ML21154A051
| ML21154A051 | |
| Person / Time | |
|---|---|
| Issue date: | 05/20/2021 |
| From: | Derek Widmayer Advisory Committee on Reactor Safeguards |
| To: | |
| Widmayer, D, ACRS | |
| References | |
| NRC-1519 | |
| Download: ML21154A051 (378) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Docket Number: (n/a)
Location: teleconference Date: Thursday, May 20, 2021 Work Order No.: NRC-1519 Pages 1-277 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 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5 (ACRS) 6 + + + + +
7 FUTURE PLANT DESIGNS SUBCOMMITTEE 8 + + + + +
9 THURSDAY 10 MAY 20, 2021 11 + + + + +
12 The Subcommittee met via Teleconference, 13 at 9:30 a.m. EDT, Dennis Bley, Chair, presiding.
14 COMMITTEE MEMBERS:
15 JOY L. REMPE, Co-Chair 16 VESNA B. DIMITRIJEVIC, Co-Chair 17 RONALD G. BALLINGER, Member 18 VICKI M. BIER, Member 19 DENNIS BLEY, Member 20 CHARLES H. BROWN, JR. Member 21 GREGORY H. HALNON, Member 22 WALTER L. KIRCHNER, Member 23 JOSE MARCH-LEUBA, Member 24 DAVID A. PETTI, Member 25 MATTHEW W. SUNSERI, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 ACRS CONSULTANTS:
2 MICHAEL CORRADINI 3 STEPHEN SCHULTZ 4
5 DESIGNATED FEDERAL OFFICIAL:
6 DEREK WIDMAYER 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 CONTENTS 2 Opening Remarks . . . . . . . . . . . . . . . . . 4 3 Staff Introduction . . . . . . . . . . . . . . . 5 4 White Paper - Risk-Informed and Performance-Based 5 Human-System Considerations for 6 Advanced Reactors . . . . . . . . . . . . . 10 7 10 CFR Part 53 - Subpart A - General Provisions 145 8 10 CFR Part 53 - Subpart F -Section 53.700, 9 Operational Objectives and Controls 10 on Equipment . . . . . . . . . . . . . . 206 11 10 CFR Part 53 - Subpart F - Section 53.800, 12 Programs . . . . . . . . . . . . . . . 241 13 Adjourn . . . . . . . . . . . . . . . . . . . . 277 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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4 1 P-R-O-C-E-E-D-I-N-G-S 2 9:30 a.m.
3 CHAIR BLEY: Good morning. The meeting 4 will come to order. This is a meeting of the Advisory 5 Committee on Reactor Safeguards Subcommittee on Future 6 Plant Designs. I'm Dennis Bley, Chairman of the 7 Subcommittee.
8 ACRS members in attendance are Joy Rempe, 9 Ron Ballinger, Walt Kirchner, Dave Petti, Vesna 10 Dimitrijevic, Jose March-Leuba, Vicki Bier, and Greg 11 Halnon. Our consultants Mike Corradini and Steve 12 Schultz are also in attendance.
13 Derek Widmayer of the ACRS staff is the 14 designated federal official for this meeting and Kent 15 Howard of the staff is the backup designated federal 16 official.
17 The purpose of today's meeting is to 18 continue our discussions with the staff on the 19 preliminary rule language for 10 CFR Part 53, 20 Licensing and Regulation of Advanced Nuclear Reactors.
21 Today's subjects include Subpart A, 22 General Provisions, and Subpart F, Requirements for 23 Operations. We will also discuss the staff White 24 Paper, Technology-Inclusive Performance-Based Human-25 System Considerations for Advanced Reactors.
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5 1 The Subcommittee will gather information, 2 analyze relevant issues and facts and formulate 3 proposed positions and actions as appropriate. This 4 Subcommittee meeting is the sixth of a series 5 scheduled to discuss the preliminary proposed rule 6 language. The ACRS Full Committee has not yet 7 scheduled a full committee meeting to discuss these 8 topics.
9 The ACRS was established by statute and is 10 governed by the Federal Advisory Committee Act, FACA.
11 The NRC implements FACA in accordance with its 12 regulations found in Title 10 of the Code of Federal 13 Regulations, Part 7. The Committee can only speak to 14 its published letter reports. We hold meetings to 15 gather information and perform preparatory work that 16 will support our deliberations at a full committee 17 meeting.
18 The rules for participation at all ACRS 19 meetings including today's were announced in the 20 Federal Register on June 13th of 2019. The ACRS 21 section of the U.S. NRC public website provides our 22 charter, bylaws, agendas, letter reports and full 23 transcripts of all Full and Subcommittee meetings 24 including the slides presented. The meeting notice 25 and agenda for this meeting were posted there.
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6 1 As stated in the Federal Register notice 2 and in the public meeting notice posted to the website 3 members of the public who desire to provide written or 4 oral input to the Subcommittee may do so and should 5 contact the designated federal officials five days 6 prior to the meeting, as practicable.
7 Today's meeting is open to public 8 attendance and we have received one request to provide 9 an oral statement on the proposed rulemaking language 10 from the U.S. Nuclear Industry Council. We have also 11 set aside time in the agenda for spontaneous comments 12 from members of the public who are attending or 13 listening to our meetings.
14 Due to the COVID pandemic today's meeting 15 is being held over Microsoft Teams for ACRS and NRC 16 staff attendees. There is also a telephone bridge 17 line allowing participation of the public over the 18 phone.
19 A transcript of today's meeting is being 20 kept, therefore we request that meeting participants 21 on the bridge line identify themselves when they are 22 asked to speak and speak with sufficient clarity and 23 volume so that they can be readily heard.
24 At this time I ask that participants on 25 Teams and on the bridge line keep their devices muted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 1 to minimize disruptions and un-mute only when 2 speaking. I need to emphasize that for those 3 listening via the bridge lines. Please silence your 4 microphones now. If they're left open, we may have to 5 silence the whole bridge line and we then have trouble 6 taking comments.
7 We will now proceed with the meeting. I 8 call on John Segala, Chief of the Advanced Reactor 9 Policy Branch of NRR, to make opening remarks.
10 And, John, if you would introduce your 11 next speaker when you're finished. John, the floor is 12 yours.
13 MR. SEGALA: Thank you. Consistent with 14 the Nuclear Energy, Innovation and Modernization Act, 15 we are committed to developing a technology-inclusive, 16 risk-informed and performance-based regulatory 17 framework for a wide range of advanced reactor designs 18 and publishing the final Part 53 rule by October of 19 2024 in accordance with the Commission's directed 20 schedule.
21 We are committed to a regulatory framework 22 for advanced reactors that achieves the goals of the 23 Commission's Advanced Reactor Policy Statement and the 24 NRC's Principles of Good Regulation.
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8 1 engagement where we release preliminary rule language 2 to solicit feedback to better inform the staff's 3 proposals and to ensure a shared understanding of what 4 will be included in the final rule.
5 As we are considering changes to the 6 previously released preliminary rule language we want 7 to ensure that we have appropriately considered the 8 feedback we have received from all stakeholders 9 including the public, industry, standards development 10 organizations, trade groups, non-governmental 11 organizations, and the ACRS. Since we are at the 12 early stages of the rulemaking process the draft 13 preliminary rule language will remain open for 14 discussion as the staff works towards providing the 15 Commission a proposed rule.
16 We are here today in the sixth ACRS 17 meeting to seek feedback on NRC's development of Part 18 53 preliminary proposed rule language for advanced 19 reactors. We previously briefed the ACRS Subcommittee 20 in January, February, March and April of this year on 21 preliminary rule language for Subparts B, C, D, E and 22 F of Part 53. Then in May we briefed the Full 23 Committee providing an overview of Part 53 structure 24 and the second iteration of the preliminary rule 25 language for Subparts B and C. We understand that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 ACRS is in the process of finalizing an interim letter 2 describing their conclusions and recommendations on 3 the released Part 53 rule language.
4 Today, as was mentioned, we plan to 5 discuss a White Paper on risk-informed and 6 performance-based human-system considerations for 7 advanced reactors and new Part 53 preliminary proposed 8 rule language for Subpart A, General Provisions, which 9 includes definitions; and Subpart F, Requirements for 10 Operations. We are looking forward to hearing any 11 insights and feedback from the ACRS today.
12 Thank you. That completes my opening 13 remarks.
14 And with that I guess I'll turn it over to 15 Bill Reckley.
16 MS. VALLIERE: Yes, John, I think Bill may 17 be having some technical difficulties, so I'll take 18 over in his stead.
19 MR. SEGALA: Okay.
20 MS. VALLIERE: If we can go to slide 2, 21 please.
22 CHAIR BLEY: This is Ms. Valliere, right, 23 Nan?
24 MS. VALLIERE: Yes, apologize. This is Nan 25 Valliere, yes.
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10 1 So as has been mentioned, this agenda 2 illustrates the topics we intend to cover today. As 3 John just mentioned, the White Paper on Human-Systems 4 Considerations, the Subpart A General Provisions where 5 we will focus on some particular definitions in that 6 section, and then after two portions of Subpart F 7 Operations relating to controls on equipment and 8 operational programs.
9 Next slide, please? This is our typical 10 life cycle slide showing our plan to develop Part 53 11 and just highlighting the fact that today we'll be 12 covering the general provisions in Subpart A and those 13 portions of Subpart F that I just mentioned.
14 Next slide, please? And this is our also 15 usual slide to show our planned and past interactions 16 with the ACRS. Here we are in May today and you'll 17 see the topics we intend to discuss in the general and 18 administrative area, and the operations area.
19 And next slide, please? With that I am 20 going to turn the microphone over to Juan Uribe and 21 Jesse Seymour to go over the Human-System Operation 22 White Paper. I'm not sure -- Juan or Jesse, who's 23 starting?
24 MR. URIBE: Good morning. This is Juan.
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11 1 audio is good?
2 MS. VALLIERE: Yes, it's good, Juan.
3 MR. URIBE: Okay. Perfect. So good 4 morning, members of the Subcommittee and everyone in 5 attendance. As Nan alluded to my name is Juan Uribe.
6 I'm a project manager in the Advanced Reactor Policy 7 Branch, which is part of DANU in NRR.
8 So as the lead project manager in 9 conjunction with Jesse Seymour, who's a reactor 10 operations engineer in the Division of Reactor 11 Oversight; and he's the lead technical reviewer and 12 main author of the paper, we will be both discussing 13 and overview of the contents described in the draft 14 White Paper that John and Nan alluded to titled Risk-15 Informed and Performance-Based Human-System Operation 16 Considerations for Advanced Reactors.
17 But before we move on I also want to 18 acknowledge the contributions from the rest of the 19 working group members for the development of this 20 paper.
21 And so hopefully by the end of the 22 presentation the hope is that you have a background 23 and a good understanding of the paper which at this 24 early stage its main purpose is really to solicit 25 feedback and comments from the stakeholders and start NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 the discussions.
2 So with that, next slide, please? So to 3 accomplish today's presentation we'll first start with 4 the background. We'll cover the nexus to 10 CFR 53 5 and really trying to answer the question that's always 6 asked, which is how does this fit and what -- how does 7 this connect to Part 53? After that we'll go over the 8 White Paper at a high level. We'll talk about sort of 9 the next steps and then we'll also have some generous 10 time for additional questions or comments that we may 11 have at the end.
12 So with that in mind, let's get started.
13 Next slide, please? Thank you.
14 So as John alluded to in his opening 15 remarks, for advanced reactor activity it's really 16 always prudent to start with NEIMA as the background 17 and what it required the NRC to accomplish. So as 18 you're aware NEIMA was signed into law January of 2019 19 and required the NRC to complete a rulemaking to 20 establish a technology-inclusive regulatory framework 21 for optional use for commercial advanced reactors by 22 October of 2027.
23 In the SRM the Commission approved the 24 staff's proposed rulemaking plan described in the SECY 25 paper and accelerated the completion timeline of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 1 proposed rulemaking to October of 2024. And as you're 2 aware, currently the staff is addressing NEIMA by 3 developing the proposed 10 CFR 53 rulemaking.
4 So while the final rule will set the 5 requirements over the different subparts, we recognize 6 that in some cases it's really the guidance and 7 support of the proposed rule that maybe the driving 8 factor to meet technology-inclusive performance-based 9 criteria that defines what we hope is a modern risk-10 informed graded approach. And so with that in mind 11 the draft White Paper to be discussed today was 12 developed under that premise.
13 The White Paper provides background and 14 initial considerations on key areas of advanced 15 reactor operations. And as I mentioned before, really 16 meant to start the discussion and solicit feedback 17 from external stakeholders such as the ACRS 18 Subcommittee today.
19 The topic of operations in general and how 20 different new advanced reactors may operate from the 21 more traditional large light water reactors has been 22 around for some time, and so I'll just list a couple 23 of different papers where the topic has been brought 24 up.
25 But on March 2020 the NRC issued SECY 10-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 1 0034, which is the potential policy licensing and key 2 technical issues for SMRs. So the requirements for 3 operator staffing for SMRs or for small or multi-4 modular facilities was discussed there.
5 On November of 2019 NEI also submitted a 6 White Paper related to microreactor policy issues and 7 there was a dedicated appendix to operator 8 requirements for automatic and remote operations.
9 And lastly, on October of 2020 NRC issued 10 SECY 20-0093 which described policy and licensing 11 considerations for microreactors where the staff 12 addressed challenges in the areas of staffing, 13 training, qualification requirements, and also 14 autonomous and remote operations.
15 And so the bottom line -- the current need 16 to develop a regulatory framework for advanced 17 reactors based on NEIMA and a recognition of the 18 likely operational deltas described in documents such 19 as those that I referenced led the staff to review 20 these topics in sort of a new and novel lens in the 21 context of the Part 53 activities.
22 Next slide, please? So with that, on 23 March 25th of 2021 the NRC released a White Paper, 24 Risk-Informed and Performance-Based Human-System 25 Operation Considerations for Advanced Reactors, to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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15 1 begin the stakeholder discussions. For the interested 2 folks the accession number for the White Paper is 3 listed on the bottom of the slide.
4 So the figure on the right has been 5 regularly presented by the Part 53 Working Group and 6 just a couple of slides ago from Nan again. And so 7 this gives you a high-level overview of the different 8 subparts that make up the rule. And really if you 9 look at the box highlighted in red under Subpart F, it 10 discusses staffing and programs. So really this is 11 where we see this paper and its evolution into final 12 guidance as supporting Subpart F.
13 CHAIR BLEY: Juan, this is Dennis Bley.
14 MR. URIBE: Yes?
15 CHAIR BLEY: Your second bullet on this 16 slide, since this will be attached to our transcript 17 as an official record, it says you introduced this to 18 the Subcommittee in March of '17. My understanding 19 was you were going to present it to us in April, but 20 it got delayed until this month. So this is the first 21 time we've had a discussion on your White Paper.
22 MR. URIBE: Yes, and I apologize if the 23 bullet is misleading. I guess I was going to cover 24 that right now. But really the intent here was that 25 on March 17th the Part 53 Working Group provided a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 1 table that key guidance in support of the rule. And 2 this was one of the entries listed there.
3 So you are absolutely correct. This is 4 the first time that we're presenting this. We were 5 originally going to provide this overview in April and 6 it got moved to May. So I apologize for the 7 misleading bullet. But I guess the intent was just to 8 say that this White Paper has been identified as key 9 guidance by the NRC staff and it was presented as such 10 in the table along with numerous other guidance on the 11 March 17th Subcommittee meeting. Does that clarify 12 the question?
13 CHAIR BLEY: It does. Thank you.
14 MR. URIBE: Okay. So again I apologize 15 for the lack of clarity on that bullet.
16 So really what that means is -- because 17 it's key guidance what that means is that based on the 18 limited or the finite amount of resources and 19 availability that we have -- this is guidance that we 20 would prioritize such that it is completed or close to 21 final upon completion of the rule on October of 2024.
22 So diving a little bit deeper into the 23 paper, the topics addressed: diverse and novel 24 operational characteristics including automation of 25 operations, staffing and qualifications of operations NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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17 1 personnel, evolution in control room concepts, and the 2 application of human factors engineering. So Jesse 3 will dive into these topics shortly. I just want to 4 cover one more slide before handing it over.
5 So next slide, please? Thank you. So 6 this is the last slide as I mentioned before we get 7 into the bulk of the discussion, and it's really just 8 another way of presenting to you perhaps the same 9 information on how the pieces fit together. It's 10 trying to address the question that we always get 11 about how this relates to the rule structure.
12 So this slide is once again attempting to 13 dissect Part 53, specifically Subpart F, and really 14 what is shown is a proposed draft of the table of 15 contents of Subpart F. I will note that the actual 16 rule numbering structure may have changed from what is 17 shown here and will continue to change give that Part 18 53 is a moving target, but for illustration purposes 19 I think it serves to get the message across. And so 20 the red box really represents where we see the White 21 Paper and eventual final guidance supporting the Part 22 53 rule. So in summary, once again it's intended to 23 support the operations portion of Subpart F.
24 Regarding Subpart F, on April 22nd at the 25 ACRS Subcommittee meeting the Part 53 Working Group NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 1 provided a breakdown on the different parts of the 2 rule including Subpart F, which one of them was 3 operations. For Subpart F as well the Facility Safety 4 Program was first discussed on the January 7th, 2021 5 meeting. Subpart F operational objectives and 6 programs will be discussed later today. And I believe 7 the proposed staffing operations language will be 8 released in the coming weeks or months.
9 So with that in mind the last thing I'll 10 mention is that the guidance in this White Paper has 11 the primary purpose of informing and supporting the 12 Part 53 rulemaking to support Subpart F, but a 13 secondary goal is to facilitate the consistent 14 treatment of advanced reactor applications that may be 15 received or are expected to be received prior to the 16 Part 53 being finalized and that would come in under 17 Part 50 or Part 52 regulatory pathways such as the 18 advanced reactor demonstration plants.
19 CHAIR BLEY: Juan?
20 MR. URIBE: Yes?
21 CHAIR BLEY: I have a question for you:
22 We don't usually get into the legal aspects of 23 licensing, but here because of the way this rule is 24 being written and now this is our first look at some 25 of the new guidance, though it might be in an early NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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19 1 draft phase, if words such as these were in the 2 statements of consideration, my understanding is they 3 are then -- they define the way the rule should be 4 interpreted. If they're in either a NUREG or some 5 other Reg Guide or something like that, then they 6 indicate things that the staff would find acceptable, 7 but they don't really become part of the requirements 8 of the rule. Is that a correct interpretation?
9 MR. URIBE: So I'll look to Bill and Nan 10 to chime in if I mischaracterize, but the expectation 11 is that the rule would set the policy, whether it's in 12 the actual rule or the statements of consideration, 13 and the guidance would simply expand upon a way to --
14 one acceptable way to meet the policy described in the 15 rule.
16 CHAIR BLEY: And the intention that this 17 would be -- in one of those forms become guidance?
18 MR. URIBE: Yes.
19 CHAIR BLEY: Okay. Thank you.
20 MR. URIBE: And so just to wrap up, again 21 the goal is at the end to have guidance that is 22 flexible, that is technology-inclusive, that provides 23 for a performance-based scaled review that is 24 commensurate with the technology and the risk hazards.
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20 1 that we can go ahead and dive into the paper.
2 Jesse?
3 MEMBER BROWN: Before you leave, this is 4 Charlie Brown --
5 MR. URIBE: I'm sorry. Go ahead.
6 CHAIR BLEY: -- you made the comment on 7 this little table that you had over here that the part 8 numbers are in flux, like the 53.8 or the 53.9 or so 9 on. Is that what -- I was just looking back. I just 10 happened to have Part F open that was submitted to us 11 and the numbers have changed since -- based on this.
12 Is that going to be an ongoing type change as you go 13 through this?
14 MR. URIBE: So again I'll look to Nan or 15 Bill so I don't mischaracterize, but I think yes, the 16 snapshot in time for when we put together this slide 17 -- as you're aware, Part 53 evolves pretty much on a 18 daily basis. And so really more than focusing on the 19 actual numbers themselves it's more on the structure 20 of how Subpart F has the staffing section. And this 21 is really what it's trying to support. So I believe 22 the short answer is yes, the numbers have changed. It 23 will continue to evolve at some point. And so really 24 it's more on the particular aspect of Subpart F that 25 we're trying to support.
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21 1 MEMBER BROWN: Okay. I just noticed like 2 security is now 53.95. In the previous document we 3 got for review it was 53.83. So I just wanted to make 4 sure I knew what -- the words are the same; the 5 numbers may change, or whatever?
6 MR. URIBE: Yes. And I'll even take it a 7 step further to say that perhaps what I have is the 8 more outdated version.
9 MEMBER BROWN: Oh.
10 MR. URIBE: What Part 53 is saying is --
11 what the working group is saying would be the more up-12 to-date or current version.
13 MEMBER BROWN: Okay. Thank you.
14 MR. URIBE: Yes.
15 Jesse?
16 MR. SEYMOUR: Next slide, please?
17 Good morning. My name is Jesse Seymour 18 and I'm a human factors technical reviewer and 19 operator licensing examiner in NRR's Division of 20 Reactor Oversight in the Operator Licensing and Human 21 Factors Branch. I'm one of the staff members who 22 developed the paper we'll be discussing and I'm going 23 to present an overview of the recent draft White Paper 24 that was issued.
25 My presentation will follow along the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 1 paper's general structure and content and there 2 shouldn't be anything that I cover here that isn't 3 generally covered in the paper already. Also the 4 paper goes into much more depth on many of these 5 topics and for the sake of time I'll just be 6 summarizing some of the key points.
7 The general approach taken in the White 8 Paper was to provide a background discussion on 9 important considerations and the existing regulatory 10 framework followed by articulating objectives for the 11 Part 53 rule to address those considerations, and then 12 to propose some potential approaches and options for 13 addressing specific aspects of those objectives.
14 Some key drivers behind this White Paper 15 are our recognition that the regulatory framework for 16 advanced reactors should be capable of addressing 17 novel operational concepts for a wide variety of 18 advanced reactor technologies, that some advanced 19 reactor designs; and again I say some, may present 20 very low radiological risk and requirements and the 21 current regulatory framework for operation of large 22 light water reactors may be unnecessary for reasonable 23 assurance of safety, and lastly that the development 24 of a risk-informed, performance-based and technology-25 inclusive framework that appropriately considers the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 1 role of humans and human system integration is 2 warranted for advanced reactors.
3 Next slide, please?
4 MEMBER BROWN: Can I ask a question on 5 that page?
6 MR. SEYMOUR: Yes.
7 MEMBER BROWN: This is Charlie Brown 8 again. You made the comment that some designs may 9 present very low radiological risk. All of them use 10 uranium to generate power, isn't that correct?
11 MR. SEYMOUR: That is correct. You know 12 --
13 MEMBER BROWN: Then how ---
14 (Simultaneous speaking.)
15 MR. SEYMOUR: -- fusion technologies.
16 MEMBER BROWN: Put aside the fusion for 17 right now, but I mean how do you have -- to produce 18 any amount of power, what's the basis for saying "very 19 low radiological risk?" It seems like that's 20 advertising in some way.
21 MR. SEYMOUR: The clarifier there is just 22 a recognition that some designs may potentially 23 present very little radiological risk, and they --
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24 1 designs to present considerable improvements and 2 safety features. But the important thing, I think, to 3 point out here is that the safety case still needs to 4 be demonstrated in every application to ensure that a 5 reasonable assurance of adequate protection exists.
6 So within the context of this paper, we 7 emphasize that factors like accident consequences and 8 the incorporation of various approaches and safety 9 features and characteristics need to be carefully 10 considered and then used to inform scalable process 11 that will address each design appropriately.
12 MEMBER BROWN: So what you're really sort 13 of implying is that due to their nature, they may have 14 less chance of dispersing radioactive materials where 15 they would be a hazard as opposed to the current large 16 light-water reactors?
17 MR. SEYMOUR: I think that's an accurate 18 assessment and, you know, moreover, I would just also 19 qualify that by saying it's strictly on a case-by-case 20 basis. We're not trying to make a blanket 21 characterization of all advanced reactor technologies.
22 MEMBER BROWN: Okay. Thank you.
23 CHAIR BLEY: Okay. Let's see, which slide 24 are we on? So yes, if we can move on to the -- okay, 25 yes, that's the slide. Thank you. So continuing on, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 1 the preceding decades have witnessed evolutionary 2 changes in areas like passive safety and modular 3 construction. The wide range of technologies that are 4 under various states of development include small 5 modular reactors, non light-water reactors, and 6 fusion-based technologies. Technologies such as these 7 warrant careful consideration of design attributes 8 that represent departures from large light-water 9 reactor designs. Consistent with this, the NRC 10 recognizes the desirability of attributes such as 11 simplified safety features of a passive or inherent 12 nature, reductions --
13 CHAIR BLEY: I'm going to interrupt you.
14 I don't see your slides -- oh, wait a minute. Never 15 mind. Please go ahead.
16 MR. SEYMOUR: Okay. Thank you. So 17 consistent with that, the NRC recognizes the 18 desirability of attributes such as simplified safety 19 features of a passive or inherent nature, reductions 20 in required human actions, incorporation of defense in 21 depth and minimization of risks associated with severe 22 accidents in advanced reactor designs. Next slide, 23 please?
24 Advanced reactors could vary in size from 25 very large to very small. Such variations are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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26 1 expected to have potential implications for both 2 source term sizes and accident consequences. So how 3 does this influence our thinking regarding operations?
4 First, accident source terms can serve as a measure of 5 the efficacy of mitigation features. Secondly, 6 advanced reactor designs may present low potential 7 accident consequences. Considerations like these are 8 informative because limiting the hazard posed by a 9 reactor facility reduces the potential for accident 10 consequences, and is the most reliable means in 11 assuring safety. Next slide, please?
12 Passive safety features and inherent 13 safety characteristics have considerable potential to 14 influence the role of personnel at advanced reactors 15 facilities. As an example, passive safety features 16 tend to shift humans into more of a defense in depth 17 role. Passive safety features can still fail under 18 certain conditions, though. In comparison, however, 19 inherent safety characteristics can be considered to 20 be absolutely reliable. The incorporation of various 21 inherent safety characteristics, passive safety 22 features, and automated safety systems into a design 23 is expected to influence the concept of operations of 24 a proposed facility which by extension, would be 25 anticipated to affect the degree of emphasis placed on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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27 1 the human factors engineering aspects of an 2 application review. Next slide, please?
3 MEMBER REMPE: I guess I would like to 4 stop you hear for a minute. I know several members 5 have pointed out that there could be some sort of 6 event, perhaps an external event or something where 7 you can't always -- it seems to me, and I think other 8 members have said the same thing, that there might be 9 some sort of external event that might mean you can't 10 rely on an inherent safety feature. Have you done a 11 thorough search to make sure -- I mean you can't have 12 a barrier that might say is inherently safe fail or 13 something? I mean I'm just kind of wondering about 14 the statement that you have inherent safety 15 characteristics that are just an absolutely reliable 16 thing, there's absolutely no way may not be able to 17 depend on it.
18 MR. SEYMOUR: Yes. That is a very 19 important topic. And what I would say here is that, 20 you know, when we use the term "absolute reliability,"
21 it is important to qualify that. And what that 22 requires is a detailed consideration of a range of 23 characteristics of the features that are supporting a 24 given safety function. So an example that I'll use is 25 that, you know, control of reactivity often involves NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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28 1 reactivity feedback mechanisms that are inherent to, 2 you know, the reactor design itself in order to 3 prevent reactivity exertions. But I think it's 4 important to note -- and this is an example that 5 certain, you know, reactor coefficients can vary over 6 core life so, you know, something that appears to be 7 an inherent characteristic at the beginning of life 8 may be less effective or even perhaps not effective in 9 a given situation at end of life. So again, you know, 10 it's important to think about that carefully.
11 And additionally, you know -- and I think 12 this really kind of gets to some of the circumstances 13 that you're driving at -- for cooling or containment 14 safety functions, you know, in those cases, the 15 passive systems that we would probably encounter for 16 those, you know, could exhibit failures that might not 17 -- might render them not being absolutely reliable.
18 So in that instance, you know, if we're thinking about 19 like a seismic event or something that results in 20 piping breaks, then perhaps a system that we thought 21 was going to be, you know, absolutely reliable might 22 not be.
23 Yet at other times, if we consider a, you 24 know, reactor that post shutdown can reject enough to 25 the atmosphere around it to remain cool, then perhaps NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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29 1 that would be something that we would consider to be 2 absolutely reliable. So it's a very careful 3 consideration that would need to be done on a case-by-4 case basis to make sure that it's being treated 5 reasonably.
6 And the last part that I would, and we'll 7 talk about this a little bit more later on, is that 8 what we're talking about now has to be considered 9 within the broader context of a defense in depth 10 philosophy. So it's assumed -- you know, there's an 11 ongoing assumption throughout this paper that even in 12 instances where we're talking about an inherent safety 13 characteristic that a layered approach to defense in 14 depth exists concurrently.
15 And I think a good summary that I'll draw 16 upon later on in this discussion is that, you know, no 17 single design or operational feature, no matter how 18 robust, should be relied upon exclusively to satisfy, 19 you know, all the various layers of defense in depth.
20 So hopefully, that answers that question.
21 MEMBER REMPE: So what's the plan? Are 22 you doing to update the white paper to reflect what 23 you're saying here, because in the past when we 24 brought this up, I know -- I think it's Bill Reckley 25 would say, yeah, it's for some cases, scenarios, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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30 1 other scenarios will -- it doesn't mean it's always 2 true. So your statement here about inherent safety 3 characteristics that you've mentioned, like 4 degradation is a great example over a lifetime, 5 absolutely reliable does not say that, and you had 6 those same statements in the actual white paper. So 7 you're going to update the white paper or something to 8 reflect these points --
9 MS. CUBBAGE: Joy --
10 MEMBER REMPE: -- and considerations?
11 MS. CUBBAGE: -- Joy, this is Amy Cubbage.
12 Bill Reckley wasn't able to joint today, but I'll just 13 say that this is, you know, a little bit of a moving 14 target, different pieces involved here, and we'll 15 absolutely make sure that we have consistency with 16 whatever Part 53 -- you know, however Part 53 and 17 other guidance describes these features. It'll need 18 to be consistent.
19 MEMBER REMPE: Great. Thank you.
20 MEMBER PETTI: I have a question. When I 21 read the paper, and I read the bullet, it seems like 22 it's definitional. It must --
23 MEMBER KIRCHNER: I agree, Dave.
24 MEMBER PETTI: An inherent characteristic 25 can be considered absolutely reliable. It must mean NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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31 1 it's always there. And if it's not, then it's not an 2 inherent safety characteristic. That's how I read 3 what your -- so yes, footnote or something in the 4 paper certainly could help.
5 MEMBER KIRCHNER: Dave, this is Walt. I'm 6 thinking the same way, that either -- the definitions, 7 which we'll come to later today, is a good place to 8 make this distinction. For me, an inherent safety 9 characteristic can vary over the life of a plant, as 10 was described by Jesse. It can be a life cycle thing.
11 It can be grade. There are a lot of things. I tend 12 to think of passive safety features as those design 13 features that are engineered into the system.
14 MEMBER PETTI: Yes.
15 MEMBER KIRCHNER: And those certainly can 16 fail under certain conditions, like a seismic event or 17 something, whereas, oh, let's just pick something --
18 moderator feedback tends to be insensitive to seismic 19 external forces and such. But these things can also 20 be changed by failure of any system within the reactor 21 itself, say an influx of water to a -- this is 22 hypothetical but water in a graphite-moderated system 23 would have a major impact on the performance of the 24 system.
25 So I think maybe some further work on --
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32 1 in the category of definitions is needed to make a 2 distinction between what's inherent and what's 3 passive.
4 MS. CUBBAGE: I think all of your points 5 are well-taken. Thank you.
6 CHAIR BLEY: Amy, this is Dennis. I'd go 7 even further than Walt was. I don't imagine where 8 there's any so-called inherent safety feature that we 9 couldn't dream up a way to disrupt it. And since this 10 is a human paper, you know, maybe somebody could put 11 the wrong chemicals in the wrong place, the cooling 12 system or something. So I think some care in this and 13 really getting rid of that language about absolutely 14 reliable or qualifying it a whole lot would be pretty 15 important. It won't just show up here. This is going 16 to show up in other places.
17 MS. CUBBAGE: Right.
18 MEMBER KIRCHNER: Dennis, this is Walt.
19 I'd go further to say that based on my own experience 20 with small modular reactors, I was chastened and 21 humbled in many instances. And I personally will shy 22 away from inherent. This idea of inherent safety is 23 more a sales pitch than it is a intrinsic 24 characteristic, because these are not immutable.
25 (Simultaneous speaking.)
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33 1 MS. CUBBAGE: Well, one example where, you 2 know, a reactor may have, you know, conductive heat 3 removal to the earth and, you know, around it, yes, 4 you could envision some way to disrupt that. But then 5 the question -- and it's broader than this operational 6 paper -- is, you know, what degree of redundancy or 7 defense in depth is required for something that is 8 based on physical properties of metal in the earth.
9 You know, these things could change your degrade, but 10 you have to look at, you know, over the lifetime, are 11 you accounting for any potential degradation or any 12 potential mechanisms that could alter that and 13 accounting for that in your uncertainty. But I think 14 this is a much broader question than the topic at hand 15 today, but your comments are all very well-taken.
16 MS. VALLIERE: This is Nat Valliere. And 17 just to add on to what Amy said, as you may recall, 18 this particular area has been identified as one of the 19 areas where some guidance -- key guidance will be 20 needed to support Part 53. So we are well-aware that 21 this is an area that needs further guidance and intend 22 to work on that.
23 MR. COUDREY: This is Chris Coudrey. I'm 24 the Chief of the Operator Licensing Human Factors 25 branch in DRO. I'll just, you know, echo what's been NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 said here, but to some degree, I think the most 2 important point here is that there is another kind of 3 layer in that defense in depth that Jesse talked about 4 that's passed the passive safety features that we're 5 kind of used to reviewing. So that next kind of layer 6 would be the inherent safety characteristics that 7 Jesse and Amy have tried to describe here. So we'll 8 take a look a little bit at some of the wording 9 related to "absolutely reliable" and make sure we've 10 got a good definition of what that next layer in the 11 defense in depth would be.
12 But just for the purposes of the 13 discussion here today, just some acknowledgment that 14 there is some additional characteristics passed just 15 the passive nature of the design that we would want to 16 consider when making some of the decisions, terms of 17 the scope of the operator licensing and human factors 18 review for advanced reactors.
19 MS. CUBBAGE: Yes. Thank you. This is 20 Amy. One last thought on that is that in the case of 21 a passive system, there may be something for the 22 operator to at least verify, that a valve has changed 23 position. You know, often passive systems have 24 valves. They just don't rely on electrical power or 25 things of that nature, and you could verify that flow NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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35 1 rates are happening. But with the inherent features, 2 it's a little harder to verify. So that role of 3 operator would be -- could be different in those 4 cases.
5 MR. URIBE: This is Juan. Can I just 6 chime in. This is a really important point that I 7 guess is worth reemphasizing as well, is that the 8 paper was released and with the purpose of starting 9 the discussion, which we're doing right now. And it's 10 all very valuable and feedback. But as Amy alluded 11 to, we'll continue to refine the language in the paper 12 so that it aligns with Part 53 and additional key 13 items that develop in the area. So we're not trying 14 to redefine a term here. We'll be consistent.
15 MEMBER REMPE: I guess I opened a can of 16 worms, but I appreciate everyone's comments and that 17 the sensitivity is going to be addressed. Thank you.
18 MS. CUBBAGE: It's a very important issue 19 so, thank you.
20 MR. SEYMOUR: Okay. Next slide, please?
21 Okay. Since autonomous operation is a major point of 22 consideration for advanced reactors, let's take a 23 moment to touch upon some key points regarding 24 automation. As an overview, automation is implemented 25 in levels that span from manual operations at the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 lowest level up to fully autonomous operation at the 2 highest level. Autonomous operation, which is full 3 automation, has the potential at its maximum extent to 4 support potentially unintended reactor operations.
5 However, even in an autonomous design, there may still 6 exist the need for humans to implement manual actions 7 under certain circumstances such as for defense in 8 depth.
9 Another factor to keep in mind is that 10 under less than fully autonomous settings, while 11 automation generally enhances operational performance, 12 there still remain other operational effects that must 13 be considered as well. A prominent example is the 14 potential for operators to potentially lose 15 proficiency with taking manual control of processes 16 when needed. Next slide, please?
17 MEMBER REMPE: So this is Joy. I guess 18 I'd like to bring up another point. In the white 19 paper, you cite Canada's SLOWPOKE-2 reactor as an 20 example of a reactor with autonomous operation. And 21 I hadn't been familiar with it, so I started looking 22 a little bit into it and actually, there was a 23 SLOWPOKE-3 that was built, and then SLOWPOKE-4 through 24 some others were designed but never built because the 25 market never materialized. Did SLOWPOKE-3 also NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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37 1 operate that way, or was SLOWPOKE-2 the only one that 2 they went that route, or do you know?
3 MR. SEYMOUR: I don't know the answer to 4 that. You know, SLOWPOKE-2 was the -- that was the 5 example that some of the staff had had some, you know, 6 some exposure to and that, you know, we had an 7 awareness of. But I can't speak to the later SLOWPOKE 8 designs, unfortunately.
9 MEMBER REMPE: So it was only a 20-10 kilowatt reactor. If you could follow-up and find out 11 a little bit more, you know, why did they not if they 12 -- if SLOWPOKE-3 was not operated that way, what was 13 the decision, why did it change? I think it would be 14 worth having a little more background on that topic if 15 you have time. Is there time? I --
16 MR. SEYMOUR: Okay. Yes -- no. Thank you.
17 I appreciate you pointing me in that direction. Thank 18 you.
19 Okay. Yes, if we could move on to the 20 next slide? Okay. Advanced reactor designers may 21 desire to incorporate load following capabilities into 22 their designs. Load following, in the specific 23 instance we're going to discuss this morning, is load 24 following where a grid control center can directly 25 adjust plant output, and that's in contrast to other NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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38 1 variations of that. But again, the specific case 2 we're talking about here is where, you know, an energy 3 control center located elsewhere is actually 4 controlling the output. It's not currently practiced 5 by commercial nuclear facilities in the U.S. on 6 account of being precluded by existing NRC 7 regulations. However, that is not the case 8 internationally. And a key example of that is France.
9 This provides a body of international operating 10 experience that we can draw upon for the subject.
11 As an example, a key takeaway from 12 international practice of this form of load following 13 -- and this is a very important point here -- is that 14 a nuclear power plant needs to be able to refuse load 15 following requests if complying with those requests 16 would violate technical specifications or otherwise 17 place the plant into an unsafe condition. Next slide, 18 please?
19 MEMBER BROWN: Can I ask -- can you go 20 back to that again?
21 MR. SEYMOUR: Yes.
22 MEMBER BROWN: The last bullet, the --
23 yes. This may be a result of my lack of familiarity 24 with the commercial since my background is navel 25 nuclear operations, but the stuff I worked on before NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 1 in some of the large, large plants was load following 2 based on parallel operations. And if a request came 3 from a central location to change power levels which 4 increase or decrease the power, as long as it didn't 5 exceed the power output requirements for the plant, 6 why would that violate technical specifications as 7 long as there was some inherent limiting operation in 8 the plant that the demand is being placed on?
9 MR. SEYMOUR: So a key example that comes 10 to mind here is, you know, there can be technical 11 specifications action statements that you get into 12 that will limit the maximum power of a plant, for 13 example, to 75 percent power or 50 percent power.
14 That would be one instance. And, you know, again, you 15 know, there are conceivable ways that, you know, 16 physical settings and controls could be utilized to 17 limit that as well. Another instance might be if a 18 plant is running autonomously, then an autonomous 19 control system could, you know, also enforce that as 20 well.
21 But the -- you know, the key -- I think 22 the key consideration here is just that, you know, if 23 an energy control center is going to try to drive a 24 plant up to its maximum output, you know, rated output 25 of 100 percent, and there is a -- you know, there's an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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40 1 administrative limit that's been placed on the plant 2 due to a technical specification of, you know, 75 3 percent or something like that, you wouldn't want to 4 be in a situation where the energy control center 5 could just kind of promptly drive you passed that 6 limit.
7 MEMBER HALNON: Hey, Charlie, this is 8 Greg. Another example is the low leakage cores on the 9 bigger reactors now have actual power-shaping and 10 quadra power tilt type tech specs that if you reduce 11 power and you have to do it quickly, you have to do it 12 with rods, you push the power profiles around to where 13 you could actually affect some of the curves that 14 you're operating to. So that's another example that 15 could possibly require review.
16 MEMBER BROWN: You're talking peaking 17 factors can change across the breadth of the core?
18 MEMBER HALNON: Yes. Ultimately, that's 19 what the tech specs are protecting, yes.
20 MEMBER BROWN: Okay. Thank you.
21 MR. SEYMOUR: Okay. If we could go ahead 22 and move on to the next slide, please?
23 Okay. Now moving on to defense in depth.
24 So what I want to do here is transition from the 25 discussion of considering, you know, the previous NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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41 1 topics now to considering the implications of defense 2 in depth operations. And this is something we touched 3 upon earlier. So the NRC has had a longstanding 4 policy of ensuring that defense in depth is 5 incorporated into the design and operation of nuclear 6 power plants. While there are a number of important 7 aspects of defense in depth, there are two key 8 philosophical principles that I wish to emphasize 9 here. Again, we did touch upon these briefly prior to 10 this.
11 So within the context of our present 12 discussion, it should be recognized that defense in 13 depth approaches should not rely solely upon a single 14 operational feature, and neither should they rely 15 excessively upon human actions. And again, those are 16 some key points that I just want to bear in mind as we 17 move forward with the discussion. So we can go ahead 18 and move on to the next slide, please?
19 Okay. So with regards --
20 CHAIR BLEY: Jess.
21 MR. SEYMOUR: Yes.
22 CHAIR BLEY: Do we have any hint of what 23 relying excessively on human actions means?
24 MR. SEYMOUR: That topic really is 25 currently at the level of that statement for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 1 purposes of this white paper and its scope.
2 CHAIR BLEY: It's pretty ---
3 (Simultaneous speaking.)
4 MR. SEYMOUR: You know -- yes, it is.
5 It's a general philosophical principle that, you know, 6 at least for this aspect of our work, you know, we're 7 trying to, you know, keep in mind as a goal. Part of 8 the consideration there is that it's a philosophical, 9 you know, position. It's a policy position. It is 10 not something that is, I would say, you know, 11 implemented into existing regulation in a very concise 12 and consolidated manner, you know, in a way that 13 would, I think, drive that particular point.
14 So again, what we're trying to do is make 15 sure that, you know, we stay aligned with the spirit 16 of that philosophical stance. But unfortunately, you 17 know, it's not much more concrete at this particular 18 moment.
19 MEMBER KIRCHNER: Can we go back, Jesse, 20 to the previous slide, to 16? Yes. Often in the 21 past, it wasn't expressed like this. It was expressed 22 as defense against human errors, not actions. In 23 other words, the system was designed to prevent human 24 errors, exceeding safety limits, etcetera, etcetera, 25 not rely excessively. We already have an advanced NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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43 1 reactor policy statement that expresses that bullet in 2 a proactive manner, but the defense in depth often is 3 used to ensure that there is defense against errors.
4 Do you see the distinction I'm making? The advanced 5 reactor policy statement, like your bullet, would 6 drive you to have less reliance on human actions, 7 intervention and so on. The defense in depth would 8 drive you to prevent human error.
9 MR. SEYMOUR: Yes. And what I -- I guess 10 how I would characterize what we are trying to say 11 here is, you know, these were specific elements of 12 broader points on defense in depth that had been, you 13 know, articulated elsewhere. And again, I think what 14 the intent was here was just to, you know, highlight 15 some specific points that, you know, we wanted to use 16 to drive such a conversation within this paper. But 17 it wasn't -- what we weren't trying do was, I think, 18 take an exhaustive approach to, you know, mapping out 19 kind of the full extent of defense in depth policy.
20 You know, what we'll see later on is that, 21 and even thus far, this paper is dealing heavily with 22 automation and autonomous operations. So again, you 23 know, within that context, what, you know, the thought 24 process was that -- you know, that second bullet 25 points at an aspect of the defense in depth, you know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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44 1 philosophy that would tend to become potentially more 2 prominent within those context. You know, if you have 3 an autonomous facility, if you have a facility that's 4 relying more heavily upon, you know, passive safety 5 features and things of that nature.
6 MEMBER BROWN: This is Charlie again. Did 7 that answer you? I didn't mean to interrupt. Walt?
8 MEMBER KIRCHNER: No. This is something 9 that is -- will continue. That's fine. Thanks, 10 Charlie.
11 MEMBER BROWN: Okay. Let me -- maybe I'm 12 speaking out of turn a little bit but in the -- all my 13 past experience, we always preached to the operators 14 of our ships that they were the first line of defense 15 and not only were they the first line of defense, they 16 were the last line of defense, because when all else 17 fails, they're the only ones left standing to do 18 something. Just -- I just hate to see us expressing 19 this idea that we don't rely on human action because 20 to me, they are, in my mind, both the first and last 21 lines of defense. And to state that you don't want t 22 rely excessively, I agree that you want to have the 23 automation take care of stuff. But the operator 24 sitting around all the time, he's still got to be 25 capable of shutting down the plant and point NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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45 1 something, whether it's a manual mode or a switch that 2 gets turned or what have you. I just -- that emphasis 3 just begs me to not think about it very positively.
4 Just my observation. That's all.
5 MR. SEYMOUR: To answer that point, I do 6 understand, you know, the sentiment and the position.
7 Again, you know, what I can do is kind of harken back 8 to the underlying intent here. You know, I think the 9 operative word is "excessive" reliance and, you know, 10 when we think about the human role in defense in 11 depth, if we have a situation where there is, you 12 know, an active safety feature that should insert, 13 say, a reactor trip but the reactor fails to trip, 14 then, you know, we absolutely want the human that's 15 there to have the tools and, you know, the ability to 16 insert that trip. So what we're trying to say doesn't 17 override anything like that.
18 Again, you know, what we're doing is we're 19 trying to highlight what -- I think what we're seeing 20 to be some of the, you know, perhaps more prominent 21 aspects of defense in depth philosophy within this 22 realm of, you know, advancements and the types of 23 safety features that are being used and, you know, 24 degrees of automation and so forth and just to kind 25 of, you know, shine light on those moving through.
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46 1 And again, you know, part of the nature of 2 this paper is that -- and this kind of gets back to 3 the discussion we were having on safety 4 characteristics, the inherent, the passive -- is, you 5 know, what we weren't trying to do is, I think, have 6 an exhaustive treatment of, you know, each one of 7 these topics but rather to, you know, lay a basic 8 framework that would be suitable for a really wide 9 variety of audiences in, you know, the first part of 10 the paper and then, you know, just to kind of 11 establish like a baseline level for some of the 12 concepts and then to move into more detailed 13 discussions.
14 Unfortunately, it's kind of a limitation 15 of the paper that, you know, we aren't more exhaustive 16 in some of these areas. So, you know, again, I'll 17 definitely note that as a potential improvement going 18 forward that, you know, we can probably stand to have 19 a more comprehensive and thorough treatment of some of 20 these background topics.
21 MEMBER BROWN: Shouldn't, in reality, 22 though, that every automated system you had in place 23 to take corrective action, protective action, or any 24 other action should always have a manual means to be 25 available to an operator? And that's not emphasized.
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47 1 I haven't seen that. I mean this has come up many 2 times, all these inherent features and we keep 3 ignoring the fact that inherent features can be very, 4 very passive. They may ought to be based -- they may 5 ought also include automated reactor trip or other 6 type stuff that you have to have some capability in 7 the long-term to do, and if they don't operate, they 8 should have some manual capability. Every automated 9 system should be able to overridden by manual 10 operations. And that's not emphasized anywhere.
11 MR. URIBE: So this is Juan. Can I chime 12 in, Jesse?
13 MR. SEYMOUR: Yes, by all means.
14 MR. URIBE: So I think this goes back to 15 as an applicant comes in, the higher degree of 16 automation that they're willing to incorporate into 17 their plan, I think that's the lens in which we would 18 evaluate that particular application. And we would 19 likely ask the same questions that are being asked 20 here to make sure, once again, that there is means of 21 having adequate protection.
22 MEMBER BROWN: I just provided the idea 23 that automation in terms of -- you made the point 24 maybe in a previous slide -- automation is great but 25 you made the point that you can become so reliant that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 1 you lose proficiency. But that's one thing on an 2 operator basis and a training basis. It still should 3 be the -- a point that if you have automation, you 4 should be able to override it and override it and 5 easily without excessive operations. So it's just a 6 thought I would throw in there that we want to make 7 sure we have that capability at all times.
8 MR. SEYMOUR: You know, I would say -- you 9 know, again, I agree that -- you know, the underlying 10 concern is extremely valid that, you know, you, from 11 a defense in depth standpoint, right, you can't just 12 be relying on a single layer. And we understand that 13 an active safety feature can fail and, you know, I 14 come from an operator licensing background and, you 15 know, the expectation that we have is that a licensed 16 operator fulfilling their role will recognize the 17 failure of, you know, an automated safety system, a 18 function and will implement that function.
19 Now as we move forward to potentially 20 autonomous reactors, and as we, you know, think 21 hypothetically about, you know, potentially an 22 autonomous micro reactor, for example, that may be 23 operating -- you know, I'll say hypothetical even 24 unattended -- then, you know, it doesn't obviate in 25 any way the need to have a back. It doesn't you know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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49 1 alleviate the need to have defense in depth. What it 2 does is it now shifts the onus to replace the role of 3 the person with something else. And it's a high bar.
4 It is, you know, and that's something that, you know, 5 we need to be sensitive to. And again, you know, to 6 the point of earlier in the paper discussing, you 7 know, some of the potential pitfalls that you can get 8 into with high levels of automation, it doesn't take 9 away from, you know, automation, you know, necessarily 10 being a good thing. Automation can be a great thing.
11 What it does is it highlights, you know, 12 the need that we can't treat automation, or many 13 things for that matter, as an end all/be all. What we 14 do is we have to take a smart and informed approach 15 and go into these things with a full understanding of 16 what the inherent limitations are.
17 MEMBER REMPE: So this is Joy again. I 18 guess I am still curious about the SLOWPOKE-2 reactor, 19 because it's, I guess, at the Royal Military College 20 in Canada. And what do they do? Do they have an 21 override? It might be really good example to learn 22 about as we go forward dealing with micro reactors.
23 Do you have any more knowledge about that?
24 MR. SEYMOUR: Unfortunately, I can't speak 25 in too much more detail about SLOWPOKE-2 other than, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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50 1 you know, kind of the cursory noting it to be a 2 potential data point that we did in the paper. You 3 know, I haven't had an opportunity to dig further in 4 the licensing. I think we do have some staff that 5 have actually been able to visit SLOWPOKE-2. So it's 6 definitely something that we dig into further and 7 again, you know, it does make for a very interesting 8 point of comparison. You know, that's why we called 9 it out.
10 MEMBER REMPE: Now --
11 MR. SEYMOUR: Again, it is --
12 MEMBER REMPE: -- I didn't mean to 13 interrupt you. Go ahead.
14 MR. SEYMOUR: Oh. But, you know, it is a 15 research and test reactor so, you know, we have to, 16 you know, go into it, I think, with the understanding 17 that there is, you know, some difference in mission 18 between that and, you know, a power reactor. But 19 again, you know, it does provide a very interesting 20 data point just in terms of, you know, core design and 21 so forth. My limited understanding of that design is 22 that a key aspect of it is that, you know, the core 23 itself can't generate too much of a power excursion.
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51 1 excess reactivity. So, you know, you're just not able 2 to have too large of a power excursion with that 3 particular design, and I think that may be one of the 4 central features. But again, I just can't speak in 5 too much depth. But, you know, it is an area that we 6 could dig into further.
7 MEMBER REMPE: The insights may be limited 8 but yes, I just am curious as to the -- and again, 9 they went with larger ones after it, and so I just am 10 curious what they learned from this Canadian program.
11 Thank you.
12 MR. SEYMOUR: Thank you. Okay.
13 MEMBER KIRCHNER: Jesse, I had to deal 14 with this kind of issue almost 35 years ago, and so we 15 -- from a design standpoint, we were looking at fully 16 automated, unattended reactor deployed in a remote 17 region. And from a design choice, we looked at the --
18 it was graphite-moderated particle fuel reactor, TRISO 19 -- one, would it -- could it sustain an uncontrolled 20 reactivity insertion accident? In other words, you 21 lose control, in this case, of reflector drums. And 22 then second, having a completely independent 23 capability of shutting down and scramming the reactor.
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52 1 requirements that we employed to address this issue of 2 automation, unattended operations and such. So that 3 --
4 MEMBER REMPE: How far did the effort go, 5 Walt? Did they build any prototypes or how far --
6 MEMBER KIRCHNER: We built a prototype of 7 the core and took it critical at Pajarito Site, and 8 that's when the program ended. But from a design 9 standpoint, it addressed these kind of automation 10 issues and not having a reactor operator present. To 11 address Charlie's concern, there was always the 12 capability to do an independent scram, so you had 13 separate and diverse systems for that and then make 14 use of what was previously talked about, passive and 15 inherent safety design features to deal with the 16 consequences, for example, of a reactivity insertion 17 accident as one of the worst case scenarios.
18 MEMBER REMPE: Did they do a lot of 19 testing with the prototype, or did they just take it 20 critical and that was it?
21 MEMBER KIRCHNER: We did the physics - -
22 confirmation of the physics characteristics and that's 23 where the project ended.
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53 1 have access to them.
2 MEMBER KIRCHNER: Yes. I'll try and get 3 those reports together.
4 MR. SEYMOUR: Thank you. I really 5 appreciate that data point. Again, that's one that, 6 as we were going through this, wasn't on our radar so 7 to speak and again, thank you for pointing that out to 8 us.
9 MEMBER BROWN: Just my suggestion would be 10 somewhere along the line in this process that 11 automation is always going to be there. It's always 12 going to be the thing that takes the initial actions, 13 but somewhere we ought to be emphasizing that the 14 operator should always be able to take manual action 15 to shut down, take corrective action any reactor plant 16 that we have regardless of what it's design is. It 17 may not be in time to prevent some in core damage, but 18 it would be enough to prevent any excursions that 19 create some type of consequences outside the core 20 itself or the reactor, whatever the configuration of 21 the internal plant is. And that's -- I don't see that 22 emphasis as we go through these various subparts or in 23 the white paper. So I'll be quiet now. I think I've 24 made my point.
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54 1 Desaulniers. If I can perhaps interest Member Brown's 2 comment?
3 MR. SEYMOUR: Yes, Dave.
4 MR. DESAULNIERS: So good morning, all.
5 This is David Desaulniers. I'm a Senior Technical 6 Advisor for Human Factors and Human Performance 7 Evaluation at NRR. Thank you for the comment with 8 respect to ensuring that the operator remains in 9 control. I just wanted to, you know, acknowledge that 10 and say I consider that view to be consistent with the 11 staff's philosophy in terms of the application of 12 human factors and how we've implemented it through 13 some of our guidance. And I'll point to, for example, 14 the interim staff guidance that we have with respect 15 to computer-based procedures where we address the 16 possibility of automation being implemented through 17 computer-based procedures. And there are guidances 18 explicit with respect to ensuring that the operator 19 remains in control and can override those automated 20 procedure steps. So I have not heard of anything, as 21 we've engaged on the advanced reactor front, that 22 would suggest to use that we would be pivoting from 23 that philosophy. Appreciate the fact that it's noted 24 that it's perhaps absent in the paper. That's 25 probably something we can address.
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55 1 CHAIR BLEY: This is Dennis Bley. Thanks, 2 Dave. That was pretty good. I must say I agree with 3 Charlie and I certainly agree on the point of being 4 able to always allow the operators to put the plant in 5 a safe stable state. On the other hand, always being 6 able to override the system is far more general than 7 I'd like to see adopted. And I just point to the 8 aircraft industry where when they first went to 9 automation and fly by wire systems, you know, the 10 pilots really objected. They wanted always to be able 11 to take over. And then a majority of accidents that 12 have been investigated, when the pilots took over in 13 cases where the automation was equipped to deal with 14 fast-changing parameters, they ended up getting into 15 more trouble than hands off operation would have. And 16 they've eliminated that in many cases. So especially 17 as we get systems that are I&C systems that are 18 operating systems, you don't want the operators to be 19 able to take over under any conditions unless it's 20 just to put it in the safe stable state. So I think 21 we need to be careful about that.
22 MEMBER BROWN: Obviously, Dennis and I 23 have a little bit of disagreement on that, some of the 24 -- I guess Boeing would have a little bit of problem 25 with that right now also since they haven't been able NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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56 1 to build their planes for a while. There are limits.
2 I agree with Dennis from the standpoint you've got to 3 be careful, but taking -- you know, shutting down a 4 reactor plant is a little bit different than trying to 5 take control of an airplane at 35,000 feet. So I 6 would be a little bit cautious about how we apply 7 that.
8 CHAIR BLEY: I wasn't talking about 9 shutting it down, Charlie. I was talking about 10 maneuvering it because you think you know better than 11 the automation, and that gets us into an area which is 12 difficult. We have the advantage we're not up in the 13 air and being able to shut it down is -- I would never 14 object to it.
15 MEMBER BROWN: I agree with you on the 16 maneuvering it. You got to be careful about overdoing 17 that. So from that -- my big issue is that when they 18 see something that they don't think is in control, 19 they ought to be able to take action and shut it down.
20 That's -- and there ought to be the capability to do 21 that. You may not ever have to use it, but if it's 22 not there, then you're toast. So we're probably close 23 to the same page. And it was a good discussion. I 24 just had to bring that up based on what I've been 25 seeing. Thanks, a lot.
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57 1 MR. SEYMOUR: Okay. I'm going to go 2 ahead, if everyone's okay, and move on to the next 3 slide. So if we could go on to the next slide, 4 please?
5 Okay. So with regards to staffing, in the 6 past, the NRC staff recognized the limitations of the 7 prescriptive staffing requirements of 10 CFR 50.54M; 8 in response, developed NUREG 1791 in order to allow 9 for increased flexibility by providing guidance for 10 assessing exemption requests to the regulations of 11 that particular regulation.
12 With that being said, we recognize that 13 licensing future advanced reactors by means of having 14 to rely upon exemptions from prescriptive staffing 15 requirements may not be an efficient or practical 16 long-term regulatory framework. Thus we recognize 17 that an alternative means that is not reliant upon 18 NUREG 1791 may be beneficial, especially if such a 19 means would rely upon analyses that can be scaled with 20 a risk of the facility. Next slide please?
21 Now I'd like to turn from talking 22 specifically about staffing in general to the specific 23 topic of operator licensing. Operator licensing has 24 a long history in the U.S. with the first regulations 25 governing it dating back to around 1956. Operator NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 1 licensing has a basis in statute with the Atomic 2 Energy Act of 1954 requiring the NRC to prescribe 3 uniform operator licensing conditions. Today all 4 license exams are approved and administered by the NRS 5 staff. However, again, we realize that advanced 6 reactor operational concepts may not align well with 7 the existing power reactor operator licensing 8 framework we have established for power reactors.
9 Some examples of changes that might be 10 appropriate for advanced reactors could include things 11 like making allowances for varying the licensing 12 examination scope on a facility-specific basis and 13 modifying simulator requirements. Keep in mind that 14 these are just some high-level examples though.
15 Bottom line, we believe that a revised approach to 16 operator licensing should be able to flexibly and 17 efficiently address a wide variety of advanced reactor 18 designs. Next slide, please?
19 Okay. So now shifting the focus slightly 20 but still, you know, within the arena of staffing, I'd 21 like to turn to the shift technical advisor position.
22 Staffing at power reactors also includes the shift 23 technical advisor, or STA, position. Unlike licensed 24 operators, STA requirements are primarily rooted in 25 Commission policy and not in regulation or statute.
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59 1 A current STA policy interpretation is that on each 2 shift, there should be at least one person on duty who 3 has a degree in physical science, engineering, 4 engineering technology, or a professional engineering 5 license. The function of this person is to provide 6 independent engineering expertise, accident 7 assessment, and technical advice as the main control 8 room operators. For now, the key takeaway that I like 9 to point out here is that the elimination of the STA 10 position at a power reactor facility would be a 11 departure from existing Commission policy as well as 12 from longstanding Agency and industry practice. Next 13 slide, please?
14 MEMBER KIRCHNER: Pardon my interruption, 15 Jesse.
16 MR. SEYMOUR: Oh, yes.
17 MEMBER KIRCHNER: This is Walt Kirchner.
18 I think on that previous slide on STA, you kind of 19 overstate, yes, this would be a departure from what's 20 going on in the current fleet, but the previous policy 21 that the Commission adopted did allow, if not for a 22 direct elimination of the STA, at least a combination 23 of the STA functions with that of the Senior Reactor 24 Operator. So there are provisions that allow for a 25 reduction in staffing with regard to how the STA NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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60 1 functions are performed.
2 MR. SEYMOUR: Yes, and I apologize. You 3 know, again, my summary did, I believe, tend to over 4 generalize things but absolutely, I think it's very 5 important to highlight that there's a distinction 6 between some of the options that are available for 7 implementing the STA role and its outright 8 elimination. And one of them that's quite common in, 9 you know, the existing fleet is that the STA will be 10 combined with another role. So there may be an SRO 11 who's fulfilling another duty on shift who also has 12 the STA -- you know, they have a degree and they have 13 a qualification to perform the STA role. And, you 14 know, many times if there's an off normal situation 15 that occurs, you know, they'll shuffle into that STA 16 role and fulfil that. So -- and again, that's one of 17 the ways of implementing that such that you're not 18 necessarily having to add a dedicated, you know, 19 person to the shift.
20 You know, I was an STA at a site earlier 21 in my career and, you know, at that particular site, 22 we used a dedicated STA, but here were other sites, 23 you know, our fleet, I believe, where they combined 24 that role with another person.
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61 1 to point out here, is that what we're talking as kind 2 of a pure case of just eliminating, you know, the 3 role, and that's where we point existing policy. And, 4 you know, going through - -and in the paper, we try to 5 go through chronologically the different policy 6 statements that had been laid out. And, you know, I 7 believe the earliest one does, you know, mention that 8 factors like upgrades to training and human machine 9 interfaces could be considerations. And then -- but 10 that we consider within the context of later policy 11 positions.
12 MEMBER KIRCHNER: Jesse, I would just say 13 that we, as a committee, had to deal with some of 14 these considerations on staffing and STA. There will 15 be an ACRS letter report issued shortly. There is an 16 attachment to that report by Member Bley that 17 addresses the STA position, and you may find that of 18 much interest in your deliberations over the white 19 paper.
20 MR. SEYMOUR: Okay. Thank you. I 21 appreciate that point.
22 CHAIR BLEY: I guess since my name was 23 brought up, I may as well jump in. The staff wrote a 24 SECY in early April that went up to the Commission.
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62 1 that was very carefully worded. And if you weren't 2 involved with it, you ought to take a look at it as 3 well. I don't think it in any way disagrees with 4 things you've said in the newspaper but it was pretty 5 well done.
6 MR. SEYMOUR: Yes. I do appreciate that.
7 I was part of the team that, you know, was involved 8 with that. I was not the author of that, however, but 9 I do have an awareness of it and, you know, again --
10 and what I'm presenting here, I'm just trying to, I 11 think, present like a very high level, you know, 12 generalization. You know, again, that SECY is written 13 within a very specific context of, you know, the new 14 scale application that I was looking at. So I'm not 15 trying to make a -- you know, take a broader 16 generalization there but again, here just to 17 highlight, I think, some of the very high level 18 considerations.
19 CHAIR BLEY: They also raised the issue of 20 it being used as a precedent and talked about it.
21 MR. SEYMOUR: Yes. Okay. I'd like to go 22 ahead and move on to the next slide if that's okay 23 unless anyone has anything else. Okay.
24 So moving on to the subject of training, 25 so in this topic, some points that I'd like to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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63 1 emphasize that, you know, training is, you know, 2 without a doubt another very important area for us to 3 consider in the course of this work. The Nuclear 4 Waste Policy Act of 1982 directed the NRC to establish 5 regulations for the training and qualifications of 6 nuclear plant operators, supervisors, technicians, and 7 other operating personnel. So this is another place 8 where we have, you know, a statutory mandate to 9 implement certain regulations and program.
10 So this Act also directs the NRC to 11 establish requirements for simulator training 12 requalification examinations, operating tests, and 13 instructional requirements. The implications of this 14 are that certain aspects of nuclear training are 15 spelled out in statute and thus are areas in which any 16 approach that we take going forward for advanced 17 reactors, we must be adherent to.
18 So while on the subject of training, I'd 19 also like to point out something that's emphasized in 20 the paper. And what I'd like to do is just briefly 21 touch upon, you know, the systems approach to training 22 which, you know, in the industry, most refer to that 23 as a systematic approach to training. But, you know, 24 again, we're invariably talking about the same thing.
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64 1 central role in current nuclear training and 2 qualification programs. And again, we refer to that 3 program as SAT.
4 So why do we emphasize the SAT process 5 here? And the reason why we do is because SAT is, 6 generically speaking, just very useful to us. So the 7 SAT process is generic in nature, and one of its key 8 attributes and one of the reasons for its original 9 adoption was that it can be adapted to essentially any 10 reactor technology or even non-nuclear technologies.
11 And that includes, you know, essentially any 12 foreseeable advance reactor designs. So again, it's a 13 very useful tool that we've relied on in the past from 14 a regulatory standpoint because of its generic and 15 adaptable nature. And again, you know, just kind of 16 discussing it a high level here. The paper does go 17 into more detail. It's something that, you know, has 18 potential application going forward as well.
19 CHAIR BLEY: I want to interrupt you just 20 a second, Jesse. I see Bill Reckley has joined us 21 now, and I wanted to check with Bill. We're only 22 halfway through the presentation slides, maybe a 23 little more than that on this paper. But it seems to 24 be getting a better amount of good discussion. Bill, 25 I'm not sure but I suspect Subparts A and -- A NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 1 especially might not take as much time as -- we had a 2 lot of -- is there any objection to letting this go on 3 a bit more?
4 MR. RECKLEY: Not at all, Dennis, and I 5 would agree with you This is a very important topic 6 and so I would just suggest we let this go as long as 7 you want, and we can make up this afternoon --
8 CHAIR BLEY: Okay, good.
9 MR. RECKLEY: -- if we need to.
10 CHAIR BLEY: Good. That's what I was 11 thinking. Given that, I think this would be an okay 12 time to take a break, and then we'll come back and 13 continue with this presentation, because this is our 14 first time to look at any ---
15 (Simultaneous speaking.)
16 MEMBER BIER: Dennis?
17 CHAIR BLEY: -- guidance. Who's this?
18 MEMBER BIER: This is Vicki.
19 CHAIR BLEY: Hi, Vicki.
20 MEMBER BIER: Before taking a break, I 21 wanted to circle back to one point that I wanted to 22 make. I don't think we need to put the slides back, 23 but it was probably around slide 16 or so, which is I 24 think there were some human factors issues other than 25 just decline of technical proficiency. One is sort of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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66 1 loss of situational awareness if the plant really runs 2 by itself, you know, almost 100 percent. And I 3 realize that even our operating nuclear reactors, 4 people are not, you know, actively driving it the way 5 you drive a car. You know, they're still spending a 6 lot of time just sitting and observing, but whether 7 that could be exacerbated.
8 And the other is just whether there could 9 be either in training or just in culture sort of a 10 predilection for letting the plant sort itself out so 11 that even in situations where things seem to be off 12 normal, these operators are very reluctant to 13 intervene. And, you know, I'm kind of reminded of the 14 old joke about, you know, all we need is an operator 15 and a dog and the job of the dog is to keep the 16 operator from touching the controls. But, you know, 17 obviously, there's a reason we want operators and we 18 don't want them that far out of a sense of kind of 19 ownership of the plant situation.
20 So I just wanted to raise those as kind of 21 additional issues beyond just proficiency.
22 CHAIR BLEY: Thanks, Vicki, and that's a 23 really important point. And two years back, John 24 O'Hara and his colleagues at Brookhaven did a report 25 in this area for NRC and Dave Desaulniers is probably NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 1 very familiar with that, maybe some of the rest of 2 you. That may be worth taking a look at that, and I 3 think expanding the discussion a little in the way 4 Vicki was suggesting is -- would be useful.
5 In any case, at this point, we are going 6 to take a break, and let's come back at 20 past the 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. We'll start promptly at 20 past the hour.
8 (Whereupon, the above-entitled matter went 9 off the record at 10:56 a.m. and resumed at 11:20 10 a.m.)
11 CHAIR BLEY: Okay. We are now back in 12 session and I'll turn the mic back over to Jesse 13 Seymour.
14 Jesse, please go ahead.
15 MR. SEYMOUR: Yes, thank you.
16 So if it's okay I'd like to move onto the 17 next slide unless we wanted to further discuss the 18 training slide.
19 (No audible response.)
20 MR. SEYMOUR: Okay. So moving on. So the 21 next topic that I'd like to move into is specific to 22 human factors engineering considerations. And another 23 central focus of our present effort involves human 24 factors engineering. At present applying human 25 factors engineering in the design of nuclear power NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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68 1 plant control rooms is required under existing post-2 Three Mile Island regulations. Our current human 3 factors engineering reviews typically focus on the 4 human-system interfaces that are located within 5 control rooms.
6 Moving forward though should include 7 examining how human factors reviews can be implemented 8 most effectively for advanced reactors. To that end 9 new approaches such as transitioning to an efficient 10 scalable human factors engineering review process and 11 thinking beyond the confines of traditional control 12 rooms should be considered.
13 As a step towards achieving this we 14 believe that a concept of operations can be a valuable 15 tool in helping us to gain the design understanding 16 necessary to conduct appropriate human factors 17 engineering reviews for advanced reactors.
18 Next slide, please? While existing human 19 factors engineering requirements focus on control 20 rooms, we also recognize that some advanced reactor 21 facilities may not opt to utilize traditional control 22 rooms in their designs. Because of this requirements 23 addressing matters associated with control rooms will 24 need to be revisited in Part 53 with an understanding 25 that the functions involved may become decentralized NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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69 1 in an advanced reactor facility.
2 One potential implication of this is that 3 human factors engineering requirements will 4 essentially need to be able to follow important 5 functions if they are relocated outside of a 6 traditional control room. Beyond this it may also be 7 necessary to account for the potential emergence of 8 functions that have no precedent within traditional 9 control rooms in the current context as well.
10 Next slide, please? Okay. So now I'd 11 like to circle back around to our earlier discussion 12 on automation, and one of the things that we touched 13 upon there was that if we take automation out to its 14 logical extreme and we have a fully autonomous 15 facility, it raises the potential for an applicant 16 looking to get a facility licensed that doesn't have 17 licensed operators.
18 So what we've done as part of this paper 19 is we've tried to think through what some of the major 20 implications of that may be, and one of them, one of 21 the implications of removing all the licensed 22 operators from a facility is that the existing 23 regulatory framework also assigns certain 24 responsibilities of licensee organizations to their 25 licensed operators that are beyond necessarily an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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70 1 operational concept. These could be things of an 2 administrative nature, but in some cases these can be 3 what we would consider to be administrative tasks that 4 have safety significance to them.
5 So a key example here are the requirements 6 that we currently have in 10 CFR 50.54x and y for 7 making departures from licensing conditions under 8 emergency situations, specifically for protecting the 9 health and safety of the public. So that's one key 10 example. And that specific requirement allows a 11 licensed senior reactor operator to make that 12 determination and to invoke that.
13 But beyond that example though there are 14 also other numerous operator administrative 15 responsibilities that are both important to safety and 16 also derive from other regulatory requirements, and 17 those responsibilities would need to be addressed as 18 well. And examples of those include things like 19 compliance with technical specifications, making 20 operability determinations, making required 21 notifications to the Nuclear Regulatory Commission 22 when required, making emergency declarations and the 23 attendant notifications to state and local 24 authorities, and also compliance with radiological 25 release limits that are imposed upon the facility.
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71 1 So again that highlights some of the many 2 concerns that would arise in a situation where there 3 were no licensed operators present. Again, these are 4 just some of the perhaps more noteworthy examples that 5 aren't necessarily operational in nature.
6 Next slide, please?
7 CHAIR BLEY: Jesse?
8 MR. SEYMOUR: Yes?
9 CHAIR BLEY: In that area any plant also 10 has key agreements with local and state emergency 11 response organizations and I don't know if those are 12 done through NRC or if they're done separately. Can 13 you talk about that at all because that could be a 14 tricky area as well?
15 MR. SEYMOUR: So that is a bit beyond the 16 scope of the paper's context. When it comes to 17 emergency planning in general, to your point, that 18 involves more actors than just the facility and the 19 Nuclear Regulatory Commission. So when we talk about 20 the planning and coordination that happens between the 21 utility and the state and local authorities, FEMA is 22 also a player. So FEMA has a key federal role that 23 deals with particularly the local and state side of 24 that as well. And additionally, there are certain 25 aspects programmatically that are being derived from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 1 EPA requirements and so forth. So but the other key 2 federal player is really FEMA that's involved there.
3 So when those agreements are reached in 4 the course of the emergency plan being developed for 5 a site there are more potential complexities that 6 could arise there as well. That is something that 7 would have to be resolved as part of the application 8 review and the licensing process.
9 So what I would say though is for the 10 purposes of this paper we really don't go too far in 11 that direction because it's a bit outside of the 12 scope. So --
13 CHAIR BLEY: I --
14 MR. SEYMOUR: -- it's not something 15 that --
16 CHAIR BLEY: I raised this more for Bill 17 than for you because somewhere this needs to be 18 considered. You know, if I'm a local guy I don't want 19 to call a computer and find out what's going on.
20 MR. SEYMOUR: Bill, did you have anything 21 you wanted to add to that?
22 MR. RECKLEY: No. It will be something, 23 Dennis, I think when we get into programs and some 24 things like security for example, fire protection, 25 some of those other areas where off-site entities NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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73 1 might get involved that we'll have to address and have 2 to coordinate, just like you mentioned, all these 3 activities that are going on at the same time.
4 CHAIR BLEY: I guess we could (audio 5 interference) --
6 (Simultaneous speaking.)
7 MR. SEYMOUR: I'm sorry. There was a 8 little bit of I think background interference.
9 CHAIR BLEY: Somebody had a mic open. Go 10 ahead.
11 MR. SEYMOUR: Okay. Yes, so I can 12 continue on unless we wanted to -- was there a 13 question I guess for me at the last part? I 14 apologize. I didn't quite hear the last --
15 (Simultaneous speaking.)
16 CHAIR BLEY: No, just go ahead.
17 MR. SEYMOUR: Okay. Thank you.
18 Okay. So moving onto the next topic. So 19 now we've covered a broad range of background 20 considerations in getting to this particular point, 21 and at this stage -- and as I mentioned before the 22 paper approaches this in stages. There's a 23 considerable amount of material on the front end of 24 the paper that is really just kind of providing some 25 of the background discussion that we've covered.
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74 1 But now what the paper moves into after 2 this are objectives. And at this stage what I'd like 3 to do is shift gears and consider some objectives for 4 our Part 53 rulemaking that can help us to address the 5 areas that we've covered thus far and again just 6 moving through these categorically.
7 The first such objective identified in the 8 paper is that the rule should recognize that staffing, 9 training, operator licensing and human factors are 10 interrelated areas. So diverse advanced reactor 11 technology is what will also necessitate us taking a 12 more integrated approach to reviewing those areas than 13 we've done in the past, and potentially applying a 14 flexible approach that we can appropriately target and 15 scale based on a particular technology.
16 And something I'll point out too going 17 through these objectives is that these objectives are 18 presented at a high level, but we're not always 19 necessarily specifying things that necessarily need to 20 go into rule language. In some cases we're speaking 21 to things that are addressed through regulatory 22 guidance documents. So again I don't want to imply 23 that everything we're talking about here is being 24 stated in terms of being an objective for something 25 that would necessarily be articulated in the rule NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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75 1 itself. Some of this would be stuff that we would 2 consider in the development of regulatory guidance as 3 well.
4 So the rule should also account for 5 varying accident consequences for these different 6 designs and for the variation that we'd see within 7 these different technologies as the applications come 8 in. Additionally, the rule should require a human 9 factors engineering program that is adequate to ensure 10 that personnel can understand plant status, take 11 action to ensure safety, and perform other important 12 technical and administrative functions that have 13 safety implications.
14 I think that's key because we're going to 15 talk more later on about scaling our human factors 16 engineering review program and approaches that we can 17 do towards that end, but the important thing is that 18 we always have to keep our eye on the ball so to speak 19 with what are we trying to accomplish, right? And I 20 think this hearkens back to some of the comments 21 earlier by the Committee, and things that we want to 22 make sure that we have a focus on are making sure that 23 people can fulfill their safety role, and human 24 factors engineering needs to be a mechanism that 25 facilitates and enables that in a reliable and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 1 consistent way.
2 So regarding that last point, and again 3 talking about the administrative functions with safety 4 implications, so stepping a bit away from the 5 operational, the human roles associated with 6 management and availability of plant-specific safety 7 functions should be taken into account. And I would 8 say that that's of particular importance when we're 9 considering human factors engineering requirements.
10 So something that we may encounter with 11 these designs are plant safety functions that perhaps 12 don't align with the traditional large light water 13 framework that we're used to. And by that I mean we 14 think of plant safety functions in terms of reactivity 15 control, heat removal, things of that nature. But 16 again it's going to be important that our approach to 17 really not just human factors engineering, but other 18 areas as well too are cognizant to what are the safety 19 functions that may be relevant to that plant and what 20 are the human roles to play and making sure that those 21 are available?
22 Next slide, please? Okay. So continuing 23 on, the rule should account for designs that do not 24 utilize traditional control rooms. And again what 25 we're saying here is that the rule needs to be able to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 account for that. We still anticipate that many 2 designs would have a traditional control room, but 3 again it's something that -- the objective here is 4 simply stating that we need to be prepared for the 5 other possibility as well.
6 Additionally, the rule should ensure that 7 the operator licensing process accomplishes the 8 following: Okay? So these would be high-level 9 objectives that any revision to the operator licensing 10 process that we come up with would need to adhere to.
11 And first and foremost it needs to comply with 12 applicable statutory requirements because we have at 13 least two pieces of legislation that mandate certain 14 aspects of the operator licensing program. And we've 15 touched up the Nuclear Waste Policy Act previously, 16 but the other one is the Atomic Energy Act.
17 Additionally, any operator licensing 18 program to be sound needs to conform with accepted 19 testing standards.
20 We also need to be able to facilitate 21 consistent and reliable licensing decisions by the NRC 22 and a goal that we should have is also to make 23 efficient use of NRC and/or vendor licensee resources.
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78 1 need to just arbitrarily cut. The key is to be 2 informed and to be efficient and to target what we're 3 doing with an eye toward safety. So we're focusing on 4 the things that are important, but we're also staying 5 relevant to the unique technologies that perhaps we're 6 going to be encountering in some of these designs.
7 And lastly; and again this is a very, very 8 high-level prominent objective in this, at the end of 9 the day we need to be able to provide a reasonable 10 assurance that operators will be able to manage plant-11 specific safety functions.
12 Okay. So if there are no questions I'll 13 go ahead and move onto the next slide.
14 (No audible response.)
15 MR. SEYMOUR: Okay. So continuing on, the 16 rule should allow for consideration of innovative 17 features that are intended to make new designs safer 18 while also accounting for uncertainties associated 19 with new approaches. And again that's a key point, 20 too. What we don't want to do is we don't want to 21 just I think become fixated with the new thing that 22 comes along and somehow think that it's foolproof, 23 right, because our operating experience and history 24 and general tells us otherwise.
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79 1 that with new technologies there may be uncertainties 2 that are inherent in that, and again our approach 3 needs to be mindful that those uncertainties, and for 4 lack of a better way to put it, the unknown unknowns 5 can still be inherent in these new technologies as 6 they come along. So something that we need to be 7 mindful, too.
8 The rule should in a non-prescriptive 9 manner -- and again I say here in a non-prescriptive 10 manner because this part contrasts with the current 11 approach taken under 50.54 -- so the rule should 12 require staffing levels needed to support safe 13 operation and allow for the possibility of 14 demonstrating that no human presence is necessary.
15 Now there are a few layers to that: The 16 first is what we're saying -- I don't think it's a 17 radical departure from our current underlying 18 perspective because what we do now is we have a 19 prescriptive staffing model and we provide an 20 exemption process where relevant analyses can be used 21 to justify variations from that, right?
22 In this case what we're saying is that 23 that burden of proof is always there, right? You need 24 to justify using appropriate analyses, what that 25 staffing number is, but the starting point is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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80 1 different, right, so vice starting from a prescriptive 2 staffing requirement and then allowing exemptions from 3 there. What this objective is stating is that we 4 would take a non-prescriptive approach and allow the 5 staffing requirement for a facility to be driven by 6 the relevant analyses. And again it's a high-level 7 objective and this is obviously an area that we're 8 still working in.
9 But the last part is something I want to 10 highlight as well, too. And when we say the 11 opportunity for demonstrating that no human presence 12 is necessary, we're going to talk about that a bit 13 more later on in this paper. And what I want to 14 highlight is that when we talk about that possibility, 15 it's something that we see is being hypothetically 16 possible, but there would be a very high bar to get 17 over to -- for an applicant to be able to justify that 18 that was safe.
19 So is it feasible that someone could 20 demonstrate that that is the case? We feel that it's 21 possible, right, but for the safety bar to get over it 22 would be high to do it. And we'll talk a bit more 23 about some potential factors that we may consider in 24 that bar.
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81 1 prescribe minimal requirements that must be met to not 2 use licensed operators at all. And again this kind of 3 ties into that last statement that I was talking 4 about. So implicit and working your way down towards 5 not having any or very many people at a site there are 6 iterations of that where perhaps a site tries to 7 justify not having licensed operators at their 8 facility. Again, that is an area that we see to be a 9 possibility. But again there would be a safety bar; 10 and in this case we feel a prescriptive safety bar, 11 that you would have to clear to be able to justify 12 that.
13 Lastly, the rule should ensure that 14 advanced reactor defense-in-depth approaches do not 15 rely exclusively upon a single operational feature or 16 rely excessively upon human actions. And again that 17 hearkens back to our earlier discussion, and I would 18 just emphasize that again that's highlighting key 19 points in a broader defense-in-depth philosophy.
20 Okay. So if we could, I'd like to go 21 ahead and move onto the next slide.
22 MEMBER KIRCHNER: Could you stop for a 23 moment on 26? This is Walt Kirchner. So here you 24 come to what I would say is one of the hard points for 25 the rule. You say that the rule -- you say may also NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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82 1 prescribe, but it would seem to me that the rule would 2 define minimal requirements to have no operators or no 3 licensed operators. And that hearkens back to certain 4 design features or safety functions.
5 And kind of reiterating Charlie's position 6 and my own words, it seems that you would have to have 7 it independent from any control system, an independent 8 reactor protection system that could put the reactor 9 into a safe shutdown condition. And that's getting 10 prescriptive, I know, when you're trying to be 11 flexible and general, but I don't see how you get 12 beyond that bar, as you said would be very high.
13 I suspect you have to define it sooner or 14 later. And that's not guidance. That's prescriptive 15 regulation. That's one member's opinion.
16 MR. SEYMOUR: I --
17 MEMBER BIER: Hi, this is -- sorry. Go 18 ahead.
19 MR. SEYMOUR: Oh, sorry. I was just going 20 to say I think that captures our current perspective 21 on that very well, right? And we'll get to the 22 criteria that we lay out in the paper later on, but we 23 see that to be prescriptive in nature. And we'll 24 present it in a way I think that falls philosophically 25 in alignment with what you're saying, but again we see NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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83 1 that as being a place where the room for being 2 flexible kind of runs out and at that point there 3 needs to be some hard criteria that need to be met.
4 And the key thing is that we have to be able to 5 establish a reasonable assurance of adequate 6 protection.
7 And again I don't want to digress too far 8 because we'll talk about it more in a little bit, but 9 I think there's something key to keep in mind: As we 10 subtract the number of people that are available to 11 operate a site we start to transition from are there 12 enough people there to go ahead and carry out the 13 tasks that need to be done? Will there be people that 14 are available to implement human interventions where 15 they're needed? And once we subtract that last 16 operator out of the equation now we've kind of flipped 17 that thought process on its head. And now what we've 18 done is removed the opportunity for human 19 intervention, right? And that's a key transition 20 point in our mind.
21 So at that point the considerations need 22 to be fresh. And when we look at how we adjust the 23 staffing number using exemptions under the NUREG-1791 24 process that we use now, again that process isn't 25 designed to get you -- and we'll talk about this more NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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84 1 later on, but that process -- it's a very robust 2 process, but it's not designed to get you down to zero 3 licensed operators. What it does is it looks at 4 workload, it looks at tasks, it looks at many factors 5 in a comprehensive manner helps you determine how many 6 people need to be there, but it just isn't designed to 7 go to that last step.
8 And part of the reason why we think that 9 a different set of prescriptive criteria are warranted 10 to make that final jump and get to zero licensed 11 operators are just because of the safety implications.
12 And a big part of that, at least to us, is that point 13 that I articulated earlier where now when you're 14 taking that last person with the training and the 15 abilities to carry out those safety-related actions to 16 provide that defense-in-depth. Now you're taking away 17 the opportunity for human intervention, right? And 18 that's a big step.
19 So the stance that we'll talk about later 20 on is that our current perspective is that there are 21 conditions under where we think it would be 22 potentially acceptable, but it's a high bar. And 23 again I think it's going to be presented a little bit 24 differently than how you just described, but I think 25 the underlying thought process is going to be in line NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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85 1 with at least I think how you've stated it so far.
2 MEMBER BIER: Okay. This is Vicki Bier.
3 I wanted to make a point that is potentially relevant 4 to this issue of taking away the last human operator, 5 but I think may also be much more broadly relevant in 6 dealing with these totally novel designs that go 7 beyond kind of the historic structure of regulation.
8 The U.K. Health and Safety Executive, or 9 whatever it's called, has historically had a much 10 less-prescriptive approach than the U.S. NRC and 11 depends, as I understand it -- it's been a number of 12 years since I looked at it, but as I understand it, it 13 depends very heavily on what they call a safety case, 14 which basically says that the burden of proof is on 15 the operating organization to convince the regulator 16 that their plans are safe, but the regulator has 17 pretty wide discretion to accept or reject their 18 plans. It's not a checklist of do you have X amount 19 of valves on Y kind of maintenance schedule, et 20 cetera? It's must more just why should we believe 21 this?
22 And it might be worth it, if you guys 23 haven't already done it, to reach out to -- I guess 24 it's the NII, Nuclear Something Inspectorate, to just 25 have some conversations about first of all what are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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86 1 they thinking about for advanced reactors, but more 2 generally how do they feel like their non-prescriptive 3 safety case approach has been working in general.
4 MR. SEYMOUR: I appreciate that. In the 5 course of the sources that we look to the sources of 6 information we drew from and putting this paper 7 together we did not -- that was not one that we drew 8 upon specifically for this paper. There were certain 9 IAEA sources of information that we drew upon, but 10 they weren't driving I think the points that we were 11 focusing on from the U.K. So thank you. I made a 12 note of that so that we can pursue that.
13 MEMBER PETTI: This is Dave. I had a 14 question in terms of the second bullet, this idea of 15 no human presence. Do you mean no human presence 16 physically at the plant? What about -- I know there 17 are designs out there where they're microreactors but 18 they're remote-controlled. The control room is 19 remote. Let's call it some central location where 20 there's maybe more than one of these microreactors so 21 that there's still an operator to interact, but 22 they're not physically at the facility. Is that 23 different than what you're talking about here, sort of 24 a halfway inbetween option or --
25 MR. SEYMOUR: So it brings up a very NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 interesting point, and that is I think the 2 differentiation that we make between talking about an 3 autonomous facility and a remotely-operated facility.
4 There is a separate project that's ongoing in the 5 Office of Research. There's a working group that's 6 digging further into some of the considerations 7 associated with remote operations. So that is a topic 8 that we specifically kind of sidestepped within this 9 current paper because there's a separate effort that's 10 looking into that.
11 The context for what we're talking about 12 here -- and again we used the human. I think we were 13 trying to talk about things a bit more generically 14 than just kind of framing an operator, right? We were 15 envisioning that we may have advanced reactor 16 facilities where people are filling several different 17 roles. You may have someone who's there who's 18 fulfilling a radiation protection role, a maintenance 19 role and perhaps doing some operational checks and so 20 forth. So again that's I think the genesis of using 21 the term human there was just to put a more generic 22 kind of hat on that.
23 But what we're really talking about here 24 I think is the autonomous facility, again kind of at 25 the hypothetical extreme where you have an autonomous NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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88 1 reactor that's located somewhere, perhaps a 2 microreactor that as an autonomous control system.
3 But for the purposes of operation there isn't going to 4 be someone who's necessarily in a control facility 5 elsewhere that has the remote ability to control. And 6 I think what's different there is perhaps there could 7 be some remote monitoring, right, but what we're 8 talking about here is independent of that remote-9 control function.
10 MEMBER HALNON: Yes, I've got just one 11 other last comment on this.
12 Dave, were you done with your line of 13 questioning?
14 MEMBER PETTI: Yes.
15 MEMBER HALNON: Okay. My suggestion on 16 this is that given the interest that we have and the, 17 I guess for lack of a better term, ambiguity that 18 surrounds it this may be more of a focused effort to 19 get it defined now and understand where it might lead 20 us in the future with designs.
21 What I would not like to see is us 22 eventually get to a point where there is no 23 circumstance from a public confidence or a regulatory 24 structure perspective that we would allow a reactor to 25 be critical, change itself, its own power levels or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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89 1 whatever that goes with the autonomous word. And if 2 we come to a philosophy that there's absolutely no 3 condition that we would accept not having a human 4 person looking at it, then there would be quite a bit 5 of money wasted in trying to develop designs that 6 could possibly meet that high bar.
7 So if in fact we're going to allow this, 8 then that's one thing, but if in the end point if we 9 decide that there's just not -- it's just not 10 palatable for us to have no human presence, we should 11 state that so that we're not trying to design plants 12 unnecessarily complex and/or able to make the 13 autonomous design.
14 So anyway, I guess my point is is that I'd 15 hate to see a bar that people might strive for and 16 never be palatable to us as a regulatory structure or 17 public confidence. So again back to my first point is 18 it might be better to get this earlier and understand 19 where our philosophy as opposed to later and allow 20 something that may never happen.
21 MR. GREEN: Jesse, do you mind if I 22 respond to this one?
23 MR. SEYMOUR: Sure, Brian. One thing I 24 just wanted to interject real quick though before 25 moving on is that -- just to highlight the points that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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90 1 -- and we'll talk a bit more about this on later 2 slides, but our current stance is that we see this as 3 being something that's achievable. It just has 4 obviously a burden of proof associated with it.
5 But another thing that we want to do in 6 the course of mapping this out and one of the drivers 7 for having prescriptive criteria for this particular 8 point is because, for lack of a better way to put it, 9 you wouldn't see it being an expensive point, right?
10 There should be some regulatory certainty 11 associated with. If you design your facility in such 12 a way that you meet kind of hard requirements and not 13 have licensed operators and now you're going to forego 14 laying the groundwork and making the investment and 15 developing the licensed operator training programs and 16 all the attendant infrastructure associated with that, 17 that's an expensive change late in the game if you 18 were to find out that, well, you didn't quite hit the 19 bar.
20 So that's an area too where in terms of 21 having a prescriptive set of criteria at some point we 22 -- and there are other considerations there too, but 23 that was one of the things that influenced our 24 thinking as that's going to be kind of an early-on 25 commitment. You're pursuing those training programs NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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91 1 and developing that or you're not. And it's not 2 undoable to change that late in the game, but it's 3 expensive, right? So I think there was a sensitivity 4 to providing a measure of regulatory certainty, and 5 that's not the end-all-be-all for being prescriptive 6 on that area.
7 But again I would just highlight that it 8 is something that we see as being achievable. But 9 again it's a delicate point, so we want to be careful 10 about it.
11 MEMBER HALNON: Yes, and my point, it's 12 achievable versus allowed. I can honestly see some 13 very high level of public pushback from some of the 14 groups that --
15 MEMBER KIRCHNER: Yes, Greg, I concur.
16 The idea of having no oversight by a human on an 17 autonomous reactor, I think that's even a higher bar 18 than just having an autonomous reactor. I think the 19 public acceptance of that will be difficult. It 20 really becomes a policy issue for the Commission in my 21 opinion just like wide-scale deployment of 22 microreactors has proliferation and security issues 23 that will also rise to a policy, if not congressional 24 level.
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92 1 Factors Team leader. We recognized early on that this 2 was going to have that sort of impact. So this was a 3 topic that was addressed in an information SECY that 4 was sent to the Commission; I believe it was last 5 October or November, just to put it on the radar. To 6 my knowledge we have not heard anything back from the 7 Commission but we wanted to at least lay the 8 groundwork for this conversation at that level.
9 CHAIR BLEY: Thanks, Brian. This is 10 Dennis Bley. Hey, I don't think we have that SECY.
11 Derek, could you make sure we have it?
12 MR. WIDMAYER: Will do.
13 MEMBER KIRCHNER: Brian, can you give us 14 the SECY number?
15 MR. GREEN: I don't know it off the top of 16 my head. I'll look it up and get it to Derek.
17 CHAIR BLEY: Thank you.
18 MR. SEYMOUR: Okay. So are we okay to 19 move on or did we want to talk further on this point?
20 (No audible response.)
21 MR. SEYMOUR: Okay. So if we could just 22 move onto the next slide?
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93 1 are controlled externally from a grid control center.
2 Separately the rule should also require that 3 sufficient information be submitted to facilitate 4 reviews as outlined within the goals discussed. And 5 the examples of such information could include the 6 following: And this isn't meant to be an all-7 inclusive list, but again with the items that we're 8 talking about there are certain information needs that 9 would be associated with that.
10 I think a good way to -- it may seem like 11 a lot of information, but something I want to 12 highlight too is that there can be value in what's 13 referred to somewhat as a data-driven approach, and 14 that is this, right? In some cases providing a little 15 extra information on the front end on an applicant's 16 part can make the review process a lot more 17 straightforward on the tail end by providing that 18 needed information.
19 And something too I want to highlight is, 20 particularly with the first item, the concept of 21 operations from the design. When we talk about 22 concept of operations, the -- and we'll cover this 23 more later on, but the picture that we as technical 24 reviewers pull up in our minds is a large light water 25 one for the most part when we are doing technical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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94 1 reviews.
2 And so when these advanced reactor 3 technologies have technologies and perhaps staffing 4 models and other considerations that are kind of 5 radical departures from that, then that can create the 6 opportunity for confusion and many rounds of 7 verification from the part of the staff. So again, 8 providing a concept of operations on the front end, 9 while not currently required, to be submitted as part 10 of an application is something that we see to be 11 perhaps a tool going forward.
12 Another area would be a functional 13 requirements analysis that describes the features, 14 systems and human actions relied upon for safety. And 15 again to clarify what we mean there, earlier I spoke 16 to safety functions, right, that plants have. For a 17 large light water reactor there's a certain set of 18 those that we've gotten accustomed to. For a 19 pressurized water reactor we're used to things like 20 reactivity control, core cooling, heat sink and so 21 forth. However, an advanced reactor design may not 22 have -- by virtue of their design may not need to 23 necessarily have the exact same set of safety 24 functions and they may not necessarily be satisfied 25 using the same names.
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95 1 So things like a functional requirements 2 analysis and its companion document a functional 3 allocation can be helpful by mapping out where the 4 designer sees those safety functions to be and also 5 what's going to meet them, right? So is it going to 6 be a passive safety feature? Is it going to be human 7 action or an active feature, in which case perhaps 8 we're going to need a provision for human defense-in-9 depth that we might not need if something was more of 10 an inherent nature per se? But nevertheless, we would 11 still consider defense-in-depth. We might just 12 consider it in a little different context there.
13 So next a staffing plan, right, with 14 supporting human factors engineering-based analyses.
15 Again, if we move to stepping away from taking a 16 prescriptive approach to staffing, then we're still 17 going to need the relevant analyses to allow us to 18 determine if the case has been made for again a 19 staffing number. And that's a process that we've 20 worked through in space with some of the other 21 application work that we've done recently.
22 Again, there's a fairly robust process 23 that's established for doing that. It's just done in 24 a different context right now. We look at it as using 25 a prescriptive staffing model as a starting point and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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96 1 then doing exemptions from there. So again a 2 variation on that.
3 And lastly, a systematic approach to 4 training-based training program for relevant 5 personnel. And again that is an area that we talked 6 about. It's actually driven -- having that training 7 regulation there is actually driven by statute. SAT 8 is a tool that we've traditionally relied upon to get 9 there.
10 So what I'd like to do now is to 11 transition away from talking about rule objectives and 12 to start talking about some of the potential 13 approaches we might use in implementing some of these, 14 unless there's any questions or discussion points on 15 these objectives before we transition.
16 (No audible response.)
17 MR. SEYMOUR: Okay. So if I could, let's 18 move onto the next slide?
19 Okay. So the first thing I'd like to 20 begin talking about is scalable human factors 21 engineering reviews. Now we've discussed objectives 22 to the Part 53 rule, but as I mentioned let's turn to 23 the final segment of this presentation in which we'll 24 start discussing some potential solutions that might 25 be employed to put those objectives into action.
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97 1 So as we go through these please keep in 2 mind that some aspects of what we're talking about 3 would potentially end up being incorporated into rule 4 language while other aspects would be developed into 5 regulatory guidance documents. So for simplicity 6 we'll just focus on the main ideas here.
7 To begin with we'll start by talking about 8 scalable human factors engineering reviews. So we've 9 initiated work under contract with Brookhaven National 10 Laboratory, and specifically we've been working with 11 Mr. John O'Hara there, to develop a method for scaling 12 the scope and depth of human factors engineering 13 reviews for advanced reactors.
14 The objective of this effort is to enable 15 the staff to readily adjust the focus and level of 16 staff human factors engineering review efforts based 17 upon factors such as risk insights and the unique 18 characteristics of the design or facility operation.
19 In the interim though -- and again, as we talk about 20 these topics today it's important to keep in mind that 21 we anticipate that we may receive advanced reactor 22 applications before this rule is out the door. So 23 something we have a mind for is what do we do in the 24 interim?
25 And what we have in the present is we have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 1 the ability to adjust the scope of our normal NUREG-2 0711 human factors engineering reviews on a case-by-3 case basis should a given license application warrant 4 a reduction in the scope of a human factors 5 engineering technical review. Again, so that's a tool 6 that we have now, but we're looking towards a more 7 comprehensive scalable process here.
8 Okay. Next slide, please? So now moving 9 into the next part, and we've touched upon this 10 somewhat to begin with. This will actually be over 11 the next few slides here talking about this.
12 So let's discuss the potential criteria 13 that could justify not using licensed operators in an 14 advanced reactor facility. To justify not using 15 licensed operators it must be demonstrated that 16 adequate protection of the public health and safety 17 will exist in the absence of any operator action for 18 preventing or mitigating accidents.
19 The following are examples of criteria 20 that could potentially be used for assessing the 21 acceptability of an advanced reactor design operating 22 without using any licensed operators: And again, this 23 will be over a few slides here, so we'll start by 24 talking about the first one here.
25 So one, the accident analysis for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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99 1 design should demonstrate that the radiological 2 consequence criteria will be met without reliance on 3 human actions for event mitigation, defense-in-depth 4 or safe shutdown. So that's the first of several 5 criteria that we'll talk through.
6 Next slide, please? Two; so moving onto 7 the second criteria, safety of the design should rely 8 upon inherent safety characteristics. And again, I 9 maintain the awareness of our earlier discussion on 10 that particular term, but again presenting this at the 11 high level that we are right now the second objective 12 is that safety of the design should rely upon inherent 13 safety characteristics. Absent an operator presence; 14 and this is something I alluded to before, the 15 absolute reliability -- again keeping in mind our 16 discussion on the need to perform some more work on 17 that particular term, that the absolute reliability of 18 inherent safety characteristics would be key.
19 And the point we're trying to make there 20 is this, right? If the opportunity for human 21 intervention doesn't exist, then the burden placed 22 upon the full layered approach to various safety 23 features is elevated, right, because you lose that 24 human intervention backup. And that's the key thought 25 that's there.
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100 1 So again, we talk about inherent safety 2 characteristics, we talk about passive safety features 3 and the various merits of those, but the key point at 4 the end of the day is just that you're definitely 5 placing a different type of focus on those safety 6 systems because you've lost your ability for human 7 intervention if they don't do that job.
8 So if you have an active safety feature 9 that fails to work, the manual human insertion of that 10 protective action is no longer available. So now what 11 do you fall back on, right? And that's where looking 12 at things from a safety function perspective -- and I 13 mentioned a functional requirements analysis and a 14 function allocation before. That's where we would 15 need to be able to trace that through and find that 16 the final -- in the end that the safety function is 17 being satisfied by something that's other than a 18 person, right?
19 So the automatic protection system has 20 failed. So what's protecting you now? Is it a 21 diverse system that kicks in? Is it something that's 22 passive? Is there going to be some inherent 23 characteristic of the design that's going to be a 24 limit hazard posed and maintain things in some kind of 25 a safe state, right? So again, I'm just bringing up NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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101 1 that particular objective at a high level. There are 2 a lot of layers to it, but that is the second aspect.
3 So three. So this would be moving onto 4 the third item. If not fully autonomous, right, then 5 the design should have sufficient autonomy to support 6 safety without human action. So at a minimum you 7 would need sufficient autonomy to support whatever is 8 needed for safety without human action.
9 If human action is needed for a startup, 10 then it may be appropriate to -- and there are some of 11 the points here. And what this brings -- the point 12 that this brings up is that a facility that runs 13 autonomously may not necessarily be able to start 14 itself up absent having a person start it up and 15 initially get it on line. And even if that startup is 16 relatively straightforward, perhaps just enabling a 17 permissive -- it allows a computer to check parameters 18 and then perform the startup in an automated way. The 19 initial startup of a facility may involve things like 20 physics testing and so forth to where people will need 21 to actually do that.
22 So if a human is needed to actually start 23 up the facility, then we suggest two potential options 24 for doing that: One would be the most 25 straightforward, which would be to have a licensed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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102 1 operator conduct the reactor startup. The other 2 alternative could potentially be to allow the 3 demonstration that safety analyses bound all 4 postulated errors that a non-licensed operator might 5 commit during a reactor startup. And we feel that the 6 higher burden there is warranted because in our eyes 7 a non-licensed startup operator's abilities would not 8 provide us as the NRC staff with the same degree of 9 assurance as those of a licensed operator, right?
10 And again, we have what we feel to be a 11 very comprehensive framework for operator licensing 12 and that provides us with the ability in part to 13 independently examine and certify operators. So we 14 get that ability to come in and to make an independent 15 assessment of their knowledge, skills and abilities.
16 So in the absence of that we would see there as being 17 a higher bar that would need to be met to let someone 18 without a license start up a reactor, irrespective of 19 how small it might be.
20 MEMBER HALNON: Why would we even 21 entertain that?
22 MR. SEYMOUR: So we have had some 23 application discussions, without getting too specific, 24 that have broached this possibility. So one of the 25 things that we've done in the course of this, driving NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 1 this paper is obviously that provides us with a data 2 point. So we've tried to envision (A) would that be 3 acceptable under any hypothetical situation? And if 4 it is, then what would the bar be for that?
5 And again, at the end of the day this 6 paper is meant to be a vehicle for having these very 7 discussions and weighing the merits of this, and this 8 is the set of options that we came up with, that 9 either (A) you don't it at all; you have a licensed 10 operator do that. Or if you do, then what do you have 11 to account for, right? Well, you have to account for 12 the fact that there is a broader range of uncertainty 13 about what this person might do wrong in the course of 14 starting up a reactor.
15 Now the design -- and I'm speaking -- just 16 coming up with a hypothetical here. If it can be 17 demonstrated that the reactor design is sufficiently 18 robust that any conceivable excursion is limited and 19 nothing would get damaged, nothing would get 20 challenged, nothing would get released by the 21 incorrect actions being taken by an operator, then in 22 our eyes it's at least conceivable that perhaps there 23 could be a means of finding that to be acceptable.
24 But again, this is an interesting point, 25 right? Since the 1950s, as I alluded to earlier, we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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104 1 have defined what controls are. And the Atomic Energy 2 Act gives us the latitude to be able to write 3 regulations provided that we do it in a uniform way 4 that mandate operator licensing. And what we've done 5 is we've defined those controls as being things that 6 affect reactivity and power. And then we've said that 7 you either have to have a license to do that or, in 8 the rare exception that you're perhaps a trainee, you 9 need to be under the direct oversight of someone with 10 a license. So this is somewhere that we haven't gone 11 before. And again, this is a point that would 12 represent a departure.
13 I think the intent of what we were really 14 trying to do here was to say if we were going to allow 15 such a thing, where would that bar at a minimum lie?
16 And again, our perspective at the present point is it 17 would be probably at that level. You almost have to 18 establish that this operator can't damage anything by 19 making those errors because we just want to have that 20 level of assurance about their abilities.
21 MEMBER HALNON: Okay. No, I understand 22 keeping an open mind and looking at everything from a 23 balanced view. I guess we want to make sure that we 24 fully understand, and I'm sure you all do, the reasons 25 for licensing is not so much -- well, to me it's three NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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105 1 facets. It's probably more, but it's capability, it's 2 your assurance as the regulator that capability and 3 proficiency is there, but it's also a public 4 confidence issue as well.
5 So there are a lot of different things 6 other than just technical capability of the design to 7 be able to do it. So we have to continue to factor in 8 those other maybe non-specific but tangential issues 9 that we have to make sure stay in the conversation.
10 It' snot just the technical capability. Can the plant 11 do it? It's other things, too.
12 MR. SEYMOUR: Yes, and if I could go 13 further what I would say is that -- and again, I come 14 from -- I mean I was a licensed operator earlier in my 15 career and I've been involved in operator licensing 16 for some years now actually conducting the exams.
17 It's an area that I personally feel very strongly 18 about and the merits of it.
19 And I would just like to say that the 20 current stance that we take within the operator 21 licensing branch on this particular topic is that --
22 because we anticipate that there could be some 23 questions raised about, quote/unquote, if these 24 reactors are very safe, why require people with 25 licenses to operate them? Why not have a more open NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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106 1 mind, using non-licensed operations within certain 2 traditional roles?
3 And what I'd like to say to that, if I 4 may, is that using the licensed operator framework 5 allows us to independently certify the knowledge, 6 skills and abilities of applicants, and that provides 7 an additional safeguard against an inadequate training 8 program leaving facility personnel unable to fulfill 9 safety roles.
10 Another thing that it does is it -- that 11 licensing comes with requirements for -- and I'm just 12 going to outline a few of them. A licensed person 13 comes with requirements for medical examinations, 14 right? So they have to be routinely given a medical 15 exam to make sure they're fit to carry out their 16 duties. And that's beyond fitness for duty 17 requirements, right? This is making sure that things 18 like your blood pressure -- you're not at risk for 19 being incapacitated due to perhaps a stroke or 20 something like that. I use that as an example, but 21 it's very comprehensive.
22 There are requirements for continuing 23 training and reexaminations. There are requirements 24 for maintaining proficiency, and there's also -- and 25 lastly this person is independently licensed under NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 Part 55, and that's separate from the facility 2 licensing. So there's the potential for individual 3 accountability for violation of our regulations.
4 And when you look at all that stuff in 5 aggregate, that provides a degree of assurance that, 6 to be quite frank, a non-licensed operator simply 7 cannot in some cases.
8 So we feel that a well-designed operator 9 licensing process can be efficiently applied to 10 advanced reactors and in some cases it may be prudent 11 to scale that program. So perhaps that exam isn't 12 going to be the same size or perhaps tested in the 13 same ways as what we do now. Perhaps the training 14 program that leads up to it might be shorter because 15 there's simply less to cover, but at the end of the 16 day that license still comes with those factors. And 17 those are things that we feel strongly about going 18 through this project.
19 MEMBER PETTI: I just had a question. I 20 worry about the public perception here. The lady that 21 cuts my hair needs a license, but we're going to let 22 someone who isn't licensed start up a reactor? The 23 autonomous control, everything that -- on the news 24 we're hearing about crashes from Tesla's -- the 25 autonomous control is not there. I mean this is still NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 1 a boundary that is not as well proven I think and 2 let's call it less-risky technologies than the nuclear 3 here. Similarly we wouldn't have -- we wouldn't allow 4 someone that wasn't a licensed pilot to sit in the 5 cockpit and take over. So I think there's some other 6 analogies outside of nuclear that we need to be aware 7 of. Thanks.
8 MR. SEYMOUR: Yes, and again to that point 9 I would just reinforce that again this is a 10 presentation of -- in this particular aspect -- and 11 I'll again -- I should be more specific and say that 12 items 3 Alpha and Bravo here are really an 13 articulation of when we distilled down the various 14 approaches I think what it kind of narrowed down to.
15 We have a long history with obviously 3 Alpha, having 16 a licensed operator conduct the reactor startup. And 17 3 Bravo presents challenges. It does. And one of 18 that we discussed internally with that are issues of 19 public perception and confidence and so forth.
20 So we present that to get the feedback, 21 exactly what we're getting in this discussion. I mean 22 that's one of the objectives here. And I would simply 23 kind of back around to this is a work in progress. So 24 we certainly haven't set our minds to permitting that, 25 allowing a non-licensed operator to fulfill that role NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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109 1 by any means.
2 MEMBER BROWN: Does that mean that once 3 you've started up the non-licensed operator continues 4 to be the non-licensed operator for all operations as 5 well? You've used the words for startup. I always --
6 once the plant's started up, what do you mean, he can 7 leave and go home and it just operates totally 8 autonomously? Is that the point of this whole 9 discussion, or this whole thought process?
10 MR. SEYMOUR: So when we were discussing 11 this particular item, 3 Bravo, obviously we had to 12 have some kind of a mental construct that we were 13 considering that within. And the framework, the 14 backdrop if you will, was that you had an autonomous 15 -- a reactor that could operate autonomously but still 16 needed someone to do something like do the physics 17 testing on the initial startup or perhaps pull the 18 rods to get it online.
19 But then again I used the point of 20 comparison and I was -- the technical specifications 21 that we have large light waters. The required 22 staffing for a facility that's required by technical 23 specifications can vary with mode, right? So once you 24 get out of a startup mode -- and again, this was just 25 kind of the backdrop that we used to think through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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110 1 this. You could be in a situation where once you were 2 out of that startup mode and the plant has ascended to 3 some given power level where it's going to continue on 4 autonomously. Then you're going to be outside of the 5 mode of applicability or that particular staffing 6 requirement, what would be applicable.
7 And again, inherent in that discussion of 8 the autonomous reactor is the fact that, at least 9 theoretically, it's able to run itself, right? It's 10 able to move through power ascension and maintain 11 itself online and so forth without necessarily having 12 a human operator there.
13 Now what that doesn't mean is that you 14 might not have non-licensed operators fulfilling other 15 roles besides things like reactivity control and so 16 forth, and large light waters are an area where we 17 currently have that. You'll have non-licensed 18 operators that are there to operate secondary systems, 19 doing rounds, things like that, but those non-licensed 20 operators aren't the people that are going into the 21 control room and starting and stopping safety systems 22 and pulling rods and things of that nature. So again 23 that's another distinction that I would just make.
24 MEMBER KIRCHNER: Jesse, going back to the 25 existing regulations, and hopefully I'm not misquoting NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 1 the numbers, 54m, 50.54m, whenever you have fuel 2 loaded you have to have an SRO on site. Roughly 3 that's -- I'm doing shorthand here. It's more 4 complicated than that. Why would not have an 5 equivalent level of oversight, whether it's an 6 autonomous reactor or a partially autonomous reactor?
7 And whether there's an SRO or an RO; we could debate 8 that, the level of training maybe scales to whether 9 it's a test size reactor or a several 100-megawatt 10 electric reactor.
11 But it would seem to me, going back to 12 some of Greg's points, that from a public confidence 13 and from a policy standpoint, boy, are we leading 14 people down to -- or are we leading the staff into a 15 situation where you get into a position with an 16 applicant that is just untenable and perhaps 17 unresolvable?
18 This is where I worry when we're scaling 19 and we're being so flexible that we are not 20 establishing any criteria, so to speak. And that's an 21 exaggeration on my part. Then because you're 22 obviously going to review each of these applications, 23 but then we interject a lot of uncertainty, regulatory 24 uncertainty into the process as well. Expectations 25 from both parties perhaps will not be matched without NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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112 1 some specificity.
2 MR. SEYMOUR: That's definitely something 3 that we've been sensitive to, working through this.
4 The public perception aspect of that, the public 5 confidence, is something that has weighed heavily upon 6 us. And again, that's why one of the things I made 7 sure I pointed out earlier was we've been licensing 8 operators for a very long time.
9 And if you think back to the timeframe of 10 that requirement, I mean, you're not too far removed 11 from kind of the birth of the Atomic Energy 12 Commission, and so forth. And so, at a very early 13 point in time, I believe, if I remember the genesis 14 correctly, it came from Congress, the initial push.
15 I think there may have been some hesitancy on the 16 Commission's part at that point to necessarily mandate 17 licensing, if I remember the background correctly.
18 But, again, that came from a public type of push there 19 to have that, so this person would have some form of 20 licensing. So, again, it's a very important point.
21 One of the ways in which we have thought 22 this is that, inherently, people can be sources of 23 desirable things and undesirable things when they 24 interact with a system. So, people can take 25 corrective actions. In the best case, right, a person NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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113 1 can realize that a system is not doing what it's 2 supposed to do, and they can intervene and start 3 taking manual actions that ensure that the correct 4 things happen.
5 But the opposite number holds true as 6 well. And that is that people can be sources of 7 errors of commission, right, where they go through and 8 perhaps they inject things that are detrimental.
9 So, one of the thought exercises that we 10 went through was we said, if you had a hypothetical 11 reactor that in the accident analysis doesn't require 12 any human interventions to show safety, if this 13 reactor has very low radiological consequences that 14 are acceptably low associated with that, and if it's 15 automated, and if there is adequate defense-in-depth 16 to where, if the means that's relied upon for safety 17 fails, or something (audio interference), that we'll 18 go ahead and still provide for safety.
19 We said, if you met that, at that point, 20 what would the person be adding within that context, 21 right? Now, obviously, there is that assurance, 22 right, that assurance that, hey, there's another way 23 layer that's there; there's still a person. And that 24 holds true.
25 But we had to also consider at that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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114 1 point that there wasn't a whole lot more, in terms of 2 the traditional way that we kind of think through 3 safety cases, there wasn't a whole lot more to be 4 added by the person, and there were potentially 5 negative things to be added by the person being there.
6 So, at that point, it became kind of a breakeven, 7 right, because the person could still be the source of 8 errors of commission, and so forth.
9 Again -- and I think this gets back to the 10 earlier point -- what we are trying to do here is 11 we're trying to say, is this something that could be 12 acceptable under the right conditions? And what would 13 those conditions be? I think there's still a broader 14 question at a policy level, to be quite honest. It 15 could be debated at what level of policy that is. The 16 Atomic Energy Act, our understanding of it is that it 17 gives us the ability to mandate licensing via 18 regulation, but it doesn't necessarily say that we 19 always have to, right? So, that's kind of left to the 20 Commission's judgment. But I think there is a policy 21 matter, and there could be a consideration there for 22 the public perception that will have to be evaluated 23 at that time.
24 I think the role that we're trying to fill 25 right now is by laying out the considerations to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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115 1 refine the best possible information, should that 2 situation be encountered where a policy decision needs 3 to be made higher up.
4 MEMBER BROWN: It's interesting, you go 5 through this -- I hate to be the heretic and throw 6 myself into this quagmire. I mean, this reminds me --
7 and I've got to go back, I don't know, it must be 30 8 years-plus -- of the argument that all autonomous 9 operations, the keying in of information, the software 10 is so reliable; it's so wonderful. No mistakes will 11 ever get made and we can sit back and, 5,000 miles 12 away, key in the parameters for what we want to do, 13 and away it goes and it starts up.
14 It reminds me of the old Therac radiation 15 treatment machines where there's an automated setup.
16 The operator keyed in the information, and then, they 17 ended up killing a bunch of people because the 18 operator keyed it in so fast that it bypassed the 19 safety settings, and they ended up overexposing people 20 to huge quantities of radiation and killing them.
21 And with this thought process that human 22 interaction is not required, or local human action is 23 not required, or some mechanical setup not depending 24 upon software, is just insane. I hate to phrase it 25 that way. Quite frankly, if I was on this Committee, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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116 1 this whole consideration would have been X'ed out.
2 I'm, obviously, not, but that's just my thought 3 process on the thing. If I was a public person 4 commenting on this, there would be holy heck being 5 raised.
6 I don't think you can be so expansive and 7 so thoughtful that we really have to consider stuff 8 that doesn't make sense. There is no answer to that.
9 So, I will just close my mic.
10 Many people probably don't remember the 11 Therac circumstances. Dennis probably does, or maybe 12 you --
13 CHAIR BLEY: There's a lot of reports on 14 that that's worth looking at. Of course, that did get 15 fixed.
16 MEMBER BROWN: Oh, of course it did.
17 CHAIR BLEY: There's a book identifying 18 cases -- it's a few years old -- called "Set Phasers 19 on Stun". It's worth looking at if you're considering 20 this stuff. It includes one national defense problem 21 that almost led to a launch by error. And it's worth 22 reading, and then, you can find the real technical 23 information, in fact, set up. Therac is that one, 24 too. When you read it there, you scream when you get 25 to the point of the diagnosis.
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117 1 And it didn't work out the way Charlie 2 said, but, yes, basically, it was in the software. It 3 got fixed, but it killed some people on the way.
4 So, I think this requires a lot of care 5 and some way to protect against the unforeseen. And 6 the Boeing example is another one. People could have 7 protected against it, and they decided not to.
8 Anyway, I don't disagree totally 9 completely, but, also, the other side of this is, if 10 you're dealing with something with demonstrably very 11 low consequences, whether they come from fission 12 products or something else, maybe it's not 13 unreasonable, but it requires enormous care.
14 MEMBER BROWN: Even though we still have 15 barbers that have to be licensed just cutting your 16 hair.
17 CHAIR BLEY: That could be a moneymaker.
18 MEMBER BROWN: Yes.
19 MR. SEYMOUR: Yes, I do appreciate the 20 pointer to the reference, and I think we do have maybe 21 a copy of that floating around in our branch, 22 actually. So, I appreciate that. I will definitely 23 try to take a look at that.
24 MEMBER KIRCHNER: Jesse, this is Walt 25 Kirchner.
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118 1 MR. SEYMOUR: Yes?
2 MEMBER KIRCHNER: Just on this topic, is 3 there some minimum bar currently that the NRC imposes 4 on research and test reactors in terms of licensing of 5 the operators?
6 MR. SEYMOUR: Yes. So, in the paper we 7 try to summarize that. And again, I'll qualify my 8 remarks by saying that my operator licensing 9 background came from the power reactor side. But, in 10 general, yes.
11 So, the staffing requirements are a little 12 bit different, in that, in general, research and test 13 reactor staffing is going to mandate that you have one 14 licensed operator at the controls, and it's going to 15 mandate that you have an SRO that's at least on call.
16 And so, again, a little bit tighter of staffing.
17 The licensing process is a bit different, 18 too. Again, those facilities generally don't have 19 simulators. So, the exam is structured a little bit 20 differently.
21 And another thing that is interesting, 22 too, is that, whereas we kind of take a one-size-fits-23 all approach on the power reactor operator licensing 24 side, because there's such a variety of the research 25 and test reactors -- if memory serves me, I think they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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119 1 range from 5 watts all the way up to megawatts, if I 2 remember right, and just a variety of complexity. We 3 actually have a graded approach to how those exams are 4 done. So, the complexity of the exam varies with the 5 type of facility being examined on.
6 So, again, there's differences in staffing 7 and there's differences in the actual licensing 8 process that's involved there. And again, the paper 9 goes into a bit more detail about what some of those 10 differences are.
11 Okay. Unless there's anything else, I'd 12 like to go ahead and just move on to the next slide.
13 And this will be a continuation of the criteria that 14 we were discussing.
15 Okay. So, fourth, license conditions 16 should be established for the facility by which those 17 administrative responsibilities with safety 18 implications -- so, for example, technical 19 specification compliance -- that would otherwise have 20 been allocated to licensed operators are reassigned.
21 And again, this is on the presumption that 22 there's no licensed operators at a facility. Those 23 important administrative tasks still need to be 24 performed by someone, right? So, for lack of a better 25 way to put it, somebody needs to be on the hook for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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120 1 ensuring that these administrative functions are still 2 completed. And we outlined examples of what some of 3 those may entail earlier. So, an example of that may 4 be, potentially via a license condition, assigning a 5 designated facility manager who now has responsibility 6 for that.
7 The last point is, again, kind of segueing 8 from strictly speaking about licensed operators, 9 although not necessarily licensed operators --
10 sometimes they are; sometimes they aren't -- for the 11 STA position, the staff would need to engage with the 12 Commission on a proposed departure from policy, should 13 an applicant propose a staffing plan that does not 14 include on-shift engineering expertise.
15 And again, this is the point that, if a 16 facility is looking to not have licensed operators, we 17 can envision that they're probably going to 18 potentially pursue not having any Shift Technical 19 Advisor as well. But, as discussed, the underlying 20 ties associated with the STA position are different.
21 So, we have to be considerate of the differences that 22 would need to be addressed there.
23 So, a key consideration would likely be 24 the applicant's ability to demonstrate that the result 25 of staffing-related analyses remain adequate, in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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121 1 absence of the on-shift expertise provided by an STA.
2 And again, that's an example of a key consideration 3 that may factor into that. But, again, that would 4 require, our understanding, our current interpretation 5 is that there would have to be a Commission 6 interaction to proceed forward with that type of an 7 exception.
8 MEMBER KIRCHNER: I'm going back to bullet 9 No. 4, Jesse. This is still Walt Kirchner.
10 Someone's on the public line with their 11 mic not muted.
12 No. 4 almost begs the question of whether 13 or not this facility manager position would have some 14 technical expertise, either like a licensed operator 15 or like an STA. Because safety implications, as 16 governed by tech spec compliance, really would require 17 someone not to be an administrator, but a technically 18 competent person in the organization.
19 MR. SEYMOUR: Yes, and this is a point 20 that we've talked about specifically in the paper.
21 It's a very interesting point. And what I'd like to 22 say is that, even if this individual wasn't licensed, 23 it doesn't mean that there wouldn't also be a very 24 good case to be made for prescribing certain training 25 requirements in that case, right? And there actually NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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122 1 is a statutory tie that's there.
2 So, the relevant portion of the Nuclear 3 Waste Policy Act, essentially, says that we will 4 establish regulatory guidance for the training and 5 qualifications of civilian nuclear plant operators, 6 supervisors, technicians, and other appropriate 7 operating personnel. So, by statute, we have the 8 mandate to establish training where it's -- you know, 9 there's now training requirements via regulation where 10 it's relevant for those categories of personnel. And 11 those categories in the actual statute are high level.
12 In the training rule, which is 50.120, we actually 13 drill down and we make finer distinctions on exactly 14 who has to have that training.
15 But, ostensibly, Part 53 will have to also 16 comply with the Nuclear Waste Policy Act, right, and 17 those training provisions that I was talking about.
18 So, what we would have to do is we would have to 19 establish a regulation that covered the relevant 20 training.
21 And the reason why I bring that up is 22 that's something that we have to do as part of this 23 rulemaking. We will have to determine what categories 24 of personnel require training.
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123 1 right? And I'm speaking, again, hypothetically here.
2 But we would be, at least from my understanding of the 3 requirements, we would be within our rights to have a 4 regulation, as part of Part 53's corollary to the Part 5 50 training rule, that would mandate that such a 6 facility supervisor position would have to have a SAT-7 based training program that would be of a sufficient 8 scope to ensure that they could fulfill those 9 responsibilities.
10 So, while this individual may not be 11 licensed, I would say almost invariably it would have 12 to be some sort of a regulatory tie that would be 13 there. And most likely, there would be, to where 14 there would actually be a regulatory hook, if you 15 will, to ensure that they are being trained.
16 MEMBER KIRCHNER: I just see you bending 17 over backwards so far and so hard to be so flexible, 18 when you have existing regulations that have served 19 you so well. And you, earlier this morning, gave such 20 a compelling argument for the purposes of having 21 trained operators.
22 So, I can go along with the autonomous.
23 Now this is one person's, one member's opinion, so let 24 me qualify this. But, rhetorically, I can see the 25 autonomous operation. I can see the problems that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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124 1 Charlie is worried about. I can see the safeguards 2 that you could insist on, like independent reactor 3 protection system to shut it down, et cetera, et 4 cetera. But, somewhere in this loop of things, 5 someone who is qualified -- i.e., at least at an RO 6 level, but probably, if it's a larger plant, at an SRO 7 level -- would, I think, be necessary to not just 8 address public confidence, but someone somewhere has 9 to have oversight of things like you've itemized 10 previously. Like in No. 4, are you really in tech 11 spec compliance?
12 So, given everything you have to do in 13 10 CFR 53, why create a new training program for 14 someone that's, quote-unquote, "a substitute for a 15 licensed operator"? Just one member's opinion.
16 MR. SEYMOUR: To kind of get back to a 17 point that I was talking to earlier, this is, for lack 18 of a better word, this is an issue that we're still 19 working through internally in our discussions, and so 20 forth. It's a complex matter that we're trying to 21 navigate.
22 And I would go back to the points that I 23 was articulating earlier about the value of having a 24 licensed operator to fill certain roles, right, and 25 what we get with having that licensed operator there.
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125 1 And what I would say, a key consideration 2 here is this, right, and this is a point that lends 3 itself naturally to debate, right? If a person is 4 trained adequately, we would expect that they could 5 fulfill a given role, right? One of the many things 6 that a licensed operator training program does is it 7 allows us to independently assess that that training 8 has been successful, right, and that the individual 9 can demonstrate the knowledge, skills, and abilities 10 to fulfill a given role, right?
11 And what happens is, if we approach 12 something in this way, right, we can go through and we 13 can start mandating certain requirements on a case-by-14 case basis that perhaps start building back up towards 15 the full scope of an operator licensing program. For 16 example, for this individual that has this additional 17 training to fulfill these tasks, maybe we would go 18 through and also mandate other requirements, right?
19 The problem that we're, I think, having to 20 work through is this, right? At what point do you 21 reach that break point where you start having to pack 22 on so many additional requirements for this person to 23 make sure that we are confident that they can fill 24 these roles, to where at this point why shouldn't it 25 just be operator licensing program? And what I can't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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126 1 give is I can't give a perfect answer to that because 2 it's an open question, right? It's something that 3 we're working through.
4 So, what we can do here in this paper --
5 and again, what I'm, I guess, trying to accomplish for 6 the presentation -- is to lay out the considerations, 7 but not in any way to say that these are matters that 8 are resolved or that we've necessarily reached the 9 perfect conclusions on each of these items. Because, 10 again, part of what we're doing here is we're trying 11 to map out the background and the considerations, and 12 then, open up exactly these types of discussions and 13 to figure out if there's a better way to go about 14 doing it.
15 And again, this individual, this is a good 16 example here. Do we have examples out there where we 17 have people that are fulfilling roles where we ascribe 18 some other requirements, and yet, it's not necessarily 19 a licensed position? Yes, we do have corollaries to 20 that. One example would be the Certified Fuel Handler 21 Program, right? And again, that's not exactly a one-22 for-one correlation here, but there are places like 23 that.
24 But, again, what we're going to run into 25 is we're going to run into kind of a philosophical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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127 1 transition point where we're going to have to say, is 2 it more effective to go through great pains to try to 3 make sure that non-licensed people can fulfill certain 4 roles, or is it going to be better to utilize a 5 licensed operator process for all those roles, right, 6 that one traditionally would, but allow those programs 7 to be tailored specifically to that design and to the 8 considerations there?
9 And again, I use those examples where, if 10 you look at it just qualitatively, right, an RTR 11 operator licensing program is generally of a shorter 12 duration. And the ongoing kind of attendant 13 requirements associated with them aren't quite as 14 extensive as what you have for a power reactor 15 operator licensing program. And by that, I mean 16 downgoing requalification, and so forth, right? The 17 requirements are still there, but they're not quite at 18 the same scope in depth.
19 And what the difference is there is that 20 in both cases you're producing an operator that you 21 have confidence in that they can perform the safety 22 role. In one case, you've got a much smaller facility 23 with different safety considerations associated with 24 that, and you're just able to achieve that with a much 25 shorter program.
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128 1 So, again, that's a philosophical bridge 2 that we won't have to cross. You know, do you go 3 through pains to try to let novice operators do more 4 or do you try to be more efficient about targeting the 5 operator licensing program, so that it's appropriate, 6 but you still maintain all the benefits that come with 7 that?
8 And that's really the best that I can do 9 at this point, is just say that those are the 10 considerations that weigh heavily on our minds right 11 now.
12 Okay. So, if everyone's okay, I'll go 13 ahead and move on to the next item. Okay.
14 And the next item is, again, talking 15 through solutions. So, what I want to talk about here 16 is a scalable approach to operator licensing 17 requirements. And again, what I spoke to earlier was 18 that we've been thinking through how do you take an 19 operator licensing program and tailor it to unique 20 technology that may be out there.
21 And so, what I want to talk about here is, 22 how could we take an approach that's flexible and 23 efficient to operator licensing? And a flexible 24 process that advanced reactor vendors and licensees 25 could use to develop an operator licensing exam NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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129 1 program for their science might consist the following 2 steps, and again, these are at a very high level:
3 But, in any case, what we foresee would 4 need to be done is everything would begin with a job 5 task analysis. So, for this unique design, you would 6 have to go through and identify the specific 7 knowledge, skills, and abilities related to the 8 facility's licensed operator role. And again, we 9 talked about how they could be different for the 10 safety functions and technology, and so forth.
11 You have to go through and you have to 12 figure out what is important for the licensed 13 operators for that particular facility. So, then, the 14 training and evaluation methods to be used would be 15 selected using a systematic approach to training 16 process. And as part of that, what you would do is 17 you would determine the exam composition.
18 And again, as I mentioned, our current 19 frameworks are out there licensing both power reactors 20 and research and test reactors. We use slightly 21 different methods to accomplish that, right? So, a 22 research and test reactor generally isn't going to use 23 a simulator to perform their license examinations; 24 whereas, a power reactor facility, the simulator is 25 considered to be the most operationally valid part of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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130 1 the whole thing, right? So, again, there's 2 differences in how we do that, and then, we achieve 3 similar results, right? We produce a licensed 4 operator that provides us with a reasonable assurance.
5 So, the next step would be that -- again, 6 because this would be kind of a tailored examination 7 process -- the vendor or licensee would then pilot the 8 proposed exam. So, this would involve going through 9 the examination to make sure that it does what it's 10 intended to do. And then, what we would as the 11 regulator is we would then review, and then, proceed 12 on to administrator those examinations.
13 So, again, in any case, what we don't 14 depart from here is, you know, retaining the final 15 licensing authority. And also, a current thing that's 16 very well established is for analysts to go through 17 and review exams prior to administration. In the case 18 of research and test reactors, they will actually 19 develop those examinations. On the power reactor 20 side, we periodically develop them. Most of the time, 21 we'll have the facility develop them. So, there's 22 differences there, but a key thing is that we're 23 always putting our eyes on that product before it's 24 administered.
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131 1 here that I just do want to touch upon is that, when 2 we benchmark internationally and when we look at how 3 some of our international counterparts officiate their 4 operator licensing and certification processes, we do 5 see that there are different ways to kind of get to 6 similar results.
7 And one of them is the example that we see 8 with Canada. In that instance, while the Canadian 9 nuclear regulatory authority there, the CNSC, while 10 they retain the certification authority, their 11 practice is actually to let the facility administer 12 the examination, right? And, of course, there's 13 reviews, and so forth, that happen on the administered 14 examination prior to licensing.
15 So, a potential option that we do put out 16 there in the paper is that, as part of this, it may be 17 appropriate, and perhaps under certain conditions, to 18 allow vendors and licensees to also administer their 19 own license examinations. So, again, that's a 20 possibility that we do draw there as well.
21 MEMBER BIER: If I can comment briefly?
22 This is Vicki Bier.
23 I don't want to argue that it would never 24 be appropriate for vendors or licensees to administer 25 their own exams, and I'm not familiar with the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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132 1 Canadian system that may be working perfectly well 2 there, but I think the experience in aviation at the 3 FAA is that allowing airline pilots to do test rides 4 for their sort of peers in their same airline has not 5 been super reliable. So, just a caution.
6 MR. SEYMOUR: I appreciate that. Thank 7 you.
8 What I will say is, we're only kind of 9 touching upon some of our ongoing work here, but we do 10 have an ongoing effort to investigate the potential 11 for this new approach to operator licensing. And 12 embedded in there -- and I'll touch upon this, I 13 believe, a little bit later on, or perhaps not; I 14 can't remember if it's on a later slide -- but one 15 thing that we're doing as part of that, we're making 16 a concerted effort to consider lessons learned and 17 things that we can take away from other industries.
18 And one of the industries that we do have 19 marked to follow up on is aviation, right? So, 20 specifically, the FAA, to take some learnings from 21 that.
22 So, yes, I do appreciate that. And what 23 I will say is that, even though we've just kind of 24 started down that path, our intention is to try to 25 learn what we can from how the FAA approaches the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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133 1 pilot certifications.
2 Okay. So, if I could, I would like to go 3 ahead and move on to the next slide.
4 So, moving on, please recall that earlier 5 I mentioned that a concept of operations may have 6 value in helping us understand a new design. And what 7 I'd like to do is just kind of talk briefly about the 8 concept of operations and why we focus on it here.
9 Because, recall earlier, what I pointed to is that 10 there may be value in us requiring a concept of 11 operations a part of these applications.
12 So, there is currently no regulation that 13 requires applicants to provide a concept of operations 14 as part of applications. And the reason why we come 15 back to this point is that new designs will likely 16 conceive of radically different concept of operations, 17 for which the staff may have little or no prior 18 understanding. Because, again, we're used to working 19 with the large light water reactors. So, some of 20 those very basic assumptions that we have in our minds 21 about how things work may be not necessarily the case.
22 So, therefore, there may be a need to 23 explicitly make the concept of operations a part of a 24 content of applications under Part 53. And the 25 description of concept of operations, we see it as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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134 1 being something that will help us to avoid confusion, 2 to understand/confirm to what extent a design relies 3 upon humans for safe operations. Again, an important 4 point. And also, to determine the appropriate scope 5 of the staff review, because, once again, we're 6 talking about pursuing scalable processes. And an 7 additional point would be to reduce the need for 8 subsequent requests for additional information. So, 9 again, this is something that we can easily see these 10 questions coming up with these designs, and we just 11 see there as being value in asking for it on the front 12 end.
13 Okay. If we could move on to the next 14 slide, please?
15 Okay. So, moving on to staffing analyses, 16 turning back to the subject of staffing, and in lieu 17 of taking a prescriptive approach to staffing under 18 Part 53, it may be appropriate for applicants to 19 propose their own alternative staffing models. At a 20 minimum, an HFE-based staffing analysis of sufficient 21 scope and depth should allow the NRC staff to 22 adequately assess the susceptibility of the proposed 23 levels would be needed.
24 Furthermore, something I'd like, I guess, 25 to go a little further into is that alternative NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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135 1 staffing models for advanced reactor applicants could 2 be informed by the existing process of NUREG-1791.
3 And what I just want to kind of emphasize there is 4 that NUREG-1791 was written for a slightly different 5 purpose.
6 So, a good analogy here is that NUREG-1791 7 provides a really comprehensive set of tools to 8 determine if staffing numbers are adequate.
9 Unfortunately, it's just written from a different 10 starting point, right? It's written to take what we 11 have in 50.54(m), and then, to move through a process 12 where we are pursuing exemptions from that.
13 So, we would have to, obviously, carefully 14 consider that process. And we do have a working group 15 right now that's working through a deeper analysis of 16 how to potentially apply key aspects of that NUREG in 17 a different way. But that's something, obviously, we 18 need to work through incrementally.
19 And additionally, it may be appropriate 20 for the Part 53 rule -- and again, this is another 21 option that we currently are considering -- it may be 22 appropriate for the Part 53 rule to also provide a 23 prescriptive staffing model as an option for 24 applicants that prefer not to conduct the staffing 25 analyses needed to support an alternative flexible NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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136 1 staffing model.
2 So, again, that's currently something that 3 is out there as part of our discussions. It is that, 4 while there is value, as we see right now, in moving 5 away from a prescriptive staffing requirement and 6 allowing for flexible staffing approaches, there may 7 also be value in having a different track that could 8 be taken by an applicant, to just perhaps go to a 9 prescriptive model with an option. So, again, that's 10 not something that we have further resolution on at 11 this point, but it is something that's currently in 12 our thinking.
13 Okay. If I can move on to the next slide, 14 please?
15 Okay. So, finally, how does human factors 16 engineering fit into the picture for advanced 17 reactors? And applications are likely to need to 18 contain specific information that is expected to be of 19 a human factors engineering programmatic nature. And 20 some examples of that are design for control 21 room/human system interfaces, or even proposals for 22 alternative staffing models, because those analyses 23 would be expected to be based on human factors 24 engineering methods. We would expect them to apply 25 human factors engineering principles there as well.
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137 1 So, we believe that Part 53 will likely 2 need to require advanced reactor applications to 3 address the incorporation of state-of-the-art human 4 factors engineering principles more comprehensively 5 than existing regulations require at present. And I 6 think a good way to put this is that the current human 7 factors engineering rule that we have in Part 50 very 8 neatly targets the control room.
9 And again, we talked about many different 10 things that could be different with advanced reactors.
11 What we don't want to do is just take a broad brush 12 and say, well, the entire facility falls under the 13 auspices of this rule, because that may not be the 14 prudent way to do that.
15 But what we do see is that we need to 16 follow important safety functions around the facility, 17 if you will, and apply those human factors engineering 18 principles where they're needed to make sure that --
19 and it's twofold -- to make sure that those safety 20 functions are going to be operated and managed 21 appropriately by the people involved with them; and 22 also, to help ensure that their continued evaluability 23 is supported. And that later point, I think that's a 24 difference between how we've traditionally thought.
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138 1 bit further into that point, an advanced reactor human 2 factors engineering program should be adequate to 3 ensure that humans can perform the full range of tasks 4 necessary to ensure, as I mentioned, the continued 5 evaluability of plant-specific safety functions.
6 And driving back to that point that I was 7 speaking to earlier, that may also extend to the 8 maintenance and testing activities related to plant 9 safety functions. An important example of this -- and 10 we go back to that hypothetical autonomous facility 11 that's just sitting there, you know, just running 12 without necessarily having a licensed operator, or who 13 knows who else, even overseeing it -- if you're going 14 to set the facility up, and you're going to perform 15 maintenance on it, and you're going to do inspections, 16 and then, you're going to start it up and walk away, 17 now there's different safety considerations that may 18 present themselves in different manners, right?
19 So, case in point, piping inspection that 20 you do on a passive safety system that's relied upon 21 as a layer of defense-in-depth may take on a new 22 significance because, if that pipe were to break, 23 there won't be anyone there to implement or that's 24 capable of implementing perhaps redundant action to 25 compensate for that.
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139 1 So, again, what we need to be aware of 2 here is that there may be cases where we need to 3 ensure that important maintenance testing and 4 surveillance activities, inspections, and so forth, 5 are properly supported by human factors engineering 6 principles to make sure that they can be done 7 correctly and consistently, and provide that assurance 8 for the continuing availability of those systems.
9 So, with that point, what I'd like to do 10 is move into one more slide, where I'll just provide 11 a brief overview of some of our ongoing work. And I 12 touched upon these earlier, but just a broad highlight 13 of our current efforts in further developing the items 14 that we talk about here.
15 So, I'd like to just outline some of these 16 efforts, and they are joint NRR and, also, Office of 17 Research projects. So, we do have working groups that 18 have been involved, you know, staff in both offices 19 involved here.
20 And these things relate to the topics that 21 I have on this slide. And so, we're working on, kind 22 of twofold, both drafting language and, also, to 23 propose regulatory guidance documents and approaches 24 to those that address the objectives and the proposed 25 solutions that we've covered within the white paper.
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140 1 So, specific examples where we are 2 currently working to draft rule language -- and again, 3 we've covered many of these going through -- the 4 prescriptive criteria that would need to be met in 5 order to allow for not using licensed operators.
6 Again, that's a work-in-progress. That is why we're 7 talking about it today and gathering the feedback that 8 we are. As well as prescribing certain types of 9 information that should be include with an application 10 made under Part 53.
11 With regards to development of regulatory 12 guidance, we are currently working to develop -- and 13 again, we've got a few projects that I'll highlight 14 here -- we're currently working to develop scalable 15 and technology-inclusive programs covering the areas 16 of staffing, operator licensing, and human factors.
17 And there's actually three interrelated, but somewhat 18 distinct efforts that are working on those three 19 areas.
20 And so, with that being said, I would like 21 to just see if there's any further questions or 22 comments. And if not, I'll go ahead and turn it over 23 to Juan Uribe for the conclusion.
24 (No response.)
25 MR. URIBE: Thank you, Jesse.
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141 1 So, hearing nothing, I know we're way past 2 the time. So, I'll just want to wrap up some 3 thoughts.
4 If you could put up the next slide, 5 please?
6 Okay. So, thank you, Jesse, again.
7 Again, the draft concepts in the white 8 paper are really meant to solicit feedback. I think 9 Jesse alluded to it, but it's not the intent to say 10 that we've decided on a certain path or not. Again, 11 we want to explore different options and get some 12 feedback, including ACRS. And so, obviously, that 13 means that the final scope and the guidance will be 14 determined or refined in sort of the coming months.
15 I also want to clarify that, up to this 16 point, we have received very limited stakeholder 17 feedback, and we do anticipate increased engagement 18 from stakeholders upon release of the proposed Subpart 19 F rule that aligns with the staffing and operations 20 concepts in the June-July timeframe.
21 We are also, to the extent that we can, 22 looking to the different efforts, so that we can 23 establish well-defined and unambiguous criteria, which 24 is critical for a performance-based, graded approach 25 related to operations. And what I mean by that is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 leveraging the results of the LMP analyses, the 2 maximum hypothetical accident, deterministic insights, 3 PRA insights, et cetera.
4 So, we definitely want to be risk-5 informed, but by no means exclusively risk-based. And 6 so, there is still an aspect that may be informed by 7 deterministic criteria, where appropriate.
8 Again, the white paper concepts are 9 intended for future 10 CFR 53 applicants, but we plan 10 to leverage the insights, the discussions here, for 11 Part 50 or 52, near-term applicants.
12 Next slide, please.
13 Additional stakeholder interactions will 14 follow in the coming months. As I previously alluded 15 to, the NRC staff is evaluating resources and 16 schedules to identify what areas to prioritize. Jesse 17 alluded to three ongoing efforts that we have, and 18 we're in coordination with the National Labs.
19 I think the concept of autonomous 20 operations versus remote, Jesse alluded to the 21 different efforts that Research is leading on the 22 remote aspect of that. So, I won't go into detail on 23 that again.
24 I'll just say that, as we move forward, 25 and we presented to you again one of the things that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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143 1 we want to do is provide a timeline or a more 2 definitive schedule with key deliverables and 3 milestones. And currently, we're putting that 4 together. So, that will follow.
5 And then, the final form of this IST is 6 still being evaluated, but we hope to follow the same 7 process that we've done before, where we can release 8 white papers to engage stakeholder discussion as we 9 complete sort of individual subparts of the guidance, 10 if you will. And then, we'll also have continued 11 discussions on how the final form of the guidance will 12 look.
13 So, with that, thanks again. I don't have 14 anything else.
15 CHAIR BLEY: Well, you've given us a lot 16 to think about. I hadn't thought about this until we 17 got all this discussion today. And maybe this ought 18 to go to John Segala, but perhaps Juan and Jesse or 19 Bill and Nan may want to comment on it.
20 You didn't give us any indication, but are 21 you looking for a letter on this in the short term?
22 MR. RECKLEY: Dennis, this is Bill 23 Reckley.
24 I think that will come as we develop Part 25 53 and make the decisions on how we're going to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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144 1 exercise this; what we're going to propose to the 2 Commission to possibly consider at this time. As Juan 3 and Jesse mentioned, this white paper was just a 4 beginning point. So, over the next few months, we'll 5 be making the decisions as to what to propose in Part 6 53, what we might acknowledge is premature to propose, 7 and so forth. And then, that would be a point for the 8 ACRS to weigh in, I believe.
9 CHAIR BLEY: Okay. That makes me feel a 10 little better because I think we have a lot to digest 11 and think about here.
12 I think at this point, unless there's 13 questions from the Committee members, this is an 14 appropriate place to break for lunch. It's one 15 o'clock, 1:06 back east. So, let come back at two 16 o'clock East Coast time. It's almost an hour for 17 lunch. We're in recess until that time.
18 (Whereupon, at 1:43 p.m., the foregoing 19 matter went off the record and resumed at 2:00 p.m.)
20 CHAIR BLEY: Okay, we are back in session.
21 I think now we're going to Bill, is that 22 right, Bill Reckley?
23 MR. RECKLEY: Yes.
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145 1 this morning.
2 MR. RECKLEY: No, no, I do as well.
3 You can start to get, also, a little 4 better grasp of the scope of our Part 53 efforts. And 5 basically, we end up touching everything. And so, 6 that's part of the challenge. So, as we go forward, 7 we'll have to come back. And you'll see this 8 afternoon when we start to talk about Subpart F, that 9 we've broken operations into the hardware, the people, 10 and the programs. And all three of those play a role, 11 and we're going to have to assess and, as you heard 12 this morning, really carefully consider the role of 13 people and how much or how willing we are in this 14 current effort under Part 53 to try to address some of 15 those potential issues. And so, that will be a future 16 meeting. We're not going to get into that this 17 afternoon.
18 The first thing we're going to talk about 19 this afternoon is Subpart A.
20 So, Liz, if you can go, yes, to slide 41.
21 And Jordan Hoellman is going to lead the 22 discussion of Subpart A, its layout, and a large part 23 of the discussion will be on the definitions.
24 So, Jordan, if you want to take over?
25 MR. HOELLMAN: Yes, great. Thanks, Bill.
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146 1 So, good afternoon. My name is Jordan 2 Hoellman.
3 You guys can hear me right, okay? Okay, 4 okay. I just wanted to make sure.
5 So, I'm a project manager in the Advanced 6 Reactor Policy Branch in NRR, and I'm excited to be 7 here today to discuss our first iteration of the 8 preliminary proposed rule language for Subpart A, 9 "General Provisions".
10 Subpart A really demonstrates the 11 iterative nature of the development of Part 53. As 12 other subparts are written and/or revised, Subpart A 13 will need to reflect the changes and capture other key 14 terms used throughout the rule. We discussed Subpart 15 A in our stakeholder meeting on May 6th, and we 16 received some good feedback from our public 17 stakeholders and industry on the Subpart A definitions 18 that we will continue to consider as we develop the 19 proposed rule.
20 So, you can move on to the next slide.
21 Well, probably the next slide, too.
22 Subpart A includes the scope, definitions, 23 written communications, employee protections, 24 completeness and accuracy of information, exemptions, 25 combining licenses, jurisdictional limits, attacks and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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147 1 destructive acts, and information collection 2 requirements.
3 We are intending to develop Part 53 with 4 largely no cross-references to either Parts 50 or 52.
5 This will require in some cases copying and pasting 6 language from either Part 50 or Part 52, where 7 appropriate. Instead of using cross-references, this 8 approach will apply throughout Part 53; at least 9 that's our initial thought.
10 Yes, Dennis?
11 CHAIR BLEY: Yes. NRC has a number of 12 published glossaries, including a very long one 13 associated with PRA. Have you bounced these 14 definitions against those other glossaries? Are we 15 redefining them? Or are we being consistent?
16 MR. HOELLMAN: Yes, so I forget what the 17 NUREG number is, but we did look at the NUREG, and I 18 think that's the glossary you're talking about that 19 defines the PRA terms.
20 Generally, we looked at NEI 18-04 and the 21 non-light water reactor PRA standard, and tried to 22 draw the definitions we're using in Subpart A from 23 those. Because, as we discussed in the rulemaking 24 plan, and subsequent documents, Part 53 is intended to 25 have, you know, using the Licensing Modernization NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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148 1 Project approach as one acceptable way for meeting the 2 rule. So, we wanted to make sure that we were being 3 consistent with those two documents for sure, and 4 then, like I mentioned, we used the NUREG. And I'm 5 sorry, I can't remember what number is, but you know 6 which one I'm talking about.
7 CHAIR BLEY: Yes, I couldn't remember --
8 MR. HOELLMAN: The one you mentioned, yes, 9 yes. So, yes is the simple answer, I guess.
10 CHAIR BLEY: Not to beat this to death, 11 but the LMP document, the new, just came out, I guess, 12 pretty recently, a couple of months ago, PRA standard 13 for non-light water reactors and that other glossary, 14 I have no idea if they're consistent in their 15 definitions, but I hope they are. It sounds like 16 you've tried to check.
17 MR. HOELLMAN: Yes, yes. So, we have, and 18 they are generally consistent. Like I said, I think 19 we tried to go with the non-light water PRA standard 20 definition more, you know, used the terms and words 21 more consistent with that document than the NUREG.
22 But, in general, yes, they're consistent. I mean, 23 some wording changes and things like that are 24 possibilities, I guess, but we wanted to make sure 25 that both the non-light water reactor PRA standard and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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149 1 the LMP document we were being consistent with those, 2 because --
3 CHAIR BLEY: I think that makes sense to 4 me, but, eventually, there will be various NRC 5 existing and new guidance documents, and I hope we can 6 stay consistent. Otherwise, we get confusion.
7 MR. HOELLMAN: Yes, and you'll see as we 8 move through some of the definitions, you know, we had 9 the discussions earlier this morning on defense-in-10 depth and inherent characteristics. And so, I think 11 that human factors white paper was issued a little 12 before the Subpart A definition section. So, there 13 may be slightly inconsistencies there, or not really 14 inconsistencies; just the words, they don't perfectly 15 line up, I guess, if that makes sense.
16 CHAIR BLEY: Go ahead.
17 MR. HOELLMAN: Okay. So, I just talked 18 about how we're planning on not using cross-references 19 and copying and pasting language from Part 50 or 52, 20 where appropriate. Instead, we think that this will 21 cause less confusion when someone is applying under 22 Part 53 about the applicability of requirements in 23 either Parts 50 and 52. If we did cross-references, 24 I guess the idea is we thought that people would get 25 confused, depending on which section they were looking NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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150 1 at, which regulations actually apply. So, that would 2 make Part 53 a, quote, "one-stop shop," and an 3 optional separate licensing framework from either 4 Parts 50 or 52.
5 So, this iteration of Subpart A reproduces 6 Part 50 in a lot of places, and currently includes 7 bracketed references to the existing requirements in 8 Parts 50 and 52, and those brackets will be replaced 9 with the applicable Part 53 requirements, once 10 developed.
11 So, we included bracketed text. I think 12 in the next iteration we put out of Subpart A we'll 13 probably include a discussion table and the bracketed 14 information would move to the discussion side of that 15 table, as we've done for the other released subparts.
16 So, in some places, the bracketed text 17 indicates references; in other places, it's intended 18 to capture things that would have been included in the 19 discussion table, and that provides an explanation or 20 indicates concepts that are still under development 21 and may be further revised or removed in future 22 iterations, as we work to develop the proposed rule.
23 Subpart A will also include many pointers 24 to other sections of Part 53 that need updating in 25 future iterations. And some examples include NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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151 1 emergency plan and related submissions in the security 2 plan, and related submissions portions of Subpart A.
3 Today, we're just going to focus mainly on 4 the definitions portion of Subpart A, and 5 specifically, focusing on terms that have been 6 discussed in previous meetings related to event 7 categories and SSC classifications.
8 As I mentioned, the definitions will need 9 to be updated, as other key terms needing to be 10 defined or identified or revised, as we continue to 11 develop the proposed rule.
12 And you'll probably notice that the terms 13 related to fusion are in gray. As you may know and as 14 we discussed in the staff's response to the rulemaking 15 plan SRM, the staff is working to develop options for 16 the regulation of fusion energy systems and the 17 staff's assessments of the potential risks posed by 18 various fusion technology, and possible regulatory 19 approaches for fusion facilities is being done in 20 parallel with developing the draft proposed rulemaking 21 package for Part 53. This work is supporting an 22 options paper to the Commission. The draft proposed 23 Part 53 rule will be developed with the aim of 24 accommodating fusion technologies as much as possible 25 to maintain flexibility for future Commission NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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152 1 direction on the appropriate approach for licensing 2 and regulating fusion energy systems.
3 So, yes, let's move on to the next slide, 4 please.
5 Advanced nuclear plant and advanced 6 nuclear reactor, these two terms, there's two things 7 we want to note on these. No. 1 is that we used the 8 term "advanced nuclear plant" throughout the rule 9 text. The term "plant" includes one or more reactors 10 and a collection of sites, buildings, radionuclide 11 sources, and structure systems and components for 12 which the license is being sought.
13 And we're currently using the NEIMA 14 legislation definition of "advanced nuclear reactor".
15 Since NEIMA is the statute that's directing the NRC to 16 develop Part 53, we thought we needed to stay 17 consistent with how NEIMA defines an advanced reactor.
18 The NEIMA definition does not exclude future large 19 light water reactors, so long as they have significant 20 improvements compared to reactors under construction 21 when NEIMA was enacted in January of 2019. And so, 22 the AP1000s were under construction at that time, and 23 some of the specific items listed in NEIMA are 24 reproduced; the definition is reproduced there at the 25 bottom of the slide.
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153 1 So, the staff is considering how to deal 2 with this part of the NEIMA definition to provide the 3 appropriate specificity. At this point, we're not 4 planning on restricting the definition of "advanced 5 nuclear reactor" in the rule. We want to maintain 6 flexibility, and in our technology-inclusive rule, we 7 want to be also responsive to Congress' direction in 8 NEIMA. So, it's likely that the technical 9 requirements would sort of provide the applicability 10 for an applicant to use Part 53.
11 And so, this was an area that we were 12 specifically interested in stakeholder views. And so, 13 there were some varying views from stakeholders on 14 this one, and the general feedback was that we should 15 be expansive in our consideration of what qualifies 16 under this definition. And some stakeholders pointed 17 to the broad construction of the definition used in 18 NEIMA. So, this is something that we're continuing to 19 consider, I think, as we move through developing the 20 rule and need to maintain our coordination with the 21 various other rulemakings in process that are touching 22 on advanced reactors.
23 So, let's move to the next slide.
24 Consensus code or standard. This we 25 talked about in a couple of meetings. And so, we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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154 1 adapted this definition from ASME and it's consistent 2 with the National Technology Transfer and Advancement 3 Act. And the definition is intended to provide 4 flexibility in the codes and standards an applicant 5 could use. The generally accepted modifier used in 6 other portions of the rule language is intended to be 7 captured by this definition.
8 The staff is actively reviewing some 9 advanced reactor codes and standards for endorsement.
10 These are intended to be endorsed with any conditions 11 in Regulatory Guides, and this endorsement via 12 Regulatory Guide approach is planned to be continued 13 for future guidance supporting Part 53. And some 14 examples include -- that you're probably aware of and 15 we've discussed -- ASME Section III, Division 5, for 16 high temperature reactors; ASME Section XI, Division 17 2, for reliability and integrity management programs, 18 and the non-light water reactor probabilistic risk 19 assessment standard.
20 So, the capture of acceptable standards 21 and guidance provides flexibility and increases 22 efficiency by avoiding routine rulemakings related to 23 the adopting and revision of incorporated standards in 24 the regulations. And while this doesn't provide the 25 same degree of regulatory stability as is provided by NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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155 1 doing a rulemaking, it does provide one acceptable way 2 of using the code or standard.
3 Some stakeholder feedback we've gotten on 4 this that we continue to consider is in item 3 of this 5 definition. So, it includes 1, 2, and 3. And the 6 comment was, "Does the standard really need to be 7 designated for nuclear power plants?" And so, this is 8 something that we're continuing to consider, as we 9 develop the rule language.
10 So, let's move on to the next slide.
11 So, "end state" and "event sequence,"
12 these terms were included here on the same slide 13 together because they're related and reference each 14 other in the definitions. The definitions are 15 generally adapted from the Licensing Modernization 16 Project, or NEI 18-04.
17 "End state" is the set of conditions at 18 the end of an event sequence. This definition is 19 intended to include the safe/stable language to make 20 it clear that there would be no further event 21 progression beyond the identified end state. And 22 that's consistent with the non-light water reactor PRA 23 standard.
24 One of the comments that we got at the 25 public meeting earlier this month is that we should NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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156 1 include the safe/stable language in this definition to 2 be consistent with that. We've also heard comments 3 from the ACRS members that safety analysis must 4 demonstrate that the DBAs achieve and maintain a 5 safety, stable, and subcritical condition.
6 So, as we have these discussions with both 7 the members of the ACRS and our public stakeholders 8 and industry, as you are aware, there's many, many 9 comments that we continue to receive and need to 10 consider holistically on how they affect multiple 11 portions of the rule language, as we continue 12 developing the complete rule or the proposed rule.
13 "Event sequence" mirrors the definition in 14 the non-light water reactor PRA standard, and is 15 intended to capture a typical PRA event sequence and 16 provides a lead into the next slides that cover the 17 various event sequences defined.
18 So, we can move on to the next slide, 19 unless the --
20 MEMBER KIRCHNER: Jordan, out of 21 curiosity, this is Walt Kirchner, are they definitions 22 -- those two previous definitions on the prior view 23 graph, are those the same in the LWR PRA standard and 24 the non-LWR PRA standard?
25 MR. HOELLMAN: Yeah, and Walt, that's a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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157 1 good question. I'm not sure. I specifically --
2 MEMBER KIRCHNER: I don't see generically 3 why they would be different, but it's just a curiosity 4 question.
5 MR. HOELLMAN: Yeah, I don't see why they 6 would be different either, but I can't, you know, 7 commit to the fact that I didn't actually compare the 8 two personally, so I apologize for not being able to 9 address that more clearly or better, but I'll check.
10 MEMBER KIRCHNER: No, go on. That's fine.
11 Thank you.
12 MR. HOELLMAN: Okay, yeah, good question 13 though. I mean, I probably should have checked that, 14 but thank you for that. So, now we'll talk about the 15 event sequences that we've defined so far for Part 53.
16 So, normal plant operation covers events 17 that are expected to occur during the planned 18 operation or shutdown of the reactor. Licensing basis 19 events are the collection of event sequences 20 considered in the design and licensing of a plant, 21 LBEs are unplanned events that include AOOs, DBAs, 22 unlikely event sequences and very unlikely event 23 sequences.
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158 1 for the design of safety-related structure systems and 2 components, and they're based on the capabilities and 3 reliabilities of safety-related structure systems and 4 components needed to mitigate and prevent event 5 sequences respectively.
6 Within the LMP, DBAs are derived from 7 design basis events or DBEs, but assume only safety-8 related SSCs are available to respond to an event.
9 So, in Part 53, the equivalent concept would mean DBAs 10 are derived from unlikely event sequences, but assume 11 only the safety-related SSCs are functioning.
12 For anticipated operational occurrences, 13 these are expected to occur one or more times during 14 the life of a nuclear power plant, and an AOO is 15 defined as event sequences that are unplanned with a 16 mean frequency of one times ten to the minus second 17 per plant year and greater, and AOOs take into account 18 the expected responses of all SSCs within the plant 19 regardless of safety classification.
20 Unlikely event sequences have a frequency 21 range below the frequency of AOOs. Within the LMP, 22 this would equate to design basis events. Both the 23 DBEs, and the licensing modernization project, and 24 unlikely event sequences in Part 53 have a frequency 25 range between one times ten to the minus second and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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159 1 one times ten to the minus fourth per plant year with 2 an accounting for uncertainties.
3 Oh, I guess we're onto the next slide, 4 sorry. These unlikely event sequences and very 5 unlikely event sequences are on slide 48. We did 6 notice and it was brought up in our public meeting 7 that we did make an error in the frequency range 8 numbers. It said five.
9 Under unlikely event sequences, it said 10 one times ten to the minus second two and five times 11 ten to the minus fourth, and that was in error, so 12 we've corrected it and highlighted it on this slide 13 just for folks' information and so we didn't continue 14 to get the comment, I guess.
15 So, very unlikely event sequences have 16 estimated frequencies well below the frequency of 17 events expected to occur in the life of an advanced 18 nuclear plant.
19 For unlikely event sequences and very 20 unlikely event sequences, they take into account the 21 expected responses of all SSCs within the plant 22 regardless of safety classification just like AOOs.
23 Within the LMP, this equates to the beyond 24 design basis events, and both the beyond design basis 25 events in LMP and very unlikely event sequences in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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160 1 Part 53 have a frequency range between one times ten 2 to the minus fourth and five times ten to the minus 3 seventh per plant year with an accounting for 4 uncertainties.
5 So, let's move onto the next slide and 6 this is going to be our SSC classifications, and so 7 these terms are under development and may be revised 8 in future preliminary proposed rule text, but we 9 wanted to provide an initial definition which will 10 hopefully help clarify the way these terms have been 11 discussed throughout our discussions on Subparts B and 12 C. So --
13 MEMBER DIMITRIJEVIC: Before you -- sorry.
14 MR. HOELLMAN: Yeah, go ahead.
15 MEMBER DIMITRIJEVIC: This is Vesna 16 Dimitrijevic. Before you go into these 17 classification, what happened with the sequences below 18 frequency of five E minus seven?
19 MR. HOELLMAN: Well, after you account for 20 uncertainties, I guess, in the LMP, I don't think you 21 need to consider them any further because you're 22 crediting the response of all the SSCs regardless of 23 their classification throughout the plant.
24 So, I think below a very likely event 25 sequence, or in LMP terms, a beyond design basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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161 1 event, I don't think we would require -- you know, I 2 get --
3 MS. CUBBAGE: So, Jordan?
4 (Simultaneous speaking.)
5 MR. HOELLMAN: I don't know. Yeah, Amy, 6 if you could help, or Bill, if you could help?
7 MS. CUBBAGE: Yeah, go ahead, Bill, yeah.
8 MR. RECKLEY: Yeah, what is -- that is the 9 cutoff for being considered a licensing basis event, 10 and so looking at the specifics of what would need to 11 be done to address that specific event sequences as a 12 licensing basis event is as Jordan was just saying.
13 It would not fall within the requirements.
14 However, the PRA will look at those lower 15 sequences, lower frequency sequences in assessing the 16 cumulative risk measures, and also looking at the 17 sequences for the other considerations that are 18 included in our discussions, things like cliff edge 19 effects. Some of the sequences might get looked at 20 just to make sure if they fall into the category of a 21 cliff edge event sequence.
22 So, they're not totally ignored, but in 23 terms of their treatment as a licensing basis event, 24 they would not be included in any of the categories 25 that we were talking about.
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162 1 MEMBER DIMITRIJEVIC: Well, you know, like 2 because you say unlikely, if you go to previous 3 slides, you know, that you have a category said very 4 unlikely event sequences. I mean, obviously sequences 5 below five E minus seven are definitely very unlikely.
6 So, the thing is you can say the very 7 unlikely event sequences are any sequence below one E 8 minus four, you know, and then because -- I don't 9 really know why it's the range here and I was 10 wondering does this come from this advanced PRA 11 standard or this is just your thoughts on these 12 subjects?
13 Because, you know, you are very likely to 14 have -- I mean, you have this at safety. You know, we 15 already saw the reactors with, you know, below ten to 16 the minus nine, the safety, the criteria, so, I mean, 17 I just don't know why is this range necessary here?
18 That was my question.
19 MR. RECKLEY: Yeah, the ranges here are 20 again for event categories and they are taken from the 21 licensing modernization, NEI 1804 and Reg Guide 1.233.
22 That's where those ranges are coming from.
23 I will say they are similar to ranges that 24 are used in other parts of NRC regulations like Part 25 70 also. Part 70 uses unlikely with the same range NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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163 1 and highly unlikely with a slightly different range, 2 and that's why we used --
3 Going to Dennis's point, it's very tricky 4 to try to come up with definitions that haven't 5 already been used, but we used very unlikely as 6 opposed to highly unlikely because highly unlikely is 7 a term used in Part 70.
8 So, we are challenged in that regard, but 9 the numbers and the ranges come specifically from Reg 10 Guide 1.233 and NEI 1804.
11 MEMBER DIMITRIJEVIC: Okay, thank you.
12 MR. HOELLMAN: Okay, thank you, Bill.
13 That was much better than I would have done.
14 So, going back to the SSC classifications 15 here, for safety related, these are SSCs and human 16 actions that warrant special treatment, and are relied 17 upon to demonstrate compliance with the safety 18 criteria in 53.210(b) or the first tier safety 19 criteria for unplanned events. So, safety-related 20 SSCs are needed to mitigate and prevent design basis 21 accidents.
22 Non-safety related but safety significant 23 are those SSCs and human actions that are not safety 24 related, but warrant special treatment and are relied 25 on to achieve defense in depth or perform risk-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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164 1 significant functions.
2 Non-safety related but safety significant 3 SSCs contribute to mitigation and prevention of AOOs, 4 unlikely event sequences, and very unlikely event 5 sequences.
6 Non-safety significant are those SSCs 7 which are not safety related, do not warrant special 8 treatment, and are not relied on to achieve adequate 9 defense in depth or to perform risk-significant 10 functions, and so non-safety significant SSCs perform 11 functions during normal operations.
12 And so as we were, you know, as we were 13 going through this, we've gotten some comments about 14 defining what special treatment for each of these 15 terms, and that's something we continue to consider as 16 we move forward.
17 And besides the DBAs, I think we think we 18 can be a little more clear on the acceptance criteria 19 needed for each event sequence and, you know, how --
20 you know, with respect to warranting special treatment 21 and how the special treatments are actually applied to 22 each SSC classification.
23 MEMBER DIMITRIJEVIC: So, why do you think 24 that -- I mean, why do you feel the need to keep the 25 safety related and the non-safety related but safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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165 1 significant, those two categories? Did you consider 2 merging those two?
3 MR. HOELLMAN: Well, so the safety-related 4 SSCs are used for developing your design basis 5 accidents, and in your design basis accidents, you can 6 only rely on the capabilities and reliabilities of 7 your safety-related structures, systems, and 8 components to mitigate those event sequences.
9 The non-safety related but safety 10 significant SSCs are intended to give a potential 11 licensee more flexibilities in mitigating your 12 unlikely event sequences, or AOOs, or very unlikely 13 event sequences. So, we --
14 MEMBER KIRCHNER: Jordan, that's not 15 consistent. Let me back you up a little bit. On 16 safety related, it would really be useful to compare 17 your definition and its application to what's 18 currently in 10 CFR 50 because 10 CFR 50 has safety-19 related SSCs applied to all design basis events.
20 So, I guess in your terminology here, this 21 would be your licensing basis events, which includes 22 AOOs and DBAs. Do you see the distinction I'm making?
23 And then it goes further. It defines them. I have it 24 actually in front of me.
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166 1 remain functional during an following design basis 2 events to assure, and then it goes on with some 3 specificity that I think some people may object to.
4 The first two things are integrity of the 5 reactor coolant pressure boundary. The second one is 6 capability to shut down the reactor and maintain it in 7 a safety shutdown condition.
8 And the third is pretty much comparable to 9 your safety criteria because it then goes on to say 10 the capability to prevent or mitigate consequences of 11 accidents which could result in offsite exposures, 12 blah, blah, blah, and those exposures are the same as 13 in your safety criteria.
14 So, some consistency probably is warranted 15 here, at least -- I'm not saying pulling the words 16 about the reactor pressure boundary because that 17 wouldn't apply to some of the advanced designs, but I 18 just -- I'm sensing some inconsistency.
19 If I heard you correctly, you said non-20 safety related but safety significant would be applied 21 for AOOs and I think it's your -- to get through the 22 AOOs and the DBAs, which in effect are the design 23 basis events or your licensing basis events --
24 (Telephonic interference.)
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167 1 from the line, the open line. Let me just pause. I 2 would require the safety-related SSCs would be those 3 that get you through essentially the design basis or 4 licensing basis events. I mean, that's the shorthand, 5 and meet your safety criteria.
6 MR. HOELLMAN: Yeah, Walt, so I think 7 maybe there's a little disconnect. I think we're not 8 -- so we're not using the same definition of safety 9 related that's used in Part 50.
10 MEMBER KIRCHNER: No, I realize that.
11 MR. HOELLMAN: Okay.
12 MEMBER KIRCHNER: That's the inconsistency 13 I wanted to point out, but I think you can construct 14 a comparable, in effect a comparable definition 15 because the one in 50 ends with the same, meeting the 16 same safety criteria as you define, but it says for 17 all design basis events and I would say the equivalent 18 here would be for all of the licensing basis events, 19 in other words, up to the ten to the minus four per 20 reactor year.
21 MR. RECKLEY: Yeah, Walt, this is Bill 22 Reckley, and we had a lot of this discussion during 23 the development, the review of NEI 1804 and the 24 development of Reg Guide 1.233 because this is a 25 difference.
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168 1 And, you know, one way to look at it is if 2 you take the definition that we have currently which 3 focuses on the dose criteria, the 25 rem number, and 4 you look at it in the context of the analyses that 5 were available when the Part 100 definition was 6 developed, the large break LOCA as basically the event 7 that would challenge that dose value, as was ATWS when 8 it was assessed at the time.
9 And so you get basically to the same 10 functions of pressure boundary and reactivity control 11 or shutdown when you apply this, when you use our 12 definition and apply it to a large light water reactor 13 as it was analyzed at least through the 1970s.
14 So, and the distinction between the event 15 categories with the licensing basis events being 16 somewhat different in that you're exactly right, what 17 you're reading. The design basis events for light 18 water reactors as it's reflected in Part 50 includes 19 anticipated operational occurrences, the equivalent of 20 what we call DBAs, and external events that --
21 The one distinction, and again, we talked 22 about it in the development of the reg guide, is in 23 the area of anticipated operational occurrences and 24 not needing to address AOOs using only safety-related 25 equipment, as has been the practice for light water NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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169 1 reactors and it's reflected in Part 50.
2 So, that is a difference. It's a 3 conscious one and it's just an area that we'll 4 continue to have to try to explain that that is a 5 difference.
6 MEMBER BIER: If I can chime in on this, 7 this is Vicki Bier. I'm coming somewhat new to this 8 and I realize that this has all been under development 9 for a long time, which I haven't been tracking.
10 But kind of getting back to Vesna's 11 original question, with the licensing for the current 12 generation of LWRs, we were kind of stuck with this 13 term of non-safety related but safety significant 14 because after safety related was defined, we learned 15 more through PRA about what was actually safety 16 significant.
17 But the terminology itself is really super 18 unhelpful and difficult to explain. You know, why is 19 it that something that is, you know, quote, unquote 20 non-safety related can actually be more important to 21 safety than something that is safety related?
22 So, I would just encourage you to look 23 back at whether there is a way that you could take 24 advantage of the fact that you are starting fresh with 25 a clean slate and, you know, possibly just have two NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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170 1 categories, safety related and not safety related, or 2 call that middle category, you know, generate new 3 terms for those, you know, safety critical versus 4 safety significant or something that is not, you know, 5 kind of an oxymoron in a sentence, so just my two 6 cents.
7 MR. RECKLEY: We'll consider it. Part of 8 the problem is that we were still trying to rely on 9 some of the infrastructure, for example, NQA1, the QA 10 standard.
11 And so in using some of the same 12 definitions and relying on that same infrastructure, 13 we thought it was probably less burden, but I agree 14 with you 100 percent. In doing that, in using that 15 same infrastructure, you bring in some of the odd use 16 of terms that have evolved over the years, so.
17 MEMBER KIRCHNER: Is there an accepted, 18 Bill or Jordan, is that an accepted definition for 19 warrants special treatment or special treatment, I 20 guess, is what I'm --
21 MR. HOELLMAN: On the next slide, we'll 22 talk about the definition of special treatment --
23 MEMBER KIRCHNER: Okay.
24 MR. HOELLMAN: -- which will be another 25 interesting discussion, I'm sure.
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171 1 Like I said, these terms, I think, have 2 been causing some confusion in our public meetings as 3 well, and so I want to make sure at least we addressed 4 Walt's question and comment there.
5 So, in Part 53, the way we have it set up 6 is the licensing basis events are defined as unplanned 7 events that include AOOs, unlikely event sequences, 8 very unlikely event sequences, and DBAs.
9 The safety-related SSCs are solely relied 10 upon to mitigate DBAs and the non-safety related but 11 safety significant SSCs and non-safety significant 12 SSCs are --
13 Well, regardless of event classification, 14 those can all be relied upon to address AOOs, unlikely 15 event sequences, and very unlikely event sequences.
16 I'm not sure if that helped or hurt honestly, but --
17 So, regardless of event classification, 18 for an AOO and an unlikely event sequence or a very 19 unlikely event sequence, you can take into account the 20 expected responses of all of the SSCs within the 21 plant, but for a DBA, they're derived from your 22 unlikely event sequences and assume only safety-23 related SSCs are available to respond to the event.
24 MR. RECKLEY: Just I'll probably confuse 25 matters a little more, but --
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172 1 MR. HOELLMAN: Sorry.
2 MR. RECKLEY: No, one difference is that 3 the Part 50 analyses, keep in mind, are stylized 4 events and you are, as Jordan is talking about, taking 5 credit or assuming the availability or unavailability 6 of equipment in a stylized assessment.
7 In what we envision for these event 8 categories, the plotting in terms of which category 9 they fit into in terms of AOO, design basis event, or 10 beyond design basis event, or using Part 53, AOO, 11 unlikely, or very unlikely event, is coming from the 12 systematic assessment that is modeling the reliability 13 and availability of equipment and you're simply 14 plotting it as the estimated frequency coming out of 15 that analysis, and so it's going to have the safety 16 related, non-safety related, all of the equipment 17 that's available in the plant, and assuming failure 18 rates and so forth.
19 So, it's a little bit of apples and 20 oranges when you compare the analysis from Part 50 and 21 the analyses that are talking about in Part 53.
22 There's some similarities, but there's also that 23 distinction, that you're mapping all of the sequences 24 under Part 53 and you're just plotting them based on 25 the mean frequencies and then accounting for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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173 1 uncertainties in putting them into the categories.
2 So, again, as we talked about last time, 3 I know a lot of this -- we have some new members and 4 we also have just the passage of time since we talked 5 about NEI 1804 and that methodology, so we're facing 6 that challenge. Maybe we can have another briefing or 7 something on that.
8 CHAIR BLEY: Bill, I think that's a good 9 idea. Maybe you'll see this in our letter, so I 10 probably shouldn't bring it up yet, but I'm going to.
11 All of what you said kind of seems 12 reasonable except when you put in language softening 13 the need for a PRA and changing it to some qualitative 14 things. Then some of the claims of things that are 15 covered in the LMP kind of disappear.
16 I don't know. We'll want to talk about 17 that, but I don't want you to spend time on that 18 today.
19 MEMBER MARCH-LEUBA: Hey, Bill, but I do 20 want him to spend a little time. You managed to 21 confuse me. You said that for your Part 53 safety 22 analysis, you will assume all your equipment is 23 working, including non-safety related?
24 MR. RECKLEY: We'll model the plant with 25 the assumed failure rates --
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174 1 (Simultaneous speaking.)
2 MEMBER MARCH-LEUBA: So, you will take 3 credit for non-safety grade as long as you build it 4 properly so it has a good reliability, but it's not 5 safety related, you will take credit for it?
6 (Simultaneous speaking.)
7 MR. RECKLEY: We'll model it and its 8 capabilities and its reliabilities, yes.
9 CHAIR BLEY: It will give it a chance to 10 work and a chance to fail.
11 MR. RECKLEY: Right.
12 MEMBER MARCH-LEUBA: So, how does anything 13 become safety related?
14 MR. RECKLEY: It becomes safety related 15 when it's serving a function that is needed to keep 16 the -- to satisfy the first tier safety criteria, the 17 one in the --
18 (Simultaneous speaking.)
19 MR. RECKLEY: -- 210.
20 MEMBER MARCH-LEUBA: Being in the model is 21 not serving a function? I mean, what I thought we 22 agreed to was if it's not safety related, it will be 23 assumed in the analysis to be failed in the most 24 damaging condition, period.
25 MR. RECKLEY: That's for the DBA.
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175 1 MEMBER MARCH-LEUBA: What are you talking 2 about, AOOs?
3 MR. RECKLEY: Design basis accident would 4 only credit safety-related equipment.
5 MEMBER MARCH-LEUBA: Okay, so explain to 6 me again what you just said. Okay, so for DBAs, you 7 will assume that the non-safety related equipment will 8 fail in the most damaging possible state, yes or no?
9 Because you're going back on what you said.
10 MR. RECKLEY: I feel comfortable saying it 11 will credit only safety-related equipment.
12 MEMBER MARCH-LEUBA: But you will model 13 the non-safety related equipment as it was there and 14 working?
15 MR. RECKLEY: Right, but that might be 16 different than to say it's assumed to be failed 17 depending on the type of equipment that you're talking 18 about, so that's my hedging, but it is analogous to 19 what is currently done for the design basis accident 20 for Part 50 in terms of how it would be addressed, 21 only crediting safety-related equipment.
22 MEMBER MARCH-LEUBA: Only crediting 23 safety-related equipment, but putting in the model the 24 non-safety related equipment as if it was working.
25 What's the difference between creating and modeling?
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176 1 Okay, think about it, and I need to see 2 this in writing then because this is not kosher, what 3 you're saying. Maybe what you're doing is okay, but 4 what you're saying is not. All right, I'm going back 5 to mute.
6 MS. CUBBAGE: This is Amy Cubbage. I just 7 want to make sure I understand your comment, Jose. I 8 mean, the PRA has to model the full facility, whether 9 the equipment is safety related or not. Is that what 10 you're getting at?
11 MEMBER MARCH-LEUBA: No, for your safety 12 analysis to demonstrate that you are -- you survive 13 your AOOs and DBAs, you have to perform an analysis 14 and you --
15 MS. CUBBAGE: Yes.
16 MEMBER MARCH-LEUBA: Okay, and for those 17 AOOs, you only credit, meaning only exist --
18 MS. CUBBAGE: Okay.
19 MEMBER MARCH-LEUBA: -- things that are 20 safety related. If you put in your model all sorts of 21 things that are non-safety grade but don't credit 22 them, but you used them, how does anything become 23 safety grade? Nothing is safety grade --
24 MS. CUBBAGE: Okay, so --
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177 1 working.
2 MS. CUBBAGE: So, I guess if your issue is 3 AOOs, then we could discuss that further, but for the 4 DBAs, it's only the safety-related equipment that's 5 credited.
6 MEMBER MARCH-LEUBA: But the non-safety 7 related equipment is assumed to be working --
8 MS. CUBBAGE: No.
9 MEMBER MARCH-LEUBA: -- in your 10 calculations?
11 MR. RECKLEY: No, so let's say you have 12 three systems that are able to remove heat. So, when 13 you're assessing the highly unlikely event, or the 14 unlikely event and highly unlikely event, you're going 15 to model those, and some of those systems are failing.
16 That's what's making them unlikely or very 17 unlikely, but the need to remove heat is a required 18 safety function, and so you're going to pick one of 19 those three as being safety related.
20 And having picked it as being safety 21 related, you're going to do the DBA analysis and 22 you're only going to credit that system. You will 23 assume the other systems are not working.
24 MEMBER MARCH-LEUBA: Are failed in the 25 most unwanted condition because they're not safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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178 1 related. That means it can fail on. It can fail off.
2 It can fail halfway. That's what -- non-safety 3 related means it can be on any state whatsoever.
4 MS. CUBBAGE: So, when you do the -- I'll 5 probably need Marty or someone to jump in here, but 6 when you do the full PRA and look at all of the 7 different sequences, then all of these different 8 scenarios are going to come into play, whether the 9 non-safety system works or it doesn't work, but when 10 we do the analysis for the DBA, you're just crediting 11 the function of safety-related equipment.
12 MEMBER MARCH-LEUBA: Correct, and the non-13 safety related equipment, what state are they in --
14 MS. CUBBAGE: Non-functioning.
15 MEMBER MARCH-LEUBA: -- when you're doing 16 your -- non-functioning. No, it has to be the worst 17 possible condition. Because it's not safety grade, I 18 don't know, I mean, like you just bought it from Home 19 Depot yesterday.
20 MS. CUBBAGE: So, we'll have to take that 21 back and think about it.
22 MEMBER MARCH-LEUBA: Yeah, you need to 23 specify what you're planning to do because, I mean, 24 we're taking all of the safeties out of Part 50 and 25 52. I mean, you're just relaxing everything and you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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179 1 have to make sure you do the things right. This is --
2 it's becoming dangerous, guys. It's becoming 3 dangerous.
4 (Simultaneous speaking.)
5 MEMBER PETTI: So, Jose, I think some of 6 this has to do with licensing modernization. These 7 are questions that I'm sure have been asked, and 8 visited, and revisited as part of Reg Guide 1.232, 9 233.
10 As I understand it, it's how Amy described 11 it. You take credit for the safety-related systems, 12 so those are DBAs. There are also DBEs which are --
13 MEMBER MARCH-LEUBA: But if a system is 14 non-safety related, it means it can fail on, off, 15 halfway. It can fail any different number of -- when 16 you do your PRA, you can put frequencies to each of 17 those, but when you're doing your DBAs and you're 18 writing your Chapter 15, you're saying I don't depend 19 on that system to be in any particular condition.
20 MEMBER PETTI: So, Jose --
21 MEMBER MARCH-LEUBA: And I know that 22 because I have analyzed all possible conditions.
23 MEMBER PETTI: So, your idea of worst 24 possible condition will come through. Let's look at 25 a class of DBEs, loss of heat removal, okay? The NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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180 1 highest frequency DBE will be the basis to turn it 2 into a DBA, and that will incorporate whether or not 3 it was fully non-functional, functional, or partially 4 functional. That will come through the PRA, I think.
5 MEMBER MARCH-LEUBA: You think?
6 MEMBER PETTI: Yes.
7 MEMBER MARCH-LEUBA: I don't think the 8 staff can explain it properly. Okay, so, Amy, you 9 think about it and come back to us in a meeting.
10 MS. CUBBAGE: Okay, I don't know what that 11 beeping noise is. Okay, thank you.
12 Yeah, we'll think about it, but I think, 13 as Member Petti was just describing, you look at all 14 of these different scenarios and all the possible 15 failures in coming up with a whole list of what the 16 events are, and then the most challenging ones end up 17 being the DBA because then you just don't get the non-18 safety systems to function.
19 I think we'll just have to figure out a 20 different way to explain it better, but we'll also go 21 back and think about it and make sure there isn't a 22 gap.
23 And I also want to respond to your 24 previous comment about dangerous. I mean, we 25 definitely need to listen and make sure that this is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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181 1 done properly, but I can assure you that the intent 2 here is that we're creating a rule that would ensure 3 adequate protection and we're not seeking to do 4 anything less than that.
5 MEMBER BIER: This is Vicki Bier again.
6 I wanted to follow up on Jose's question maybe coming 7 at it from a different perspective.
8 If some of those items that are non-safety 9 related but safety significant are relied on in the 10 PRA for AOOs, et cetera, should they be subject to the 11 same QA and other requirements as safety-related 12 equipment?
13 And I realize, you know, there can be 14 some, you know, legitimate reasons for not doing so, 15 but that may be kind of the other way of looking at 16 this question.
17 MR. RECKLEY: And we'll get to that on the 18 next slide with the identification of the appropriate 19 degree of special treatment.
20 MR. HOELLMAN: Are we ready to go to the 21 next slide?
22 MR. RECKLEY: Yeah, because there's a 23 little bit of overlap, so, I mean, the --
24 MR. HOELLMAN: Yeah.
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182 1 Dr. Bier, but maybe when we talk about special 2 treatment, it will come back and support the question 3 you just had.
4 MEMBER BIER: No, that's totally fine.
5 You didn't cut me off at all, so it's fine.
6 MR. RECKLEY: Okay.
7 MR. HOELLMAN: Okay, so special treatment 8 are those requirements such as measures taken to 9 satisfy the functional design criteria, quality 10 assurance, and programmatic controls that provide 11 assurance that certain SSCs will provide defense in 12 depth or perform risk-significant functions.
13 And so for safety-related equipment, we're 14 envisioning that these would be things like your tech 15 specs and Appendix B QA requirements. For non-safety 16 related but safety significant equipment, we're 17 thinking, I guess, that these special treatments will 18 be things like the appropriate surveillances, 19 reliability assurance programs or other licensee 20 programs that would give a licensee a little more 21 flexibility.
22 MEMBER PETTI: I have a question. Can you 23 just go back to the previous slide?
24 MR. HOELLMAN: Sure.
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183 1 is the special treatment under safety related the same 2 as the special treatment under non-safety related but 3 safety significant? By the definition, it seems to 4 imply it's the same.
5 MR. HOELLMAN: Yeah, and this is a comment 6 we got from industry as well, and I think we need to 7 do a better job of clarifying this. I think, you 8 know, it depends on what the functional design 9 criteria that the SSC is required to perform sort of 10 what the special treatment would be.
11 So, like I said for safety related, I 12 think we're envisioning the safety-related stuff would 13 have tech specs and Appendix B QA requirement type 14 special treatments.
15 For the non-safety related but safety 16 significant stuff, I think we're envisioning more like 17 reliability assurance programs and other licensee 18 programs. That would be special treatments for those 19 things. You know, I think it's hard generically maybe 20 to say what would it be, but I don't want to -- go 21 ahead.
22 MEMBER PETTI: No, it just seems like you 23 need to bifurcate your special treatment. You know, 24 for safety related, it means one thing, but for safety 25 significant, it means something else.
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184 1 MR. RECKLEY: Yeah, one way to look at 2 this, Dave, is that our starting point is commercial 3 grade, and so special treatment is what is done beyond 4 commercial grade equipment, and so as Jordan said, we 5 got this comment before.
6 Special treatment then is broken down into 7 two categories. For safety related, it's the most 8 restrictive special treatment, and for non-safety 9 related but safety significant, it might be lesser 10 special treatment, but still something beyond 11 commercial grade equipment.
12 But we may have to look for a way to make 13 that distinction a little better. Often, special 14 treatment is used in normal conversation to mean 15 something between commercial grade and safety related.
16 We're using it --
17 (Simultaneous speaking.)
18 MR. RECKLEY: -- to encompass both, right.
19 MEMBER PETTI: Okay.
20 (Simultaneous speaking.)
21 MEMBER DIMITRIJEVIC: This is Vesna again.
22 This sounds really -- I mean, basically is something 23 less because we don't really care about safety-24 significant components?
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185 1 trying to merge all of the new and, I mean, old and 2 medium new with something completely new and it's just 3 not working well.
4 I mean, that's what's happening. You want 5 to have less treatment for something which is risk-6 significant? I mean, that doesn't really sound --
7 (Simultaneous speaking.)
8 MEMBER DIMITRIJEVIC: -- more treatment 9 for safety related, which is like, you know, 10 prescriptive? I mean, that --
11 MS. CUBBAGE: This is Amy Cubbage. I'd 12 just like to say that there is some precedence for 13 having, you know, a graded approach to different 14 equipment if it's not safety related.
15 We do have precedence with the regulatory 16 treatment of non-safety systems in the previously 17 certified designs. We have things like this for 18 different rules, be it ATWS, or Station Blackout, or 19 the mitigation of beyond design basis events.
20 So, this isn't a totally new concept to 21 have a grade of equipment that's less than safety 22 related but does not have regulatory treatment to 23 ensure that the appropriate availability design is 24 fulfilled.
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186 1 50.69, I mean, special treatment for this, because 2 that's added on, you know. It used to be just safety, 3 non-safety, and then this risk significance came and 4 it was added on, but now it should already be merged.
5 You know, so, I mean, you know, that 50.69 6 should already be -- we should have grown out of that 7 and merged those categories. That's easier to say 8 than to do, so I know you have a tough task, so.
9 MEMBER BROWN: This sounds vaguely like 10 what we were dealing with, Dennis, correct me if I'm 11 wrong, regulated treatment of non-safety RTNSS. Is 12 this trying to crank that concept into the Part 53?
13 MR. RECKLEY: Well, it is what Amy 14 mentioned as an example of how we've tried to do this 15 in the past. We're just trying to do it from the 16 beginning, so as -- you know, and recognize --
17 Basically we're trying to recognize that 18 the systems that are binary, which one can argue if 19 you go back far enough into the 1960s, the system was 20 binary, safety related or unregulated, that we had 21 trouble with that and that's why it evolved into being 22 basically three tiers through 50.69 or through 23 augmented quality when we brought in Station Blackout 24 and ATWS, how we've handled other things, that we 25 ended up with a middle category.
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187 1 You know, we do that now with important to 2 safety. Important to safety includes both safety 3 related and non-safety related equipment. So, what 4 we're trying to do here is again capture basically 5 that same system, but hopefully -- I'm starting to 6 question that --
7 Well, it's clear that maybe we haven't 8 been as successful as we wanted, but from the 9 beginning to set out that there were these different 10 classes of equipment and that equipment could be 11 serving a safety-significant function even though it 12 was not safety related, again as we've recognized in 13 50.69, as we've recognized in ATWS and Station 14 Blackout, and under RTNSS as you just mentioned, 15 Charlie.
16 So, and the reason that we didn't want --
17 you know, under 50.69, the one category we don't have 18 is safety related but non-safety significant, so the 19 fourth category.
20 MEMBER DIMITRIJEVIC: Yes, but I don't 21 understand how you don't have that category --
22 MR. RECKLEY: Well --
23 MEMBER DIMITRIJEVIC: -- because as you 24 said, you know, when you have -- I don't understand 25 how did you augment it because they would be so smart NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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188 1 doing the safety classification that now any non-risk 2 significant will be there.
3 I don't know understand how did you merge 4 it because that safety classification, safety and non-5 safety, and then you do risk significant 6 classification, who guarantees there will now be 7 safety equipment and non-risk significant?
8 MR. RECKLEY: Well, I think if we look at 9 how the safety classification is done and that it's 10 limited to functions that are needed to keep you below 11 the dose threshold defined in 53.210(b) as mentioned 12 there, the 25 rem at the exclusionary boundary, that 13 is going to exclude things being termed safety related 14 that aren't risk significant.
15 So, I just, I don't know that it's -- how 16 to explain it any better than that, but the way it's 17 being done, I think, more or less guarantees that you 18 won't have safety-related equipment that's not serving 19 a risk-significant function.
20 MEMBER BROWN: Bill, relative to your 21 categories up in the place where we start doing the 22 safety in Part C, or B, or whatever it is, the safety 23 criteria which deal with site boundaries and radiation 24 dose, et cetera, at least as they're written, but how 25 does that --
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189 1 I'm trying to categorize that in terms of 2 like the reactor trip system we have. Is that a 3 safety-related system or not?
4 MR. RECKLEY: It will most likely be 5 safety related, and for every -- the reason I caveat 6 that is for any reactor that we've seen, and up to 7 this point, we've looked at reactors that were in the 8 hundreds of megawatts, and so the inventory and the 9 resulting source term on a release had the potential 10 to exceed 25 rem.
11 And so those reactors that we've looked at 12 have all had safety-related protection systems because 13 the heat removal systems needed to be sized at a 14 reasonable point, which required that the reactor be 15 able to be shut down, so they all have required a 16 safety-related shutdown system.
17 The reason I caveat is -- and again, we 18 are just now starting to look, but there are people 19 who say that some of the reactors on the microreactor 20 side that have small inventories, relatively small 21 inventories --
22 This is yet to be proven, so, you know, 23 I'm a doubting Thomas myself, but the inventories are 24 small, that the reactivity feedback mechanisms might 25 be able to limit the power levels and that they may, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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190 1 some claim, be able to show that you don't need a 2 protection system to keep you less than 25 rem, which 3 under our criteria would be safety related.
4 Now, if that were to come to pass, the 5 shutdown systems are still going to be safety 6 significant. They're still going to be risk 7 significant in the PRA, and so they would require 8 special treatment, this is my guess, but they may not 9 be safety related if the inventory and the inherent 10 nature of the reactor is able to keep the dose less 11 than 25 rem at the boundary.
12 So, this is yet to be shown, but we're 13 trying -- just like this morning when we were talking 14 about the operations. We haven't reached any 15 conclusions yet, but we're trying to build this that 16 if someone can design that machine, the rule could 17 support it, not predicting that it can be done, but 18 some claim that it can.
19 CHAIR BLEY: Bill, this is Dennis. I want 20 to jump in with a couple of things. One is I think 21 you're pretty close to what you need and I think you 22 got to work on a little bit of the language as this 23 has come up, but these more recent issues I think are 24 well covered and we'll dig into them more deeply.
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191 1 ought to have a presentation in the not too distant 2 future on at least some parts of the LMP, and there 3 I'm getting it after you've got your billions of 4 billions of scenarios in the PRA, how you pick out the 5 licensing basis events and how you get to the design 6 basis accidents and AOOs, and how it's projected you 7 actually analyze them for the safety analysis.
8 And I think that's a presentation we need 9 because it's very -- well, a lot of us were not on the 10 committee at the time that was talked about, certainly 11 back when 1860 was done, and even for the NGNP white 12 papers, and then later for the LMP, so I think that 13 would be useful.
14 (Simultaneous speaking.)
15 MR. RECKLEY: Okay, we'll coordinate with 16 Derek.
17 MEMBER BROWN: Bill, Dennis, my basic 18 quandary is that everything is getting so fuzzy. I 19 mean, it sounds like we don't really care whether the 20 reactors shut down or not as long as we meet those 21 limits. I mean, that, if I was the public, that's 22 what I would hear.
23 CHAIR BLEY: When you brought up the 24 RTNSS, the RTNSS wasn't easy restrictions. RTNSS came 25 up because the way we've always done this, there was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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192 1 quite a bit of equipment in some of the passive plants 2 that, although they wouldn't meet the definition of 3 safety grade or safety related, once the clock ticked 4 out for three or four days, you really needed them, so 5 we didn't want to ignore them.
6 And the same thing as Bill said came up 7 back under Station Blackout and under ATWS, they were 8 outside of the design basis accident space and still 9 it looked very wise to have some controls, and so it 10 was actually expanding the areas of controls.
11 And now they're trying to formalize that 12 in a coherent way, and I think it keeps what you 13 wanted from the past, but it also includes what's come 14 up in the last ten, 15 years.
15 MEMBER BROWN: Dennis, when I got here, 16 the RTNSS was -- I got here in 2008, but I didn't have 17 any past association. So, I know RTNSS just popped up 18 on the radar when I got here on a couple of occasions 19 and I had it explained to me.
20 I just -- I don't disagree with this 21 safety significant category. It's just that when I 22 listen to some of the other explanations about nothing 23 matters except dose at the boundary and we can have 24 reactors perk along, it just seems like slippery 25 slopes are getting more prevalent clear across the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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193 1 horizon and that's all. That's just a concern. We 2 ought to go on, at least from my perspective, with the 3 presentation, I guess.
4 (Simultaneous speaking.)
5 MEMBER MARCH-LEUBA: I'd like to support 6 what Charlie just said. I 100 percent support that.
7 I mean, this is a slippery slope. And please, Bill, 8 for the next presentation, show me an example.
9 Prepare it where a protection system is 10 not safety grade and it becomes safety significant 11 because if it cannot possibly produce 25 rem on the 12 boundary, how can it be risk significant?
13 So, if it's not safety grade, it's not 14 risk significant, and therefore the protection system 15 is completely up to you to whatever you want to 16 design. We're not going to look at it.
17 So, please think it through and provide us 18 an example how a protection system that is not needed 19 for safety related or what I call safety grade will 20 become safety significant. I just don't see it.
21 Thank you.
22 MR. HOELLMAN: Okay, Bill, do you want me 23 to keep going?
24 MR. RECKLEY: Please, Jordan.
25 MR. HOELLMAN: Okay, so --
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194 1 MEMBER BROWN: As fast as you can.
2 MR. HOELLMAN: All right, all right, I 3 asked before I started how much time. I didn't ask 4 you guys, but I asked internally how much time we 5 should spend, so I'll try to run through it quickly.
6 So, we had a little discussion of defense 7 in depth earlier this morning, and so here is the 8 definition we have included in Part 53, the inclusion 9 of multiple independent and redundant layers of 10 defense in the design of a facility and its operating 11 procedures to compensate for potential human and 12 mechanical failures so that no single layer of 13 defense, no matter how robust, is exclusively relied 14 upon.
15 And so we've got some stakeholder feedback 16 on this. They generally are supportive of the 17 definition here and asked for this definition to match 18 the requirement as it's described in Subpart B.
19 So, move onto the next slide and we'll 20 talk about design features, inherent characteristics, 21 and functional design criteria.
22 So, for design features, these are 23 essentially all SSCs, the active and passive structure 24 systems and components and inherent characteristics 25 that contribute to eliminate the total effective dose NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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195 1 equivalence to individual members of the public during 2 normal operations and prevent or mitigate consequences 3 of unplanned events.
4 Inherent characteristic, this is what we 5 discussed earlier today, and our definition in Part 6 53, at least in the initial iteration here, we're 7 defining it as an attribute of a design feature that 8 has such a high degree of certainty and performance 9 that the uncertainties need not be quantified.
10 So, it's, I think, similar to what Jesse 11 presented earlier with the, you know, absolutely 12 reliability, but in this case, we're defining it as 13 the uncertainties would not need to be quantified for 14 these.
15 And so we wanted to define this because in 16 our second iteration of Subpart B, we revised language 17 in the defense in depth section that no single 18 engineered design feature, human action, and/or 19 program control, no matter how robust, should be 20 exclusively relied upon to meet the first tier safety 21 criteria for unplanned events.
22 So, then we'll talk about functional 23 design criteria, and I'm not sure if this is -- maybe 24 this will help Jose in your question to some extent.
25 So, functional design criteria are the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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196 1 requirements for the performance of the SSC, so what 2 design criteria such as a leak grade or cooling 3 capacity are needed for each design feature.
4 So, for safety-related SSCs, these 5 criteria define the requirements necessary to 6 demonstrate compliance with the first tier safety 7 criteria, which is the first tier safety criteria for 8 unplanned events.
9 So, that's your immediate threat to public 10 health and safety, the two-hour dose below 25 rem at 11 the EAB or the duration dose below 25 rem at the low 12 population zone boundary.
13 For non-safety related but safety 14 significant SSCs, these criteria define the 15 requirements necessary to meet the second tier safety 16 criteria for unplanned events.
17 So, these are -- that's the criteria 18 appropriate to address potential risks to public 19 health and safety, and at this time, we're using the 20 QHOs for that criteria, which would be supported by a 21 systematic analysis and would enable a risk management 22 approach to operations.
23 So, I think, I guess the confusion around 24 the two tiers is maybe leading to more confusion with 25 how the SSCs are defined and the event categories to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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197 1 some extent.
2 So, in following Member Brown's direction, 3 let's keep it going.
4 MEMBER REMPE: Well, just a second. This 5 is --
6 MEMBER MARCH-LEUBA: Oh, no, no.
7 MR. HOELLMAN: Okay, let's --
8 (Simultaneous speaking.)
9 MEMBER MARCH-LEUBA: Go ahead, Joy.
10 MEMBER REMPE: This is Joy and I 11 apologize. I had to join this meeting late, but I 12 sent some comments to Dennis and Matt, and did they --
13 could you -- did you already discuss about the staff's 14 current thoughts about having the two tiers? Are you 15 still wanting to go with those two tiers or are you 16 having second thoughts on it?
17 MR. HOELLMAN: So, we're still 18 considering, you know, what to do with that comment.
19 You know, I think we've heard it multiple times in 20 both our public meetings and our ACRS subcommittee 21 meetings, so I don't think I have a better --
22 (Simultaneous speaking.)
23 MR. HOELLMAN: I don't think I have a 24 better answer for you, Joy, than that unfortunately at 25 this time, but --
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198 1 MEMBER REMPE: That's fine. I just was 2 curious because -- I apologize if you guys already 3 discussed this.
4 MR. WIDMAYER: Hey, Dennis? Excuse me for 5 interrupting. This is Derek. We've lost the public 6 line. Yeah, maybe we should take a break and see if 7 we can get it back.
8 CHAIR BLEY: I wanted to get through these 9 last two slides, but okay. It's now almost 1:30 here, 10 so it's 3:30 there. Let's come back at -- oh, this 11 will be our last break, so take 20 minutes. Try to 12 get the phone line back. I don't know. We will start 13 at quarter 'til.
14 MR. WIDMAYER: Okay, thanks.
15 CHAIR BLEY: We're in recess, and God, I 16 wish we got through these, but we'll finish the last 17 two slides there when we get back and then get to the 18 hard one.
19 MR. WIDMAYER: Okay.
20 (Whereupon, the above-entitled matter went 21 off the record at 3:26 p.m. and resumed at 3:45 p.m.)
22 CHAIR BLEY: Well, we're back in session.
23 It's a quarter 'til the hour. And we have a question 24 from Jose March-Leuba that we didn't get to so we're 25 going to start with that. Jose?
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199 1 MEMBER MARCH-LEUBA: Yes. Back on the 2 Slide 51, which we have on the screen, with respect to 3 the absolute reliability of inherent staff, you 4 mention here in the definition a high degree of 5 certainty.
6 Nothing is inherent in this life. I mean, 7 you can think of gravity, not including the wording 8 here on gravity, but you can overcome it with 9 centrifugal force. So even gravity is not inherent.
10 So I think in the definition you need to 11 do something similar which you do with the definition 12 section which defines a frequency above which you 13 don't have to qualify uncertainties and below which 14 you already are aware of you have to.
15 So define, I mean, inherent is really a 16 passive feature that has a frequency or a probability 17 of failure, which is lower than the 10 to the minus 4, 18 10 to the minus 6, 10 to the minus 12. You pick the 19 number.
20 So the way you have it here, I'm an 21 applicant. I can claim my feature is inherent, and I 22 submit an application and now we start fighting. You 23 should have an absolute kind of frequency that you 24 need to demonstrate. Okay? Keep going.
25 MR. HOELLMAN: Okay. Good thought there, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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200 1 Jose. We'll consider it. And like I mentioned 2 before, we know that some of these things are -- you 3 know, it's iterative. And we know we need to make 4 changes as we continue to revise the language.
5 So I'm ready to move on to the next slide 6 if there is --
7 MEMBER DIMITRIJEVIC: There is --
8 everybody -- actually, this is Vesna. I went and 9 checked some data on one of the application on 5069 in 10 very advance light water reactor. It's a public 11 information. One of the plants which have applied 12 50.69.
13 If you go back to the slide with 14 classification, because in the discussion, we said 15 there would not be several related components which 16 are non-significant.
17 So this is advanced light water reactor 18 with three things. They have applied 50.69. So they 19 have all four categories.
20 So the category which we said will not 21 exist anymore had about 15,000 of SSCs. So they have 22 deep classification for about 75,000 SSCs.
23 So category which was safety-related but 24 not really significant had about 15,000 SSCs. And 25 that one disappear and sort of merge here. The NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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201 1 category which we are still keeping non-safety 2 related, safety significant had only 500 components, 3 less than 1 percent. That's a very small percentage.
4 So my question is if we already absorb a 5 category which we know is much bigger in existing 6 light water reactor, why cannot we absorb this small 7 category into the safety related?
8 That's all my comment is based on the data 9 existing today in very limited 50.69 application.
10 Okay? I just wanted to bring this data and say that 11 this category is very small existing light water 12 reactors.
13 CHAIR BLEY: Thanks, Vesna. You can go 14 ahead now.
15 MR. HOELLMAN: Okay. I think we're on 52, 16 52? Okay. So this is essentially the end of the 17 presentation here. This talks about the other 18 requirements in Subpart A.
19 Like I mentioned at the beginning, it 20 reproduces Part 50 in a lot of places and currently 21 includes the bracket references to the existing 22 requirements. And those brackets will be replaced 23 with the applicable Part 53 requirements once 24 developed.
25 We're intending to develop Part 53 with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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202 1 largely no cross-references to Parts 50 or 52. And 2 Subpart A will include many pointers to other sections 3 of Part 53 that will need updating in future 4 iterations and that includes things like the emergency 5 plan submittals and security plan submittals.
6 So Slide 53. So this is just the general 7 feedback we've received from our public meeting on May 8 6. Like we mentioned earlier, the definition of 9 advanced nuclear plant, the feedback we received is to 10 allow broad applicability for advance reactors under 11 Part 53 and that the technical requirements should 12 generally define or determine the applicability of 13 designs that can apply under Part 53.
14 We had this discussion under the design 15 basis accident today and said we'll be working to 16 clarify the relationship between the DBA and unlikely 17 event sequences. And very unlikely event sequences 18 that under consensus cares and standards, the 19 stakeholders questioned the use of must in the 20 definition. And this is related to -- the definition 21 currently says it must -- the current standard must be 22 designated for nuclear power plants. And the comment 23 was sort of what about other industries?
24 Under defense in depth, the comment we've 25 gotten was to update the Subpart B to match the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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203 1 definition in Subpart A because the definition 2 excludes the clause that a new single feature should 3 be exclusively relied upon.
4 The safe stable end state, this is sort of 5 related to the comments some of the ACRS members 6 provided related to, you know, maintaining the end 7 state in a subcritical condition but also to include 8 the words safe and stable in the definition.
9 And then as we discussed and had some 10 significant discussions, I guess, earlier this 11 afternoon, the terms for the SSC classifications is 12 going to require some work and clarifications as we 13 move forward with the rule.
14 So with that, I think we spent longer than 15 I was intending to spend in this section. So if 16 there's any further questions, we're happy to take 17 them or if we want to just continue moving into the 18 Subpart F discussions, we can do that as well.
19 MEMBER MARCH-LEUBA: I'd like to make a 20 comment because I'm glad you're taking up our 21 suggestions of the safe, subcritical end state.
22 I just want you to consider the 23 ramifications for those reactors that say we can 24 overheat my reactor, and we will eventually reach an 25 equilibrium. Depending on how you do it, you must NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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204 1 still be a k factor equal to 1. So when we say 2 subcritical, we have to make sure we know we mean 3 subcritical or a k factor of 1 regulated at the low 4 power is acceptable.
5 So think about the consequences of what 6 you do and it's because of our fault but. I'm just 7 giving you -- some of these reactors when they get 8 hot, they have feedback. But they eventually have to 9 run on certain power to maintain that temperature.
10 Okay. Thank you.
11 MR. HOELLMAN: Thanks, Jose.
12 MEMBER BROWN: This is Charlie Brown. I'm 13 going to add just a thought process. I've gone back 14 and looked at the first tier and second tier safety 15 criteria in Part D, I think.
16 And everything focuses on radiation and 17 dose, 10,000 years, 25 miles -- I mean, site 18 boundaries, 25 rem, et cetera, and even the second 19 tier is more vague.
20 I really think we've missed the boat, this 21 is my opinion only. I'm throwing it out. Isn't one 22 of the first tier safety criteria that if it should 23 fall under the safety-related equipment thought 24 process is that any -- you should be able to maintain 25 the reactor subcritical and shut down throughout the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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205 1 temperature range.
2 And the second one is you should be able 3 to remove heat and cool the plant down. And those, to 4 me, in addition to the radiation requirements would 5 effectively establish the top tier level of safety 6 criteria regardless of how you categorize the 7 evidence.
8 I think we missed the boat by focusing 9 only on radiation and letting the plant sit around and 10 perk either critical or bubble around at 300, 400, 500 11 degrees. I think that's a mistake. But that's my 12 thought process is that ought to be thrown into the 13 hopper for consideration of re-evaluating of what we 14 consider as the first tier of safety criteria.
15 MEMBER PETTI: Charlie, the only concern 16 I have is that molten salt reactors have to stay 17 hotter than 300 degrees or they'll freeze their 18 coolant.
19 MEMBER BROWN: That's a good reason for 20 not ever using them. I've been there and done that, 21 okay, after the sodium plant that the Navy had.
22 MEMBER PETTI: That's a different 23 question. But there's some practical issues.
24 MEMBER BROWN: That's a different -- if 25 that's the point, you can always work an exemption if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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206 1 you want to deal with that. But my point is basic 2 plants ought not be hot, and they ought not be 3 critical.
4 So if there's an exception to that along 5 the way for one of the plants, fine, you deal with it.
6 But you don't assume that and let it run everything 7 else. That's the end of my comment on that as if I 8 won't say it again sometime.
9 MR. HOELLMAN: Thanks, Member Brown. So 10 if we're ready to move on, I guess we can do that.
11 And I think I'm turning it back to Bill.
12 MR. RECKLEY: Okay. Liz, do we want to 13 you, yes. And go on to 57. So we're going to shift 14 gears and talk about Subpart F. Obviously, all these 15 things are interrelated so much of the discussion we 16 had under definitions comes out of the discussions we 17 had under B and C, Safety Objectives and Design and 18 Analysis. So many of those issues will also come up 19 as we talk about the requirements that are being 20 proposed for operations.
21 So that is Subpart F. Looking at the life 22 cycle, this would be the subpart that would address 23 the operations phase. Yes, go ahead, Liz, so 58.
24 So the first section within Subpart F is 25 53.700. And it just lays out the objectives of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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207 1 subpart and lays out the structure that I mentioned 2 earlier that things are generally laid out with the 3 first sections addressing hardware or structure 4 systems and components in terms of maintaining their 5 capabilities, reliabilities, configuration control, 6 those kind of things.
7 The second group of sections within 8 Subpart F would go to personnel. And that was this 9 morning's discussion. We're just now beginning to 10 develop the rule text.
11 And we'll bring those back to this 12 subcommittee in the next few months. But you heard 13 some of the challenges this morning as to trying to 14 provide a wide range of possible requirements and 15 plant designs and the role of personnel within those 16 designs. But we'll be back with that. We're not 17 going to talk about that today. Like I say, we're 18 just beginning to draft those sections on personnel.
19 And then the last series of sections 20 within this subpart talk about programs. And so we'll 21 talk about that today. So we're going to talk about 22 hardware and programs, and we'll be back on personnel.
23 So that's Section 53.700, just kind of the 24 objectives, basically saying licensees have to have 25 measures in place to address all three of these NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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208 1 things, the care and maintenance of equipment, 2 requirements on personnel and the development, 3 implementation and maintenance of programs.
4 So if we go on to Slide 60, a requirement 5 that we've included in this preliminary language is to 6 have a transition into operations from construction 7 and manufacturing so that was Subpart E if you 8 remember.
9 This is often taken into account in 10 startup test programs in Chapter 13 of an FSAR for 11 example. And it may be one of the comments we added 12 in the discussion table is this could be considered a 13 program. And we'll talk to stakeholders as to whether 14 it would be more appropriate to have within the 15 programs, but I don't think there was much 16 disagreement on the need to have such a transition 17 plan. It was just how it would be treated within the 18 regulations.
19 But our thinking in including it in this 20 preliminary language was, again, we've constructed 21 Part 53 as a life cycle kind of format and so the idea 22 of having a requirement to transition from one part of 23 the life cycle into the next seemed to make some 24 sense.
25 CHAIR BLEY: Mr. Reckley?
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209 1 MR. RECKLEY: Yes.
2 CHAIR BLEY: Joy Rempe had some questions 3 about this. Joy, are you actually here?
4 MEMBER REMPE: I'll be here the rest of 5 the day. I just was curious, but I almost think you 6 kind of answered it. But when I first saw the slide, 7 I was curious how this is done for Part 50.
8 And I think your answer is it's done, but 9 it's not done in a requirement because you're doing 10 this because you're doing the whole life cycle in Part 11 53. The fact that you're now requiring the programs, 12 procedures and controls are going to be, I guess, 13 submitted.
14 And then I guess a related kind of 15 question that I had mentioned earlier and, again, I'm 16 not sure what Dennis and Matt mentioned earlier. But 17 I still think a diagram to show everything that's 18 going to be needed for Part 53 would be helpful and 19 compared with what's required in Part 52 and Part 50 20 because there are things like this that I don't think 21 you see as part of the regulation. What are your 22 thoughts on that?
23 MR. RECKLEY: It was a good suggestion.
24 And as we have time and as we develop the statement of 25 considerations, I think a tool like that is going to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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210 1 be useful.
2 Some of this does exist when you look at 3 regulations within Part 52 for example. You'll see 4 provisions for when things need to be done, you know, 5 in large part. Even the 103(g) finding has some 6 components of a transition from construction to 7 operations.
8 So these things do exist. They are mixed 9 between the license and actual requirements in either 10 Parts 50 or 52. But, again, we're open. I don't 11 think anyone is arguing. The concern from 12 stakeholders was is this yet another document to be 13 prepared and then reviewed and so it ends up actually 14 maybe accomplishing the same thing but being more 15 formal and therefore more expensive?
16 So that's one of the things we'll look at.
17 Again, no one is really arguing that the transition 18 plan needs to exist. It's just the format and so 19 forth so.
20 MEMBER REMPE: It will be a document that 21 will need to be submitted and reviewed. I'm not 22 necessarily thinking that's a bad idea. But I think 23 they've got a point, right? I mean, it never was 24 something that was submitted and is on docket was it?
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211 1 parts.
2 MEMBER REMPE: Yes.
3 MR. RECKLEY: And throughout a whole bunch 4 of different programs, right? And so that's the 5 difference. And the concern is if we're going to do 6 it here, then do it here. But if we're going to do it 7 here and in addition it needs to be done in a whole 8 bunch of other places, then it becomes repetitious.
9 So, again, we're looking at it and we're 10 going to talk to stakeholders about -- again, the idea 11 is not controversial. It's just how it's carried out.
12 MR. HOELLMAN: It's really like the PSDAR 13 on the back end of the cycle.
14 MR. RECKLEY: Yes. And, again, people 15 have such programs. It's just how they're documented 16 and whether they're siloed within a bunch of different 17 programs and then collectively they add up to an 18 overall transition plan and so forth.
19 So, again, we'll look. And I only bring 20 up that this might be an area where stakeholders come 21 back with an alternative. To the degree it's 22 equivalent, we would be amenable to that so.
23 So if we want to go on then to 73.220, 24 this is the major section or there's two or three 25 major sections related to equipment within this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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212 1 subpart. And so 73.220 maintaining capabilities is 2 really the focus in terms of making sure that the 3 structure systems and components are going to perform 4 during operations.
5 So basically then 220 lays out that the 6 role of the section is to do exactly that, to make 7 sure the capability and the reliability of the 8 equipment is consistent with what was assumed in the 9 analyses and so forth to show that you meet the safety 10 criteria in Subpart B, 53.210(b) and 220(b). So 11 that's the unplanned events, the design basis accident 12 and the other licensing basis events.
13 The section is broken down into Paragraph 14 (a) and Paragraph (b) with Paragraph (a) being for 15 safety related. And as we will talk about on the next 16 slide, that's primarily technical specifications.
17 And Paragraph (b) that defines the 18 controls for non-safety related but safety significant 19 SSCs. And that's done through reliability assurance 20 programs and other special treatments.
21 So if we go on to Slide 61, this lays out 22 this, section lays out, the requirements for technical 23 specifications. They are similar but with some 24 differences from the current requirements in 50.36 for 25 tech specs. So, again, it's tied back to the safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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213 1 functions.
2 And the first tier safety criteria, 3 53.210(b), and as was mentioned a couple times 4 already, that is those required safety functions to 5 keep the dose of the exclusionary boundary less than 6 25 rem.
7 It lays out some things like the inventory 8 of radioactive materials because that's important to 9 ensure that the ultimate source term and the release 10 is as was modeled in the analyses. It's also a 11 provision that's specifically called out in the Atomic 12 Energy Act.
13 And then operating limits, and operating 14 limits would be those things, again, needed to set out 15 the overall requirements on the plant to make sure 16 that should an unplanned event occur that the releases 17 would be constrained as modeled.
18 So this would include traditional things 19 like power levels. For gas cooled reactors, we think 20 this would include the SARDO limits, the limits on 21 radionuclides. Actually even for a light water 22 reactor, it might include the activity limits within 23 the primary system to make sure that any leak or tube 24 rupture or whatever would be below the limits.
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214 1 as safety related, the requirement would be that 2 limiting conditions for operation and surveillance 3 requirements would be defined. These are pretty 4 consistent with the current use of those terms and 5 requirements as they would be defined under 50.36 for 6 the current light water reactors.
7 And then the rest of the section, or the 8 rest of the tech specs would address design 9 attributes. That, again is something similar to the 10 current 50.36 requirements to establish the things 11 like geometries or whatever that aren't expected to 12 change but are critical to meeting the requirements.
13 Administrative controls, again very 14 similar to the existing requirements that would 15 include various programs, reporting requirements, et 16 cetera. And then when appropriate a decommissioning 17 section, again, consistent with the current 18 requirements.
19 Now of note our first iteration did not 20 include two things from 50.36 and that is the concept 21 of safety limits or limiting safety system settings.
22 That notion of not including those is those concepts 23 were derived and largely developed for the light water 24 reactors.
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215 1 the equivalent of those for other technologies, it's 2 in my view led to more confusion and it's kind of 3 artificial. And that's usually because the safety 4 limits for light water reactors and then the 5 associated limiting safety system settings is based on 6 the existence in light water reactors of actual 7 phenomena like departure from nucleate boiling where 8 the pressure limit on a pipe that can lead to a quick 9 failure of a barrier.
10 And in some designs, those phenomena where 11 you get a step jump in behavior don't exist. You get 12 degradation. Ultimately, you would get a failure.
13 But you don't have these phenomenologically driven 14 step changes.
15 And so we think we can address that in 16 other sections of the tech specs without bringing in 17 the concepts of safety limits or associated limiting 18 safety settings, which, again, looking back on history 19 when we tried to do that for some reactor 20 technologies, it was somewhat artificial and didn't 21 really -- it didn't really work, and it didn't serve 22 the same purpose as it did for light water reactors.
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216 1 limits, LCOs or maybe even design attributes. So it's 2 not as if they're going to -- that some critical 3 parameter is not going to be addressed. It just won't 4 be addressed exactly the same way it is under 50.36.
5 The other thing that deviates from 50.36 6 is the existence of the four criteria in 50.36 for 7 what needs to be included as an LCO. And the logic 8 there is that basically we are defining that LCOs need 9 to be provided for any safety-related SSC.
10 That generally aligns with the first three 11 criteria in 50.36. And the remaining criterion in 12 50.36 was the risk significant criterion that was 13 added the last time 50.36 was updated in 1990s, I 14 guess.
15 And that was added to address, as we have 16 talked about before, some of the lessons learned from 17 PRAs and the fact that there were systems like core 18 isolation cooling, RCSI and BWRs and maybe -- that's 19 the one that comes to mind that was a risk significant 20 system that wasn't addressed in tech specs or need not 21 at the time, may not have been addressed for some 22 plants in tech specs.
23 The logic for not including it in Part 53 24 and under this section is that those systems will be 25 picked up under Paragraph (b). The risk significant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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217 1 functions will have special treatment. They'll just 2 be special treatment out of tech specs.
3 And so this is one of those things that 4 might be subtle. But it's what we're trying to 5 accomplish by having the two classes -- our 6 preliminary language has called them tiers and that's 7 caused some confusion. But two groupings of criteria 8 and then trying to carry them through all of the 9 subparts.
10 So once you take what we're currently 11 calling the second tier, that's going to be addressed 12 by Paragraph (b). The reason that it needed to be 13 picked up in 50.36 in the 1990s was because we didn't 14 have that in Part 50 at the time, keeping in mind this 15 was before 50.69 and so forth.
16 So unless we had otherwise required it by 17 some rulemaking, then it wasn't going to be included.
18 And the fact is when we did the rulemakings for things 19 like station blackout in ATWS, we did not typically 20 say that that needed to be in tech specs as part of 21 the rulemaking.
22 So that is the difference in the tech 23 specs in Paragraph (a) for plant equipment. Then if 24 we go to Paragraph B on Slide 62 --
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218 1 minute?
2 MR. RECKLEY: Sure.
3 MEMBER KIRCHNER: On 61, Bill -- this is 4 Walt Kirchner. So, yes, I appreciate the dilemma of 5 -- and perhaps the fact that creating safety limits in 6 tech specs is something of an artificiality, but many 7 of these advanced reactors have thresholds, 8 bifurcation points, not too unlike a large LWR.
9 And so my sense was always that the 10 different way of thinking about defense in depth is 11 you established operating limits so you did not 12 challenge your safety systems or your protection 13 systems in particular unduly.
14 You, therefore, put perhaps in some cases 15 an overly conservative cushion into your plant 16 operating envelope, but it served a useful function.
17 And I would submit that I could give you a handful of 18 examples of places in advance reactor designs where 19 you will have similar thresholds or non-linearities 20 where you cross a certain boundary and you're in big 21 trouble, whether it's boiling in a sodium cooled fast 22 reactor or whether it's not being able to control the 23 chemistry of a molten salt solution or even more 24 complicated a molten fuel kind of reactor.
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219 1 exist with these advanced designs where that operating 2 limit, safety limit, provides, again, as I said, 3 cushion is not a good word. But it provides some 4 margin in terms of time of response or not challenging 5 your safety systems too frequently and et cetera.
6 So it has provided, I think, a useful 7 function for the current fleet. And I'm just trying 8 to mentally, quickly think through why you wouldn't 9 want a similar kind of defense in depth or cushion for 10 advanced reactors, notwithstanding all the advertised 11 wonderful, inherent and passive safety features.
12 If you push hard, you will find that these 13 advance reactors have similar threshold issues. It 14 was more an observation than a question.
15 MR. RECKLEY: And that's fine. And I 16 agree with you. And sodium boiling or time and 17 temperature for TRISO fuel, those things have been 18 brought up in the past in earlier attempts to do this 19 as potential safety limits.
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220 1 and the need to trip the plant very quickly.
2 And so I agree with you, the sodium 3 boiling or time and temperature limits are likely even 4 to be used for those associated technologies. But 5 those parameters, at least in what I've seen, are not 6 challenged in the anticipated operational occurrence 7 realm. They're challenged more in the lower frequency 8 events, in what we're calling the unlikely or very 9 unlikely event categories.
10 And then another thought that we had as we 11 were looking at this is over time in terms of the 12 associated limiting safety system settings, if you 13 look at the history of the tech specs, and that's what 14 we're talking about here, what would be in the tech 15 specs, the history of what is included in tech specs 16 under the limiting safety system settings has changed 17 a lot over the decades.
18 And now with the response times and even 19 the set points being in licensee controlled documents, 20 the limiting safety system settings requirement is 21 basically to have them and the details are elsewhere.
22 And so, again, we just kind of looked at 23 the big picture and thought all of the things you 24 mentioned I am confident will be caught somewhere.
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221 1 safety limits, the reporting of an exceedance of a 2 safety limit and the connection of the safety limit to 3 the safety system settings, it was not clear we needed 4 to continue that in light of -- like I said, the 5 changes that have been made for light water reactors 6 and somewhat the artificial -- it's not artificial.
7 It's not the same in terms of the need for a quick 8 response as it has been for light water reactors and 9 the definition of those safety limits like D&B and 10 pressure settings.
11 So, anyway, I understand your point and as 12 we go through this and look at it, I think certainly 13 there will be more opportunities once the whole 14 picture gets together to revisit how we're doing this.
15 So any other questions on the tech specs?
16 Okay. Liz, if we go to 62. This becomes one of the 17 important differences in my mind for what we're 18 proposing under Part 53.
19 And it goes to a lot of the discussion 20 that we even had a little while ago. This binary 21 notion, we're trying to get away from the binary 22 notion of safety related or useless. There's a whole 23 category of events, of SSC's, I'm sorry, that are 24 captured in the middle and we think provide a valuable 25 way to rebalance the overall program.
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222 1 So this is what 50.69 was aimed to do to 2 some degree for the operating fleet. To understand 3 that what started out and reflected the history 4 reflected a binary logic, resulted in safety-related 5 equipment that was not safety significant.
6 And so what we're trying to do is to 7 acknowledge that there is this safety significant 8 category. And actually if it's considered in the 9 design, we think the numbers that were given earlier 10 changed dramatically and that there will be more 11 safety significant SSCs then there will be safety-12 related SSCs.
13 And so that is one way to try to focus 14 safety and make sure that the structure systems and 15 components needed to prevent an immediate threat to 16 public health and safety are constantly maintained.
17 That's what we tried to do under the first tier. And 18 that equipment is controlled by tech specs, and it 19 would be rigid prescriptive kind of controls.
20 But then once you go down into a broader 21 risk management concept beyond an immediate threat to 22 people and, you know, we're talking about what the 23 appropriate criteria for the second tier would be.
24 We're currently using the QHOs.
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223 1 under a risk management approach more of these SSE's 2 can fall under a licensee program. But the 3 regulations still exist. It gives more flexibility.
4 But it's not all or nothing logic. It is you now have 5 regulatory controls on a large set of equipment that 6 are contributing to the safety case.
7 And they will require special treatment, 8 meaning they're going to require something beyond just 9 putting them in and hoping they work. They're going 10 to take surveillances. They're going to take -- maybe 11 for a station blackout we had augmented quality, not 12 necessarily full Appendix B but something beyond 13 commercial grade.
14 They may require looking at the 15 environment that they're expected to operate under and 16 therefore a special treatment would be assigned to say 17 I can't just buy off any shelf. I need to ensure that 18 it will operate under a certain environmental 19 condition.
20 And so all of that would need to be done 21 to show that they maintain the capabilities, the 22 availabilities and the reliabilities consistent with 23 the design activities and the analysis activities I've 24 done to show that they meet the second tier safety 25 criteria.
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224 1 I think that should be 220(b). I have a 2 typo there in the slide.
3 So then it goes on to say the licensee 4 would have to have controls to say who has the 5 authority over the configuration of that equipment.
6 And they would have to have the program in place to 7 say what special treatments are needed, again to 8 reflect the analysis, and how they would be 9 maintained.
10 You know, this is a step in our view to a 11 more performance based system given the licensee has 12 a little more latitude but ensuring the reliability of 13 the equipment.
14 We spoke, I think, last time under Subpart 15 C that this is also one of the things that's needed 16 for the previous Commission decisions on not requiring 17 the single failure criterion. That if you're going to 18 use a probabilistic approach to looking at a wide 19 combination of equipment failures and then using that 20 probabilistic analysis to justify whereas an 21 alternative to the single failure criterion, then you 22 need to maintain those reliabilities. And so this is 23 the way to come back in and somewhat more so -- well, 24 definitely more than what is done now to maintain the 25 reliability of the equipment to make sure it matches NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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225 1 up with the assumptions in the PRA.
2 That's done now. But given the PRA 3 generally isn't supporting the licensing case as it is 4 in this under Part 53, there is not as strong a 5 connection to the reliability assurance programs as 6 there would be under Part 53.
7 That's going to put it in as a -- well, 8 our preliminary proposal to put it in as a specific 9 requirement as we're doing here under 720(b).
10 MEMBER HALNON: Hey, Bill. This is Greg.
11 Just a thought going forward on this. The programs 12 and the guidance is going to be probably more detailed 13 than obviously this is going to be. And we'll 14 probably be needing to look at that as well as a very 15 focused industry interaction with the guidance that 16 will come out for this.
17 But I do see down the road a collision 18 with NPO and what they might require. So just a 19 thought going forward is to one, take lessons learned 20 out of the issues we've had within the recommendations 21 and other NPO programs colliding with the service time 22 and all this other stuff that occurred over the last 23 three, four, five years in the industry and make sure 24 that we at least don't conflict the regulatory 25 guidance or regulatory language with potential NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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226 1 guidance that we need to come out in the future that 2 is consistent with what may be required inside the 3 industry NPO requirements.
4 You know, I don't know what NPO is doing 5 about advanced reactors and whatnot, when they're 6 going to get engaged or if they are engaged, but 7 certainly they will be interested in this piece of it.
8 MR. RECKLEY: Okay. Yes, thank you. And, 9 again, this whole notion of going to a more risk 10 management approach does come closer. I think we can 11 maintain the separation because we would still be 12 saying you need to control the reliability to the PRA 13 assumptions because those were used in your safety 14 case versus ultimately NPO may be going even beyond 15 that in terms of what might be possible or what might 16 be desirable from beyond the safety case discussion.
17 But it does move us closer together I would admit. So 18 thank you very much.
19 So with that, I think we can -- if there's 20 no questions on the special treatment of non-safety 21 related equipment, we can go on to Slide 63. Section 22 73.730, the preliminary language, looks very similar 23 to the maintenance rule that's currently in Part 50.
24 The notion was you need to have such a similar program 25 for structure systems and components.
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227 1 Maybe one difference was because 2 maintenance -- this is a history of Part 50 and its 3 evolution, right, that as rules were introduced there 4 were decisions that needed to be made each time on the 5 appropriate scope with the rule.
6 One of the things that we were thinking 7 under the maintenance aspect of 730 was we didn't need 8 to come up with a scope because we had already, 9 through the safety criteria, defined the scope. And 10 so this is another case where we were hoping it was a 11 little more clear because the scope of this activity 12 is the safety related and non-safety related but 13 safety significant SSC's.
14 Otherwise, the requirement is basically 15 the same as the existing maintenance rule in 50.65.
16 So I think we can go on then to Slide 64, Design 17 control.
18 We've had some discussion here as to 19 whether this was needed or was captured by quality 20 assurance and other requirements. But we do have a 21 design control requirement here under 740 that 22 basically just says that a licensee has to have a 23 program to make sure that as they look at making 24 design changes to the plant that they assess it 25 against safety, security, emergency preparedness, to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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228 1 make sure that there is no unintended consequences.
2 Again, as we look there are some people 3 that say stakeholders who have said this might be 4 duplicative of some other requirements that we have in 5 terms of quality assurance. We'll look at that.
6 The notion wasn't controversial that you 7 need to control the design. It's just whether this 8 specific requirement was needed.
9 So if we go on to 65 and we can talk about 10 just some of the feedback -- some of the feedback, and 11 this goes largely to some of the continuing 12 communication issues that we have that some 13 stakeholders questioned our need to regulate the non-14 safety significant SSCs. Actually, this kind of goes 15 back to the definition, and even including having a 16 definition for non-safety significant SSCs.
17 As you can as we were going through this, 18 actually all the requirements point to either safety-19 related or safety significant SSCs.
20 There has been also with stakeholders some 21 discussions similar to what we've had today going to 22 really having two categories of equipment versus one.
23 And I think we'll continue to be faced with that 24 again.
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229 1 being that radical in having two groups of equipment 2 that are addressed by regulations. We think we can 3 point to plenty of examples where we already have 4 regulations for non-safety related equipment.
5 But then as I said, there's some 6 discussion about the duplication of requirements.
7 Some of that comes out of trying to separate the 8 subparts by the life cycle stage.
9 And so the previous slide on design 10 control, the question would be why not address that 11 within design? So, again, once we get the whole 12 picture together, I think people will be able to look 13 at say it could be organized a little differently or 14 maybe to avoid the perception of duplication, some 15 wording changes might be made.
16 So, let me see. I just got a flash that 17 somebody lost audio. Is that the bridge again?
18 MR. WIDMAYER: No, go ahead. It's just 19 one person, yes.
20 MR. RECKLEY: Okay. Okay. And then as 21 we've talked -- and this will become hopefully more 22 clear as we're able to put the puzzle pieces together 23 and make sure amongst ourselves and then also in our 24 interactions with stakeholders that there's a 25 consistency across the subparts.
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230 1 We're facing a little bit of difficulty 2 now in that we are just timing-wise, you know, we're 3 out of sync a little bit. Like as was mentioned, we 4 prepared what we're giving today, you know, from our 5 point of view. This was what we did a month ago, and 6 we're already working on another subpart and still 7 talking about subparts that we prepared at the 8 beginning of the year.
9 So once we get it together and go through 10 it cleanly to make sure that things are consistent, we 11 hope things will fall into place a little better.
12 And then we did have one comment in terms 13 of the maintenance rule that as systems are 14 implemented that may be able to automate surveillance 15 and testing that we not preclude that. So we'll look 16 at the language to see about that.
17 I'll be honest. That's not an area I'm 18 well versed in. So I'll have to investigate that a 19 little bit. So that is those sections, four or five 20 sections within Subpart F that relate to hardware.
21 We can see if there's any general 22 discussion on those or any additional thoughts on any 23 particular of the sections.
24 The next area we'll talk about is the 25 program section.
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231 1 MEMBER PETTI: Bill, just a comment that 2 I now more fully understand why you went with the tier 3 structure. I think if you just look at Subpart B you 4 don't get it. It's when you get here in some of these 5 details. So I understand, you know, you've got to see 6 the whole elephant, and we're not and so.
7 MR. RECKLEY: Thanks, Dave. And our 8 language in retrospect, we've said many times 9 internally at least we probably shouldn't have used 10 tiers. I wish we had used some different term. And 11 we may very well going forward.
12 And this goes to Joy's question earlier 13 about whether we're revisiting the two tiers. How you 14 call them and how you structure them, you know, we're 15 open for discussion.
16 But what we're trying to address is 17 basically to narrow both the equipment and the human 18 actions and as we'll talk about shortly the programs 19 that are needed to accomplish one goal. And for us as 20 we currently have it defined, that is prevent an 21 immediate threat to public health and safety, which 22 we're using the 25 rem as the measure.
23 And then another goal to address beyond 24 that what is needed to actually make sure that a 25 future plant poses no more risk than a current plant.
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232 1 So if there are people who can tell us how to do that, 2 you know, we're very welcoming to suggestions.
3 But I think, or I hope, everyone would 4 agree that the current operating fleet goes well 5 beyond just limiting the risk profile of existing 6 plants by our regulations is better than just 7 surviving the design basis accident and whether you 8 measure that through the dose or you measure it also 9 through the other elements of the current definition 10 of safety-related, however you want to look at it, 11 plants go beyond that.
12 And what we're trying to capture in Part 13 53 is that acknowledgment. So you have one set of 14 requirements to meet that very stringent requirement 15 and then another set of requirements meeting some 16 other criteria that say the risk profile has to be at 17 least as good as the operating fleet. And so that's 18 what we're trying to do as Dave said. Maybe it 19 becomes more clear as we put the pieces together.
20 But also in that risk management 21 philosophy of what we currently call the second tier 22 is the additional operating flexibilities that are 23 provided within Part 53 for the licensees that is a 24 mix for the operating fleet.
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233 1 fleet was based on a set of regulations that at least 2 in the beginning was binary, it included a whole bunch 3 more safety-related equipment, right, that then we 4 later said was not safety significant.
5 We're trying to avoid that here while 6 still acknowledging that the safety significant 7 equipment is in fact safety significant or risk 8 significant and therefore needs to be controlled under 9 the regulations.
10 So that's the goal. If there's better 11 ways to do it, you know, we're engaging stakeholders.
12 And I think we'll come upon it. Some of the wording 13 might change. But that general notion is what we were 14 trying to do.
15 MEMBER HALNON: This is Greg. One caution 16 about what you said, and I'm not sure I heard it 17 right. But, you know, the operating fleet now has --
18 you mentioned that, you know, they go well beyond the 19 requirements. And that is true. And there is a 20 yearning for maintaining much more, you know, a better 21 risk profile of these plants.
22 That is both based in safety and based in 23 the economics of operating the plant from breaker to 24 breaker. So it's a reliability issue from the 25 standpoint of electrical generation. And I'm not sure NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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234 1 that you want that to infiltrate into the regulations.
2 And that's that first bullet on the stakeholder 3 feedback. The NRC does not need to regulate NSS SSCs.
4 You can see the undertones in that right away so.
5 MR. RECKLEY: I agree. And I do need to 6 be careful because it's both the risk profile that 7 they achieve, which has an element, like you said, of 8 going beyond what's required. But also the risk 9 profile just due to our regulations, which is what I'm 10 trying to address, which goes beyond just the design 11 basis accident.
12 MEMBER HALNON: Yes, I agree with that.
13 I wanted to make sure that all dimensions were being 14 looked at.
15 MR. RECKLEY: No, no. And that's a very 16 good point and a good caution going forward that our 17 goal is to drive them to be at least as safe as we 18 require the operating fleet to be not necessarily to 19 go that extra measure of what's actually accomplished.
20 MEMBER HALNON: And what's required for 21 the electrical generation piece of it.
22 MR. RECKLEY: That's right.
23 MEMBER HALNON: Because some of these 24 advance reactors may not be generating electricity.
25 MR. RECKLEY: Right.
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235 1 MEMBER HALNON: So they may not be breaker 2 to breaker type.
3 MR. RECKLEY: Right. Yes, and we can 4 certainly expect -- well, I shouldn't -- I'll forego 5 what we might expect in terms of a shakedown period as 6 they actually work their bugs out. But anyway, Joy, 7 I think maybe you had a --
8 MEMBER REMPE: Wow. You're good. You're 9 even seeing when I unmute. I guess I just wanted to 10 also say along similar to what Dave is saying that it 11 does make sense if one thinks about the LMP and other 12 things besides 10 CFR 100 and the dose criterion to 13 have the two tiers. My question was more that, man, 14 there's going to be a lot that will have to be changed 15 if you don't have those two tiers and, of course, if 16 you -- in these other parts.
17 But I do think more explanation is needed 18 that we've heard during these meetings to hopefully 19 communicate to others.
20 MR. RECKLEY: Okay. I definitely agree 21 with that. In addition to the technical, we've had 22 communication challenges. And that's largely on us.
23 So we're trying.
24 MEMBER KIRCHNER: Bill? This is Walt 25 Kirchner.
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236 1 MR. RECKLEY: Yes.
2 MEMBER KIRCHNER: On this second tier 3 safety criteria, which is tied to the QHOs, which 4 becomes a really difficult regulatory bar to put in 5 the regulation and then measure.
6 Have you thought of -- I'm thinking, you 7 know, since in much of what we've heard in the last 8 few meetings, we have a problem because you're trying 9 to be so catholic, that's lower case c catholic, in 10 your approach, I have too many metaphors here perhaps, 11 but the tail often is wagging the dog.
12 And what I'm referring to is these small 13 microreactors that aren't going to be so 14 microreactors. Put the military applications aside 15 for the moment, they are probably going to be in the 16 megawatt class to be anything approaching economic 17 viability. So they will have significant source 18 terms. We can't just discard that out of hand.
19 But for your existing NPUF and test and 20 research reactors, don't you use a 1 rem threshold as 21 kind of a bar so to speak for measuring safety 22 consequences? One would also need to have a 23 frequency, I guess, contribution to that. But is it 24 feasible to define some simple dose as a surrogate for 25 the QHOs?
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237 1 MR. RECKLEY: Yes. We're looking at that 2 in terms of both is there a potential surrogate for 3 the QHOs? Also if you're going to approach it through 4 a maximum hypothetical, what might the dose criteria 5 be?
6 And, you know, that gets into a whole --
7 you know, we talked about trading margins against 8 regulatory requirements. And there may be different 9 thresholds depending on what you're trying to trade 10 off, right?
11 One rem is often used for the emergency 12 planning in terms of the evacuation. We're proposing 13 the 25 rem as a measure or as a criterion for an 14 immediate threat to health. There are other criteria 15 that might come into play.
16 And I don't have a specific example, but 17 there is a dose criterion within the protective action 18 guidelines for relocation, for example, 500 millirem 19 a year. The Part 100, I mean, the Part 20 dose values 20 of 100 millirem that we currently include within the 21 first tier criteria, those kind of numbers could be 22 used in regards to research and test reactors.
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238 1 as a quick measure, right?
2 If they go above that and they have to 3 show then that they're going to use the guidance under 4 or the requirements under Part 100, which is the 25 5 rem number, then an additional set of controls come on 6 for those few RTRs that have that kind of potential.
7 So, yes, we are looking at how these 8 various dose thresholds might be used and especially 9 when you start to look at what they may be proposing 10 in terms of some of their other programs, like we 11 talked about this morning on staffing.
12 You know, one question. When would you 13 ever consider letting a facility be unstaffed? Well, 14 would that be the 25 rem number? The 1 rem number?
15 The 500 millirem a year number? The 100 millirem out 16 of Part 20? You know, those are questions. I don't 17 have an answer for any of that so.
18 MEMBER KIRCHNER: I think that -- one 19 thing that occurs to me is that you often invoke the 20 second tier and associate it with defense in depth 21 measures by the non-safety but risk significant 22 systems.
23 And preventing, for example, getting to 1 24 rem and triggering the PAGs for an EP if you required 25 an EP strikes me as a kind of a -- again a kind of a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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239 1 threshold that's recognized by the EPA.
2 One could quibble. One could say the 500 3 millirem that you mentioned. But it certainly then 4 makes sense out of invoking defense in depth and 5 requiring the RTNSS equivalent treatment for those non 6 -- let me get the terminology right again, the non-7 safety related risk significant systems.
8 MR. RECKLEY: Right, right.
9 MEMBER KIRCHNER: It's just I've been 10 thinking about this since your very first 11 presentation, again, because you're dealing with such 12 a broad array of potential designs with all their 13 unique quirks and such, getting surrogates for CDF, 14 delta CDF and things like that obviously would be an 15 enormous challenge and a set of complexity.
16 So I keep thinking for regulatory 17 certainty falling back on something that's perhaps 18 either through PRA and/or deterministic means readily 19 demonstrated with approved methods to again to achieve 20 some kind of certainty amidst what's going to be such 21 a broad array of technologies that come before you.
22 MR. RECKLEY: Okay.
23 MEMBER KIRCHNER: That was an observation 24 not a question.
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240 1 you know, we'll combine that with the other 2 observations and comments we're getting. And the use 3 of the QHOs is a topic that a lot of people are 4 commenting on. So we will certainly take it into 5 account. Anyone else before --
6 MEMBER BIER: Yes.
7 MEMBER BIER: Bill, this is Vicki Bier 8 again. It seems to me that one of the reasons why the 9 current fleet has been required to achieve higher 10 levels of safety is basically because of the backfit 11 rule. As we go along, we learn things, and then some 12 of those things are worthwhile fixing and then we 13 mandate fixes to them. And so, on the one hand, it's 14 hard to imagine how you could be smart enough to 15 identify all those fixes ahead of time, but I think 16 this does show kind of an inherent difficulty of 17 managing as diverse a fleet as the advanced reactors 18 might turn out to be because you no longer have the 19 opportunity to learn from one and then apply to lots 20 of others.
21 So I don't have any concrete suggestions 22 but just something to think about on that issue of is 23 it feasible to achieve the same level of safety as the 24 current fleet.
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241 1 into the -- very good point and, you know, as we get 2 into the other subparts in the next few months, we 3 will maintain that requirement in case we need to 4 employ it and learn a lesson that Part 53 didn't 5 address very well. So, yes, we will, as get into 6 Subparts H and I, the licensing parts, talk about 7 those things.
8 Okay. Seeing no other hands, let's go 9 down, Liz, to -- yes, go ahead. So the next few 10 sections that we're going to talk about are the 11 program requirements. Again, our expectation is that 12 the personnel will be in the middle here. We just 13 don't have language yet for personnel.
14 So we'll go onto programs, slide 69. It 15 starts off with Section 53.800, which basically sets 16 out a generic requirement for the licensees to 17 identify any required programs. The subpart is going 18 to include some that we have little doubt will be 19 needed, but we also, sort of what Vicki just 20 mentioned, not knowing exactly how this might evolve 21 for technologies for which we have very little 22 experience. 53.800 basically puts an onus on the 23 licensee to identify any needed programs to satisfy 24 the first or second tier safety criteria.
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242 1 sections, these are ones that we know ahead of time 2 that they will need. So if we go onto slide 70, 3 radiation protection, this is pointing to Part 20 and 4 also including some preliminary language to address 5 the controls for routine affluence. And so we picked 6 up basically the requirements from 50.36(a) as a 7 starting point. Some of these programs we're still 8 working internally, but we felt comfortable enough as 9 a starting point to go ahead and include, that they 10 would need a radiation program to meet Part 20 and 11 they would also need some kind of program related to 12 routine affluence, the monitoring, and the reporting 13 of those as they're currently addressed within 14 50.36(a). So I don't think there's too much 15 controversy on that one.
16 Slide 71 goes to emergency planning. And, 17 likewise, we know ahead of time there will be likely 18 some programs needed for emergency preparedness. We 19 have, as most of you are aware, a proposed rule out, 20 and we're kind of looking at that and seeing how to 21 coordinate that proposed rule with this one. So we 22 didn't really prepare much text yet when we released 23 the program requirements, other than we'll need to 24 coordinate it with that rulemaking.
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243 1 security. We are currently preparing some language 2 for Part 73 on physical and cyber security. We're 3 looking at areas, like access authorization, fitness 4 for duty, material control and accounting. This 5 preliminary version that we put out didn't include 6 that language. We're preparing that now. We'll be 7 releasing it in the next few weeks to start to engage 8 stakeholders on that language.
9 So Jordan mentioned that we were trying to 10 keep 53 basically free from referring to 50 and 52 for 11 fear of getting into confusion between those 12 regulations. But there are other parts of the 13 regulations that we do plan to continue to point to, 14 and those are related, for example, I mentioned Part 15 20 and Part 73 on security, Part 26 on fitness for 16 duty. Those will continue to be parts that address 17 those requirements, and Part 53 will point to those 18 versus trying to include everything within Part 53.
19 But, again, the shorthand on security 20 programs under Section 53.830 is that we didn't 21 include text. It will be coming out in the next few 22 weeks.
23 So if we go then to slide 73. This gets 24 into the operational quality assurance program, and 25 really the only note, I guess, that we have is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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244 1 consistent with the other subparts. We've tried to 2 say or use language like generally accepted, consensus 3 codes and standards as a way to accommodate proposals 4 to use something other than NQA-1, which is basically 5 the current primary code or standard used for quality 6 assurance.
7 But other than that, what you'll see 8 written in terms of the individual items are taken out 9 of Appendix B, and there were those that we thought 10 were applicable to the operating phase of a facility.
11 MEMBER BROWN: Bill?
12 MR. RECKLEY: Yes.
13 MEMBER BROWN: I mean, so you've 14 genericized it with the consensus codes and standards.
15 I mean, what's wrong with NQA? What's wrong with the 16 addition? Why isn't that just a cut-and-paste? You 17 don't have to refer to it, but, I mean, it's been 18 tried and true, it's been used throughout the 19 industry, and now we're going to go argue about 20 whether some other standard is okay on some future 21 advanced plant?
22 MR. RECKLEY: Yes. I think we would 23 continue to say that NQA-1 is an acceptable approach.
24 And given our preparing Part 53 to basically reflect 25 Appendix B as it's currently written, although we're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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245 1 distributing it throughout the subparts, NQA-1 would 2 continue to be a way to meet that requirement.
3 Stakeholders have suggested that ISO standards, 4 quality assurance standards from other countries or 5 maybe even IAEA, might be used and they wanted us to 6 remain open to that.
7 So what we're trying to say is -- and the 8 current requirements don't specifically point out NQA-9 1, I don't believe. NQA-1 as it was developed and 10 evolved along with Appendix B, you can definitely see 11 the relationship between Appendix B and NQA-1. But 12 all we're trying to say is maybe there's an 13 alternative, and, if someone wants to propose it, then 14 we would be amenable to an alternative if it can be 15 shown to be effective. So nothing is wrong with NQA-16 1, but there may be an alternative.
17 MEMBER BROWN: Well, there's always an 18 alternative. I mean, it probably has less 19 requirements. I just, it's just beyond my thought 20 process to take stuff that obviously was barely 21 applicable and is well used and well understood and 22 just not go ahead and go with it. Now you're going to 23 set yourself up to have to compare every other 24 proposal against what's present done, which is a huge 25 manpower effort and review effort. It's just -- I'll NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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246 1 stop there.
2 MR. RECKLEY: Yes, okay.
3 MEMBER PETTI: Isn't it fair to say that 4 some of it had to do with international supply chains, 5 some of the stakeholder input? You know, you want to 6 buy something from Europe, it's not NQA-1, it's ISO, 7 whatever the number is, that sort of stuff.
8 MR. RECKLEY: Yes, that's the driver from 9 the stakeholders' perspective.
10 MEMBER KIRCHNER: But could you pull this 11 all forward to Part B, Subpart B, Bill, and just say 12 NQA-1 or comparable? I'm kind of resonating with 13 Charlie. I know some of the stakeholders are 14 interested in selling their reactors offshore and/or 15 using an international supply chain, but it seems to 16 me, one, that this could be pulled forward -- I think 17 we discussed this in one of our past meetings --
18 instead of being repeated so many times throughout 19 because then you fall into the trap of, well, I'm 20 going to take the 18 sections of NQA-1 -- here I've 21 got, I can't count the bullets, but let me just say 22 ten -- ten bullets out of NQA-1 for this subpart and 23 then another subpart I'm going to use, just to be 24 arbitrary, 12 points out of NQA-1 or Appendix B, 25 whichever way you like to refer to it. It would seem NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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247 1 to me that moving it, once you get all the pieces 2 together, this might be a good candidate to move to 3 Subpart B and just use the ASME NQA-1 as the standard 4 or say or equivalent, so you're not into rulemaking, 5 just the change of reference to one standard.
6 MR. RECKLEY: The structure we are 7 thinking about, and, along the lines, you said to see 8 if the repetition and so forth, it would just be more 9 clear if we condensed it to one place, whether it be 10 Part 53, Appendix B, or Subpart B. We are looking at 11 that, and that comment has been engaged from some 12 other stakeholders, as well.
13 But, overall, I guess I'd just say for 14 Part 53 we're not looking to change the actual 15 regulatory requirements. And I'll have to go back and 16 look again. I don't think NQA-1 is one of the 17 standards we incorporate by reference into the 18 regulations anyway. It's well established, as Charlie 19 has said, but it's not actually currently in the 20 regulations, I don't believe. I'll have to go back 21 and look.
22 MEMBER KIRCHNER: It takes its form as 23 Appendix B, but if you compare Appendix B to NQA-1 24 you'll get a --
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248 1 organized.
2 MEMBER KIRCHNER: -- almost a one-for-one 3 match.
4 MR. RECKLEY: Right.
5 MEMBER BROWN: The point being it has 6 nothing to do with advanced reactors.
7 MR. RECKLEY: I understand. It's quality 8 assurance.
9 MEMBER BROWN: Many of these have nothing 10 to do with advanced reactors. It's just reducing the 11 standards and reducing, making it a mush ball. That's 12 all. It's very frustrating. Go ahead.
13 MR. RECKLEY: Okay, all right. If we go 14 into slide 74, this actually is somewhat, somewhat 15 new, and so what we're proposing in the preliminary 16 language is to look at what is needed to look at 17 degradation. And so we looked at a number of programs 18 that relate to this, and one would be aging 19 management. As most of you are aware, that comes into 20 play for license renewal. Our proposal and the 21 preliminary language is to require an aging management 22 program from the beginning.
23 Another one that goes to degradation, at 24 least somewhat, is the requirement that often is 25 included in tech specs to monitor the cyclic and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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249 1 transient loads, looking largely at fatigue. And then 2 the last one, last sub-bullet there, is a requirement 3 to have a program that looks at materials and 4 potential chemical interactions, temperatures, 5 radiation, other factors, and to have a program to 6 ensure that degradation is not challenging the ability 7 of an SSC to perform its function.
8 And so that's somewhat of a new 9 requirement. You can go into some parallels for light 10 water reactors. As degradation mechanisms have been 11 identified that we've put in place requirements to 12 look at things like intergranular stress corrosion 13 cracking or some other phenomena that was not 14 identified at the beginning of operations. So what 15 we're proposing to require is a program where the 16 licensee has to have a program in place that looks at 17 that, looks for those possible mechanisms.
18 So I see a couple of hands up. I must 19 have struck a nerve on this one. So, Ron, I think you 20 got your hand up.
21 MEMBER BALLINGER: Yes. I mean, my 22 goodness, over the past 30 years, this degradation 23 management and inspection criteria has evolved to the 24 point where it's extremely well established. I mean, 25 you know, I think this is an easy one to deal with.
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250 1 You have Section 11 for primary system boundaries and 2 things like that. You have the degradation 3 management, which has evolved, oh, my goodness, over 4 20 years to the point where, you know, ISI, periodic 5 in-service inspections.
6 So I think, I mean, I just think this is 7 an easy one to identify. And it's well established 8 and it has reduced the number of -- what do you want 9 to call it -- surprises to a number which is almost 10 vanishingly small.
11 MR. RECKLEY: Well, good. And our 12 thinking in including it reflects partly that, the 13 history, and also partly the importance of this when 14 you're going to look at new technologies with 15 different coolants, different materials, different 16 fuels, different a lot of things, different 17 temperature ranges, and the need to do this from the 18 beginning.
19 MEMBER BALLINGER: But then you just do, 20 like, Section 5 for high temperature materials. You 21 just let the code, the codes will evolve to 22 accommodate those materials.
23 MR. RECKLEY: And we have, actually, now 24 under review the reliability and integrity management 25 section of, I guess that's at Section 11 of ASME, so NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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251 1 that's currently under review.
2 MEMBER BALLINGER: Yes, yes.
3 MR. RECKLEY: So, yes, I think I'm 4 agreeing with you. I think the pieces fall together.
5 But I'm not sure if you go into Part 50 you'll find 6 something as clear as this. And sometimes the 7 stakeholders are pointing out where they think 8 additional requirements are being added. This was one 9 of them. Again, under our view, we think it makes 10 sense, given what has evolved and also the importance 11 of looking at this for new designs, new operating 12 plants.
13 MEMBER BALLINGER: Yes. I mean, the 14 overall process by which the existing sets of criteria 15 and procedures had evolved over time can be used as a 16 -- what do you want to call it -- a template for going 17 forward for advanced reactors as a starting point.
18 MR. RECKLEY: Okay. Yes, I would hope so.
19 Vicki, I see your hand up, as well.
20 MEMBER BIER: Sorry. I think I just 21 didn't put my hand back down maybe.
22 MR. RECKLEY: Okay, okay. So that is the 23 proposal under 53.850 to have an integrity assessment 24 program or degradation management program or whatever, 25 ultimately, it ends up being called.
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252 1 That then goes another program on the next 2 slide, slide 75.
3 MEMBER KIRCHNER: May I -- Bill, this is 4 Walt.
5 MR. RECKLEY: Sure.
6 MEMBER KIRCHNER: Can I make an 7 observation? This goes back to now, your underlying 8 approach here is with using PRA means that you need to 9 maintain design control reliability for the equipment 10 that you identify as either safety related and/or not 11 safety related but safety significant or risk 12 significant. And it seems to me this would be an 13 adjunct to any comprehensive reliability assurance 14 program that you would, among other things, look at 15 the effects of aging and such. So I'm not objecting 16 or anything. It just seems to me it would have to go 17 kind of hand-in-glove with any kind of reliability 18 assessment that's continued through the life of the 19 plant and feed that so that you can remain assured 20 that your plant configuration meets the reliability 21 goals that were put into the PRA to, you know, provide 22 the design basis for the plant.
23 MR. RECKLEY: Exactly. That's what we 24 were, that's the way we envisioned it and why we 25 included it. Now, again, as we go into this, how it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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253 1 actually gets captured were free. For example, a 2 couple of slides down the road we'll talk about ISI 3 and IST. Some people have said this would be 4 addressed or could be addressed within ISI and IST --
5 I'm sorry, in-service inspection and in-service 6 testing. If that turns out to be the case, you know, 7 we can change the language just to make sure that this 8 is captured under those programs. We were just under 9 this preliminary language trying to make sure that we 10 captured it for exactly the reason you just said, 11 Walt.
12 So I think we can go on then to slide 75, 13 fire protection. We're currently looking at the 14 language that we have in here, which is largely 15 borrowed from existing requirements. But it's another 16 area where I don't think there's too much controversy 17 that a fire protection program during operations will 18 be needed in order to control combustible materials 19 and so forth. So, again, we just largely borrowed 20 this. We're looking at it and engaging stakeholders 21 on the exact wording, but the program would be laid 22 out with the same objectives of making sure that, you 23 know, that you're doing adequate things to prevent 24 fires, detect, and suppress them if they should occur.
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254 1 incorporate NFPA 805 the way it's -- I went off and 2 looked at the 50.48 to see how that's done, and that's 3 specified and you all didn't, you didn't include that 4 in this.
5 MR. RECKLEY: Yes. And this is another 6 case, to be honest, where we didn't think it was 7 controversial to say there's going to be a fire 8 protection program, but we're currently working with 9 the fire protection staff to see how to actually build 10 that in, not only here but if you'll remember back 11 when we talked about Subpart C on design and analysis, 12 fire protection was just a bullet.
13 MEMBER BROWN: Yes.
14 MR. RECKLEY: And so we need to expand 15 both, and we'll have to address the approach for 805 16 within both of these subparts.
17 MEMBER BROWN: I'm obviously not, that was 18 going on, a lot of stuff going on when I got there, 19 but I thought it's not absolutely required that any of 20 the existing fleet adopt 805, is there? I mean, I 21 thought that --
22 MR. RECKLEY: Well, that's, that was 23 optional. Right.
24 MEMBER BROWN: Yes. And do we want it to 25 be optional for advanced reactor plants? It seems to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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255 1 me we went through a lot of effort through all the 2 meetings. I remember John Stetkar and, I guess, 3 Dennis, I thought you participated in that in terms of 4 its development and application. And it obviously is 5 a part of -- what is it? Appendix B or -- I can't 6 remember which appendix it is. It's down at the end.
7 It's down at the end. It says, you know, it's 8 obviously optional, but it just seems to me after all 9 that work to get it implemented and some people 10 implementing it and using it that the why throw the 11 baby out with the bath water if that's an improved 12 method, improved performance, or an improved set of 13 criteria for doing fire protection.
14 MR. RECKLEY: Yes, so that's a great 15 question. And, again, where we are right now is we're 16 looking, 805 was specific to light water reactors, but 17 the general concept and another question is, you know, 18 we've talked a fair amount about the non-light water 19 reactor PRA standard to the degree that fire is going 20 to be addressed as an event within the PRA.
21 So all of that are things that we know we 22 need to work out, so great, great point, Charlie. And 23 I think, as we develop the next iteration for fire 24 protection, you'll see a little better argument than 25 just the placeholders that we have in here now.
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256 1 MEMBER HALNON: Charlie, this is Greg. I 2 think there would be some pushback to say that it was 3 better than the previous version based on how 4 difficult it was to transition. If we can get our 5 arms around that, 805 or at least the concepts in 805 6 would be really good. But the actual transition from 7 Appendix R to 805 was very painful and expensive, and 8 I think there's a lot of plants that got into it and 9 reversed their stance on it.
10 MEMBER BROWN: I understand that. I 11 understand that, but that's when you have to 12 transition. But if you're starting with a clean 13 sheet, it just seems it would be -- and if that's a 14 preferred method that NRC prefers because details in 15 it, I mean, the approach to doing it is much better 16 than trying to transition from an old program to a new 17 one.
18 MEMBER HALNON: Right. I agree. If 19 there's some better way of doing it and certainly up-20 front.
21 PARTICIPANT: Excuse me. There's a note 22 in the chat about the bridgeline. I just wanted to 23 alert you.
24 MR. WIDMAYER: Hey, Dennis, Thomas lost 25 the bridgeline again. He said he could get it back in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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257 1 about five minutes.
2 MR. DASHIELL: It's not lost. I just need 3 a five-minute break to reconnect batteries going out.
4 CHAIR BLEY: We can take a five-minute 5 break. I just wanted to throw in something before we 6 leave. There are a lot of extenuating circumstances 7 that led to a lot of the problems that were on many 8 sides with 805. But we always expected the PRA 9 methodology that was recommended there to turn in to 10 what PRA methodology might be here. We just got the 11 revised version of the other PRA, so we'll be looking 12 at that and we'll come back to this after we looked at 13 that one and see what those bring up.
14 So come back in five minutes. 3:35 here, 15 about 35 there. We'll recess for five minutes.
16 (Whereupon, the above-entitled matter went 17 off the record at 5:30 p.m. and resumed at 5:35 p.m.)
18 CHAIR BLEY: Okay. We are now back in 19 session. Bill, maybe I can sneak in a quick question.
20 I've just been thumbing through the new standard.
21 There's a lot of pages about fire. Has anybody 22 compared it with the approach laid out in 805? I kind 23 of imagine that it's much different.
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258 1 the PRA standard, so I can't answer.
2 MR. STUTZKE: Hi, Dennis. None of the PRA 3 standard was basically taken from the old WR standard.
4 So, you know, it should be consistent with NFP 805.
5 As Bill pointed out, 805 only applies to LWRs.
6 CHAIR BLEY: Yes. But the methods used to 7 do the PRA part --
8 MR. STUTZKE: Are all the same, yes.
9 CHAIR BLEY: Oh, same, yes. Okay.
10 Thanks. Go ahead, Bill.
11 MR. RECKLEY: Okay. Thank you, Marty.
12 So, yes, so we can move on. Again, this is an area, 13 fire protection, I'll just say, is an area, I think 14 you'll see changes in the next iteration, but, as a 15 placeholder, I don't think it's controversial that 16 there would be.
17 Somewhat the same is true for ISI and IST, 18 in-service inspection and in-service testing, that 19 they would have to have programs. We will be looking 20 and engaging stakeholders to see, as Dr. Ballinger was 21 saying, whether some of the degradation mechanisms 22 would already be picked up. For example, if they're 23 doing ISI and there's the new division for Reliability 24 and Integrity Management, the RIM, is already included 25 in ISI, maybe parts of that.
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259 1 So I don't think it's controversial that 2 we have them. It is, it's something we'll engage 3 stakeholders about how to best organize it.
4 Because this is technology-inclusive, we 5 don't currently plan to specifically incorporate by 6 reference the ASME as we do in 50.55(a). So, again, 7 we would have the requirements to have these programs 8 and have them take advantage of generally accepted 9 consensus codes and standards but not incorporate by 10 reference ASME. That's probably the biggest 11 difference, but that's also an issue we talked about 12 and it's something that affects a couple of different 13 sections as we go to that approach.
14 So I think we can go to 77.
15 MEMBER BALLINGER: This is Ron again.
16 MR. RECKLEY: Yes.
17 MEMBER BALLINGER: I mean, I think we need 18 to -- I hate to use a buzzword because it will be a 19 non-buzzword two years from now. But, in effect, the 20 evolution of ISI and other degradation models and 21 things with the existing plants have resulted in 22 amounting to what you're operating is a digital twin 23 during actual operation of the plant. And it's 24 evolved to the point where if you combine online 25 monitoring with degradation models that have been NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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260 1 developed, what you have is, by the time a refueling 2 outage occurs, everything that needs to be done with 3 respect to materials degradation is already identified 4 and everything. And my guess is that with advanced 5 nuclear plants it will be the same thing only in 6 spades much better. And when you combine PRA with 7 existing models, you'll be able to identify potential 8 areas long before they occur. And so, in effect, what 9 the advanced reactor is is you're basically flying an 10 engine in a commercial aircraft.
11 MR. RECKLEY: And, you know, we're aware 12 of the DOE programs and other efforts in areas like 13 digital twins. You know, one question, and I'll just 14 leave it out there if you have a suggestion, is that 15 that might be a very useful tool. We're not, at least 16 in the preliminary language, we didn't include 17 anything that that would be a requirement. But maybe 18 that would be a key area where that tool would be used 19 to meet whatever requirement we do put in Part 53. I 20 think that's the way we were looking at that, the 21 advances in areas such as that.
22 MEMBER BALLINGER: Yes, I would expect 23 that would happen no matter what. And, you know, I 24 said this once before, but on the NASA side for 25 satellites and things like that they have a system in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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261 1 place where they're basically flying a model along 2 with online monitoring and the like, comparing what 3 they think should happen with what is actually 4 happening and computing what they term time to 5 critical effect, which means all of a sudden something 6 is not looking very good. And because they need time 7 to take action, the operator gets a notice that says 8 you have this many whatevers before we're going to 9 shut this thing down. And so I'm guessing that that's 10 certainly going to be the case for some of these 11 remotely-operated or unattendedly operated plants.
12 MR. RECKLEY: Okay. Thank you. And, 13 again, our idea, at least at this stage, is to try to 14 make sure we don't preclude anything like that 15 advancement but probably not require it.
16 MEMBER BALLINGER: I mean, I personally 17 don't believe that an unattended operation or 18 autonomous operation would even be allowed without 19 some form of this kind of process in place. I've been 20 converted to PRA. Sorry.
21 MR. RECKLEY: Okay.
22 CHAIR BLEY: Can I make one quick comment?
23 MR. RECKLEY: Sure.
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262 1 it's in Section 4.3.9, and it runs for about 100 pages 2 from 162 to about 200. It's not a hundred pages.
3 Anyway, that's where you can find it.
4 And the standard doesn't tell you what 5 methods to do. It tells you what concepts you have to 6 address and to address those if you use the 7 methodology that was laid out with 805, and you get, 8 I think, all of them, but I haven't checked them all.
9 MR. RECKLEY: Joy, did you have --
10 MEMBER REMPE: Just a quick comment. It's 11 not just DOE looking into digital twins. Research has 12 been looking into it. You're aware of that, right?
13 MR. RECKLEY: Yes, yes, yes.
14 MEMBER REMPE: Okay, great.
15 MEMBER BALLINGER: Two years from now it 16 will be called something else.
17 MEMBER HALNON: What was the impetus 18 behind that last bullet there, provide results to the 19 plant manager? Was that just not carefully worded, or 20 is that some deficiency that you're trying to plug a 21 hole?
22 MR. RECKLEY: I think we either, I think 23 we were pulling that out of either guidance or 24 existing requirements. I'll be honest, I don't 25 recall.
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263 1 MEMBER HALNON: Okay. It just looks odd.
2 You know, this is, we're so mature in our corrective 3 action programs now that anything much less than the 4 accepted criterial probably would be, would still be 5 provided a corrective action program. It will wash 6 out in the future, I'm sure.
7 MR. RECKLEY: Okay. Thank you, yes.
8 Okay. I think then we can go onto 77, which is our 9 attempt just to pick up another programmatic 10 requirement out of both Part 70 and out of Part 50 for 11 criticality safety programs. And our preliminary 12 language was just to refer to Part 70 and then ask 13 stakeholders if they wanted us to explore some of the 14 alternatives that were included in 50.68, which takes 15 advantage of some of the things you would expect to 16 have in a large power plant, as opposed to Part 70 17 that was addressing it for any licensee that might 18 have special nuclear material.
19 And so the feedback was that they wanted 20 us to explore this, so we probably will be looking at 21 50.68. It has some light water-specific elements to 22 it, but trying to potentially build in an alternative 23 for criticality safety for advanced reactor 24 facilities. So, again, the feedback was not 25 controversial that we proposed to have such a program, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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264 1 some suggestions on possible changes.
2 So 78 goes into something we had 3 previously provided to this subcommittee because this 4 was an element of Subpart F that we had thought about 5 and prepared several months ago. And so now it just 6 is finding the place for it to rest in its natural 7 place, which was 53.890. And this is again our 8 proposal which is another new thing that we're 9 thinking about for Part 53, and that is a more risk-10 informed approach, putting more of the onus on the 11 operator to look for possible risk reduction measures, 12 to define some criteria and some actions if they come 13 across those. This has gotten a mixed reaction with 14 stakeholders.
15 But the preliminary language in the 16 current material that we provided didn't change 17 significantly. Some of the comments, even from the 18 subcommittee, was the inclusion of a lot of the 19 administrative requirements made it a fairly long 20 program description, and maybe some of that could be 21 relocated to guidance. And we're looking at that, but 22 we didn't have time to do it before this release. So 23 we basically just made a few editorial changes and 24 included the previous version of the Facility Safety 25 Program.
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265 1 So, again, just quickly, it would require 2 a licensee to look for new information in regards to 3 hazards and plant performance. And if an issue 4 crossed a threshold, then they would be required to 5 assess whether it was cost effective to do a risk 6 reduction measure. So the concept would be they could 7 use the guidance, they could use the numbers, for 8 example dollars per person RIM, that is used for 9 things like severe accident management alternatives, 10 our backfit process, those established procedures that 11 use the same general approach to assessing whether a 12 risk reduction measure is cost effective.
13 So if we go on to slide 79, again, I 14 hadn't really planned and I don't think we need to go 15 into detail because we had provided this in a previous 16 meeting. So this just basically says what I just 17 said, assess risk reduction measures.
18 The criteria and the thresholds that we 19 have in here, again, were just placeholders as we get 20 into the next phase and start to engage stakeholders 21 on specifics. If we decide to keep the program, then 22 we'll get into that more detailed discussion of what 23 are the actual thresholds and numerical values within 24 these sections.
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266 1 we're looking at the possibility of taking out some of 2 the administrative details because it did make the 3 program read as if it was bigger and longer than the 4 other programs, which is probably not the case. It 5 was just that we included a lot of detail.
6 And so we can go on to 81. And that's, 7 again, we talked about this in previous engagements.
8 So we can go to the last program that we 9 have, which is for procedures and guidelines. And so 10 this is just laying out a lot of the requirements that 11 are currently reflected in various regulations and 12 also usually in the administrative section of tech 13 specs, but we're calling out a specific program to 14 maintain procedures and guidelines and to integrate 15 those procedures and guidelines across the spectrum of 16 possible operations from normal operations, 17 maintenance, all of the procedures that would be 18 required to carry out the programs that we just talked 19 about. Then down to emergency operating procedures 20 and, ultimately, to various forms of guidelines, like 21 accident management guidelines, extreme damage 22 management guidelines, and so forth. So this is, 23 again, just kind of collecting in one place the 24 requirements to have an integrated set of procedures.
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267 1 some high-level feedback. I did this as we went 2 through the various sections, but there were some 3 discussion whether some of the programs, like the 4 transition to operations, was new and excessive. And, 5 again, that would largely be determined on whether it 6 could be captured more efficiently somewhere else.
7 Same with integrity assessment. If it's already going 8 to be included within ISI and IST, then it wouldn't be 9 required and calling it out as a separate program 10 would add unnecessary burden, provided that it is 11 captured somewhere else. And we have an ongoing 12 discussion, like I said, with a mix of reactions to 13 the Facility Safety Program.
14 The last public meeting had a little bit 15 of contention, so that's all that last bullet was 16 trying to say is whether the interactions we've had to 17 date have been all that productive was brought up by 18 both stakeholders on all sides. So people are getting 19 a little frustrated, I think.
20 So, with that, I think we can be finished.
21 We had a few backup slides on previously-released 22 sections, but, to be honest, I think we talked about 23 many of those things as we went through the 24 presentation. So I'll just turn it back to you, 25 Dennis.
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268 1 CHAIR BLEY: Thank you very much, Bill.
2 And I agree with you. We probably covered them, and, 3 if not, I'm sensing burnout from the way things are 4 going here toward the end.
5 MEMBER BROWN: Could I make one last 6 comment or observation?
7 CHAIR BLEY: Charlie, go ahead.
8 MEMBER BROWN: Yes. Just an eyeball 9 check. There's 47 pages for Subparts A through F.
10 Parts B and C occupy 7 or 8, and all the rest is 11 miscellaneous stuff. It seems like we have, we've 12 shortened up what the design objectives and basic 13 design requirements criteria are, but we put a lot 14 more detail in than a lot of the other places. It 15 just seems to me that that's something we ought to 16 consider: are we being too shy on the types of items 17 would fall under Part B safety objectives and what you 18 need to do to meet those and Part C for the design and 19 analysis.
20 CHAIR BLEY: Thank you, Charlie.
21 MEMBER BROWN: Just an observation.
22 CHAIR BLEY: Vicki, go ahead.
23 MEMBER BIER: Yes. This is just a general 24 observation, and I'm sure it's not anything that staff 25 hasn't already thought of. But it strikes me that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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269 1 every place that you try to provide flexibility, such 2 as, you know, alternative QA protocols or whatever 3 else, is creating a greater burden for staff 4 eventually to have to review all those and decide 5 whether it meets the intent and, you know, a more sort 6 of checkbox regulation system is going to be easier to 7 administer.
8 So I don't think that's a reason not to do 9 it. There may be reasons why we need that flexibility 10 to account for the diversity of what's being proposed, 11 but I think it's important that staff think through 12 both whether they have, whether the agency has the 13 right expertise to review all those topics and just 14 the time that it will take to review all of them may 15 be greater than if it's a more cookie-cutter approach.
16 CHAIR BLEY: Thanks, Vicki. We'll have 17 the time after public comments for members to make 18 comments. If any members have questions for the 19 staff, now is the time for that.
20 MEMBER BIER: Okay.
21 CHAIR BLEY: Okay. At this point, we had 22 a request for a statement from the Nuclear Industry 23 Council, and I think, Mr. Cyril Draffin, are you going 24 to do that again?
25 MR. DRAFFIN: Yes, I am.
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270 1 CHAIR BLEY: Please go ahead.
2 MR. DRAFFIN: Thank you. I'm Cyril 3 Draffin, the Senior Fellow for Advanced Nuclear at the 4 U.S. Nuclear Industry Council. We have five points 5 and a couple of other comments.
6 First, the promised reduction in 7 operational burden is not apparent. The NRC staff 8 promised that Subpart F would enable a significant 9 reduction in operational burden as compared to Parts 10 50 and 52 and the increased burden in Subparts B and 11 C were necessary to get that operational burden and 12 decrease it.
13 But now having seen some of the parts at 14 least of Subpart F, it's not clear what the benefits 15 are. The preliminary language seems to result in a 16 decreased burden, some flexibility, and doesn't 17 provide any enhancement to safety in comparison with 18 Parts 50 and 52. The only major benefit may be you 19 don't have to seek exemptions for the large LWR-20 specific requirements.
21 At the last Part 53 meeting, industry 22 asked NRC to explain if the regulations NRC were 23 adding were needed for safety. So something to 24 consider.
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271 1 hasn't addressed how to limit the staff open-ended 2 application reviews with guardrails other than 3 established backfit, especially when increasing 4 regulatory burden. And Joy brought that up as just a 5 comment recently to think through the process, 6 including the review part. So it's not just the 7 regulations and guidance but also how it's going to be 8 implemented.
9 Second point, the Subpart F, at least in 10 its current form, does not seem to be that useful. A 11 number of developers think the current approach is 12 flawed and recommended the retraction of this proposed 13 Subpart F language and we start from a clean sheet of 14 paper. So that's pretty strong.
15 Third, the Subpart F programs which we 16 just discussed are redundant to existing programs and 17 may not be needed at a larger envelope program. We 18 talked about, as mentioned, the Facility Safety 19 Program, I don't think there's a mix of industry 20 reaction. I think it was pretty much uniformly 21 against it. This program, it came from the railroad 22 industry, so there's not much support for that program 23 as being duplicative of things that are being done 24 already.
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272 1 agree with the comments that Bill made. It is new.
2 It may be duplicative, and so the question is what is 3 trying to be accomplished that is not being done 4 already, and so that's something that needs to be 5 thought through. And then some parts, like 53.830, 6 may be too prescriptive in terms of what kind of 7 program is required rather than what are the 8 attributes you're trying to achieve.
9 The fourth point -- and I'm scurrying 10 along because I know of the lack of time -- the 11 operational objectives seems broader than needed. The 12 transition to operation, as was mentioned, seems to be 13 a little excessive with additional documentation 14 required. It seems to be broader than needed and 15 53.7(c) and 53.720(a), 2 and 4. I'm open on the 16 details, but we can fly, as needed.
17 Fifth point, considering the new paradigms 18 for human systems are appropriate. Human systems 19 interactions are changing with new designs; additional 20 advances; additional twins, as we just mentioned; and 21 in monitoring technology. And the NRC is 22 appropriately considering the need for new approaches 23 for operations of autonomous small reactors with some 24 passive safety features. As discussed this morning, 25 some developers believe that on-site personnel may not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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273 1 be needed to safely operate their reactors.
2 So to finish, just touch on a couple of 3 points that we've made previously to Subparts B and C, 4 we do not agree with this second version of Subparts 5 B and C. There's many areas where the language is 6 increasing regulatory burden, and the major issues the 7 industry brought up really weren't changed in B and C.
8 We do find categories or tiers one and two confusing, 9 that PRA insights are what are important for design, 10 not particularly results. We don't believe that PRA 11 should be elevated to a compliance tool. It's 12 important to do, it's useful, but not as a compliance 13 tool necessarily. And the phase for a simplified 14 approach to a PRA, the graded PRA, has merit, and the 15 NRC is thinking about that and that's good.
16 ALARA is an important concept, as you 17 said, for the practice, but we don't think it should 18 be included as a formal regulation, in part because of 19 its subjectivity and complexity in the design phase.
20 Defense-in-depth also, as we mentioned, is an 21 important design philosophy supporting safety case but 22 thinks it should be in guidance or out of the 23 regulations. As implied in the definitions, you could 24 use programmatic controls rather than just another 25 piece of equipment to achieve defense-in-depth.
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274 1 The protection standard is important, and 2 we disagree with having it dropped from Subpart B. We 3 do support consensus codes and standards. I think it 4 is important for using the international supply chain 5 and allowing people to be effective in how they're 6 developing that. And, finally, Part 53 can have 7 predictability and flexibility.
8 So I'll end with the fact there's a lot of 9 work. We appreciate your insightful ideas and issues 10 you've raised, but, as of now, industry does not find 11 the language as written to be particularly something 12 they would want to use. So thanks for your 13 opportunity to comment.
14 CHAIR BLEY: Thank you, Mr. Draffin. We 15 will read your -- well, they're in the transcript.
16 Thank you for your comments.
17 Thomas, can we get the public line open?
18 While he's getting it open, anybody on the public line 19 who would like to make a comment? As soon as the line 20 is open, please state your name and, if you have an 21 affiliation, that, too, and make your comment. Do we 22 know if the line is open? Thomas or Matika or Derek, 23 if you can tell me you're working on it at least, I'd 24 appreciate it.
25 MR. DASHIELL: The public line is open.
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275 1 CHAIR BLEY: Thank you. At this time, 2 anyone listening in on the bridgeline who would like 3 to make a comment, please state your name and give us 4 your comment, please. Hearing none, let's close the 5 public line again.
6 At this time, any final comments from the 7 members would be appreciated. Just open your mike and 8 talk if you want to make a comment.
9 MEMBER BROWN: This is Charlie again. I 10 thought Vicki made a good comment relative to all the 11 flexibility adding a lot of extra burden on the staff.
12 CHAIR BLEY: Thanks, Charlie. Anybody 13 else?
14 MEMBER BROWN: Good observation. No, I 15 said everything I wanted to say in the meeting.
16 CHAIR BLEY: Jose?
17 MEMBER MARCH-LEUBA: Yes. I think this is 18 a bad rule and it should be reconsidered. A rule that 19 allows a reactor that has no operators, local or 20 remote, is a bad rule. A rule that would allow not 21 having a protection system, safety grade or not, is a 22 bad rule.
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276 1 shall have operators on your power plant. And then if 2 you have some special characteristics, you can ask for 3 an extension. But I think the flexibility this rule 4 is trying to produce is something that is making my 5 brain blow up.
6 This is a bad rule, and the staff should 7 reconsider doing it properly. Thank you.
8 CHAIR BLEY: Thanks, Jose Greg, did you 9 have a comment?
10 MEMBER HALNON: I don't know if I can 11 follow that or not. I think I was going to just 12 comment on the burden on the staff translates and 13 magnifies to the burden on the industry. So if, in 14 fact, we made an assessment that shows burden is 15 increased, we need to reassess those areas.
16 CHAIR BLEY: Thanks. Anybody else?
17 MEMBER PETTI: Dennis, this is Dave. I 18 just wanted to say that I thought the recommendation 19 to have a briefing on LMP and the associated now reg 20 guides would be worthwhile because I think there's a 21 lot of discussion here and examples given that would 22 not, that would violate the procedures and the 23 approaches described in those reg guides.
24 CHAIR BLEY: Thanks. Anybody else? Well, 25 I'd like to thank the staff for the presentation and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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277 1 all the discussions, the long one this morning on the 2 white paper. We're glad to see some of the beginnings 3 of some of the more general guidance. We need to 4 study that more, at least I do, and understand it 5 better and see how it will link together with the 6 rule.
7 So thanks to everyone for our 8 participation. At this time, this meeting is 9 adjourned.
10 (Whereupon, the above-entitled matter went 11 off the record at 6:08 p.m.)
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U.S. Nuclear Industry Council (USNIC) Comments NRC Advisory Committee on Reactor Safeguards (ACRS) Meeting Future Plan Designs Subcommittee Preliminary Rule 10 CFR Part 53 20 May 2021 I am Cyril Draffin, Senior Fellow for Advanced Nuclear at the U.S. Nuclear Industry Council.
- 1) Promised reduction in operational burden not apparent
- NRC staff promised that Subpart F would enable significant reduction in operational burden as compared to Parts 50/52--- and increased burden in Subparts B&C was necessary to decrease operational burden.
- Having now seen portions of Subpart F, not clear what benefits are. Preliminary language seems to result in increased burden, limits flexibility, and provides no enhancement to safety in comparison with Parts 50/52. The only apparent major benefit to Part 53 is limiting need to seek exemptions to large LWR specific requirements.
- At last public Part 53 meeting, industry asked NRC to explain if regulations NRC were adding were needed for safety
- Part 53 process not addressing how to limit staff open-ended application reviews (with guardrails other than established back-fit), especially when adding regulatory burden
- 2) Subpart F (at least in its current form) may not be necessary
- Some believe the current Subpart F approach is flawed, and recommend a retraction of proposed Subpart F language and restart from a clean sheet of paper.
- 3) Subpart F Programs are redundant to existing programs and do not need to be included in a larger envelope program to be successful.
- There is not mixed reaction from industry on Facility Safety program (53.890), which NRC took from the railroad industry. Industry is not supportive of this new regulation on top of existing NRC programs.
- Integrity Assessment Program (53.850) is new and duplicative and seemingly anunnecessary regulation
- Too prescriptive (e.g. 53.830 (a) that lists specific programs rather than applicable attributes)
- 4) Subpart F Operational Objectives is broader than needed
- Transition to Operation (53.710) has additional requirement from 50.52
- Broader than needed (e.g. 53.710(c), 53.720 (a)2&4 , can give examples if you wish)
- 5) Considering new paradigms for Human- Systems are appropriate Human-system interactions are changing with new designs and digital advances in monitoring technology. NRC is appropriately considering need for new approaches for operations of autonomous smaller reactors with more passive safety features. As discussed this morning, some designers believe on-site personnel may not be needed to safety operate their reactors.
1
Reminder of prior comments
- We do not agree with 2nd iteration of Subparts B and C. There are many areas where language is increasing regulatory burden over Part 50 or 52, and NRC made no major changes to address industry concerns in updating Subparts B and C.
- Tier or categories 1 and 2 are confusing, with opportunities for unintended consequences.
- PRA insights are what are important for design, not specific numerical results. We dont believe the PRA should be elevated to a compliance tool as part of the application, especially for a construction permit. Phased or simplified approach to PRA and a graded PRA for a range of licensing paths and technologies has merit, but have not seen in guidance.
- ALARA is an important concept and good practice, but do not believe ALARA should be included in Part 53 formal regulation, in part because of its subjectivity and complexity in the design phase.
- Defense in Depth is important as a design philosophy in supporting an adequate safety case.,
but DID details should be described in guidance, not regulation. Could use programmatic controls rather than another piece of equipment to achieve DID.
- Adequate Protection Standard is important and we disagree with dropping formal reference to reasonable assurance of adequate protection standard.
- Support use of consensus codes and standards
- Part 53 rule can have predictability, as well as flexibility.
2
Advisory Committee on Reactor Safeguards (ACRS)
Future Plant Designs Subcommittee 10 CFR Part 53 Licensing and Regulation of Advanced Nuclear Reactors Subparts A and F Preliminary Proposed Rule Language and White Paper on Human Factors May 20, 2021 1
Agenda 9:30am - 9:40am Welcome / Introductions / Logistics / Goals 9:40am - 11:15am White Paper - Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors 11:15am - 1:00pm Subpart A - General Provisions 1:00pm - 2:00pm Lunch Break 2:00pm - 4:00pm Subpart F - Section 53.700, Operational Objectives and Controls on Equipment 4:00pm - 5:30pm Subpart F - Section 53.800, Programs 5:30pm - 6:00pm Discussion/Closing Remarks 2
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 Treatment Considerations Factors Programs Security, EP Other Plant/Site (Design, Construction, Configuration Control)
Clarify Subpart A Analyses (Prevention, Mitigation, Compare to Criteria) Controls General Provisions and Plant Documents (Systems, Procedures, etc.) Distinctions Between Subpart J Admin & Reporting LB Documents (Applications, SAR, TS, etc.) Subparts H & I 3
NRC Staff Engagement Plan ACRS Interactions Framework Safety Criteria Design Siting Construction Operations Decommissioning Licensing General/Admin Sept 20 Nov 20 Dec 20 Jan 21 Feb 21 Mar 21 Apr 21 May 21 Jun 21 Jul 21 Consolidated Technical Sections Aug 21 Sept 21 Consolidated Technical Sections Oct 21 Nov 21 Consolidated Rulemaking Package Dec 21 Jan 22 ACRS Full Committee Feb 22 Mar 22 Apr 22 May 22 Draft Proposed Rulemaking Package to the Commission Jun 22 Jul 22 Aug 22 Sept 22 Oct 22 Concept/Introduction Discussion Interim Staff Resolution 4
Risk-Informed and Performance-Based Human-System Operation Considerations for Advanced Reactors NRR - Division of Advanced Reactors and Non-power Production and Utilization Facilities NRR - Division of Reactor Oversight May 20, 2021
Agenda
- Background
- Nexus to 10 CFR 53
- White Paper Considerations - Overview
- Next Steps
- Questions/Comments 6
=
Background===
- Nuclear Energy Innovation and Modernization Act (NEIMA) requires the NRC to complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use for commercial advanced nuclear reactors by 2027.
- NRC currently developing 10 CFR 53: "Licensing and Regulation of Advanced Nuclear Reactors."
Supporting Guidance
- In some cases, guidance in support of proposed rule may be the driving factor to meet technology-inclusive, performance-based criteria that define a modern risk-informed graded approach.
- Development of key guidance to 10 CFR 53 - Draft White Paper guidance to be discussed today was developed under that premise.
7
Key Guidance White Paper: Risk-Informed and Performance-Based Human-System Operation Considerations for Advanced Reactors*
- Supports Subpart F: Operations
- Presented as Key Guidance to ACRS subcommittee on March 17, 2021
- Topics address diverse and novel operational characteristics, including automation of operations, staffing and qualifications of operations personnel, evolution in control room concepts, and the application of human factors engineering (HFE).
- Draft white paper released to begin external stakeholder interactions 8
- Released March 25, 2021 (ADAMS Accession No. ML21069A003)
Nexus to 10 CFR Part 53 Subpart F Operations preliminary rule language
- Primary purpose of this paper is to inform and support Part 53 rulemaking by proposing guidance related to operations (subpart F).
- Secondary goal is to facilitate the consistent treatment of advanced reactor applications that are received prior to Part 53 being finalized.
- Goal is technology-inclusive, scaled review approaches to the extent practical.
9
White Paper Considerations Main Topics
- The regulatory framework for advanced reactors should be capable of addressing novel operational concepts for a wide variety of advanced reactor technologies.
- Some advanced reactor designs may present very low radiological risk and requirements in the current regulatory framework for operation of large light water reactors (LWRs) may be unnecessary for reasonable assurance of safety.
- The development of a risk-informed, performance-based, and technology-inclusive regulatory framework that appropriately considers the role of humans and human-system integration is warranted for advanced reactors.
10
White Paper Considerations New Technologies and Safety Characteristics
- The preceding decades have witnessed evolutionary changes in areas like passive safety and modular construction.
- Technologies that are under various stages of development include small modular reactors (SMRs), non-LWRs, and fusion-based technologies.
- Such technologies warrant careful consideration of design attributes that represent departures from large LWR designs.
- The NRC recognizes the desirability of attributes such as simplified safety features of a passive or inherent nature, reductions in required human actions, incorporation of defense-in-depth (DID), and minimization of the risks associated with severe accidents in advanced reactor designs.
11
White Paper Considerations Smaller Source Term Sizes and Reduced Accident Consequences
- Advanced reactors could vary in size from very large to very small; such variations are expected to have potential implications for both source term sizes and accident consequences.
- Accident source terms can serve as a measure of the efficacy of mitigation features.
- Advanced reactor designs may present low potential accident consequences.
- Limiting the hazard posed by a reactor facility reduces the potential for accident consequences and is the most reliable means of ensuring safety.
12
White Paper Considerations Passive Safety Features and Inherent Characteristics
- Passive safety features and inherent safety characteristics can influence the role of personnel at advanced reactors facilities.
- Passive safety features tend to place humans into a DID role.
- While passive safety features can still fail under certain conditions, inherent safety characteristics can be considered to be absolutely reliable.
- The incorporation of inherent safety characteristics, passive safety features, and automated safety systems influence the concept of operations and can affect the emphasis placed on the HFE aspects of an application review.
13
White Paper Considerations Automation of Plant Operations
- Automation is implemented in levels that span from manual to autonomous operation.
- Autonomous operation (full automation) has the potential to support unattended reactor operations.
- Even in an autonomous design, there may still exist a need for humans to implement manual operations under certain circumstances, such as for DID.
- Automation generally enhances operational performance, however other operational effects must be considered as well (e.g., operators losing manual control proficiency).
14
White Paper Considerations Load Following
- Advanced reactor designers may desire to incorporate load-following capabilities into their designs.
- Load-following where a grid control center can directly adjust plant output is not currently practiced by commercial nuclear facilities in the U.S. because the practice is precluded by existing NRC regulations; however, that is not the case internationally.
- A nuclear power plant needs to be able to refuse load-following requests when complying with such requests would violate technical specifications (TS) or result in unsafe conditions.
15
White Paper Considerations DID and Advanced Reactor Operations
- The NRC has had a long-standing policy of ensuring that DID is incorporated into the design and operation of nuclear power plants.
- The key principles of note within the present context are that DID approaches should:
oNot rely solely on a single operational feature oNot rely excessively upon human actions (or programs).
- The role of humans in DID at advanced reactors is an area that may need further development.
16
White Paper Considerations Staffing
- The NRC staff previously recognized the limitations of the prescriptive requirements of 10 CFR 50.54(m) and developed NUREG-1791 in order to allow increased flexibility to LWRs and provide guidance for assessing exemptions to the regulations in 10 CFR 50.54(m).
- Licensing future applications for advanced reactors by exemption from prescriptive requirements may not be a practical long-term regulatory framework.
- An alternative means that is not reliant upon NUREG-1791 may be beneficial, especially if such a means were to rely upon analyses that can be scaled with the risk of the facility.
17
White Paper Considerations Operator Licensing
- The NRC has been licensing reactor operators since the 1950s.
- The Atomic Energy Act of 1954, as amended (AEA), requires the NRC to prescribe uniform operator licensing conditions.
- All license exams are approved and administered by the staff.
- Advanced reactor operational concepts may not align well with the existing power reactor operator licensing framework.
- Examples of appropriate changes for advanced reactors may include allowances for varying licensing examination scope on a facility-specific basis and modified simulator requirements.
- A revised approach to operator licensing should flexibly and efficiently address a wide variety of advanced reactor designs.
18
White Paper Considerations Shift Technical Advisor (STA) Position
- Staffing at power reactors also includes the STA position.
Unlike licensed operators, STA requirements are primarily rooted in Commission policy, and not regulation or statute.
- The current policy is that, on each shift, there should be at least one person on duty who has a degree in physical science, engineering, engineering technology, or a PE license.
- The function of this person is to provide independent engineering expertise, accident assessment, and technical advice to the main control room operators.
- The elimination of the STA position at a power reactor facility would be a departure from existing Commission policy, as well as from longstanding agency and industry practice.
19
White Paper Considerations Training
- The Nuclear Waste Policy Act of 1982, as amended (NWPA), directs the NRC to establish regulations for the training and qualifications of nuclear power plant operators, supervisors, technicians and other operating personnel.
- The NWPA also directs the NRC to establish requirements for simulator training, requalification examinations, operating tests, and instructional requirements.
- The Systems Approach to Training (SAT) plays a central role in current nuclear training and qualification programs.
- The SAT process is generic in nature and can be adapted to any reactor technology, including those associated with essentially any foreseeable advanced reactor designs.
20
White Paper Considerations Human Factors Engineering
- The application of HFE in the design of nuclear power plant control rooms is required under existing post-Three Mile Island regulations.
- Current HFE reviews typically focus on the human-system interfaces (HSIs) located within control rooms.
- Moving forward should include examining how HFE reviews can be implemented most effectively for advanced reactors.
- New approaches, such as the application of scalable HFE review processes and thinking beyond the confines of traditional control rooms, should be considered.
- A Concept of Operations can be valuable in gaining the design understanding necessary to conduct appropriate HFE reviews.
21
White Paper Considerations The Evolving Concept of the Control Room
- Some advanced reactor facilities may wish to not utilize traditional control rooms in their designs.
- Requirements addressing matters associated with control rooms will need to be revisited in Part 53 with an understanding that the functions involved may become decentralized in an advanced reactor facility.
- HFE requirements will essentially need to be able to follow important functions if they are relocated outside of a traditional control room.
- It may also be necessary to account for the potential emergence of functions that have no precedent within traditional control rooms as well.
22
White Paper Considerations Additional Organizational Considerations; No Licensed Operators
- For a fully autonomous advanced reactor design, it should be noted that the existing regulatory framework also assigns certain responsibilities and authorities to licensed operators. A key example are the requirements of 10 CFR 50.54(x) and (y) for departures from license conditions.
- Beyond this, there are numerous other licensed operator administrative responsibilities and authorities that are both important to safety and derived from regulatory requirements; such responsibilities and authorities would need to be addressed as well.
oThese include compliance with TSs, operability determinations, NRC notifications, emergency declarations, and radiological release limit compliance.
23
White Paper Considerations White Paper alignment with 10 CFR 53
- The rule may recognize that staffing, training, operator licensing, and human factors are interrelated areas; diverse advanced reactor technologies necessitate integrating the review of these areas under a flexible approach.
- The rule may account for varying accident consequences in assessing staffing issues.
- The rule may require an HFE program adequate to ensure that personnel can understand plant status, take action to ensure safety, and perform other important technical and administrative functions with safety implications.
oHuman roles associated with the management and availability of plant-specific safety functions will need to be taken into account when considering HFE requirements.
24
White Paper Considerations White Paper alignment with 10 CFR 53 (contd)
- The rule may account for designs that do not utilize traditional control rooms.
- The rule will need to ensure that the operator licensing process accomplishes the following:
oCompliance with applicable statutory requirements (i.e.,
AEA and NWPA);
oConformance with accepted testing standards; oFacilitation of consistent and reliable licensing decisions by the NRC; oEfficient use of NRC and vendor/facility licensee resources; oProvision of reasonable assurance that operators will be able to manage plant-specific safety functions.
25
White Paper Considerations White Paper alignment with 10 CFR 53 (contd)
- The rule may allow for consideration of innovative features intended to make new designs safer, while also accounting for uncertainties associated with new approaches.
- The rule may, in a non-prescriptive manner, require staffing levels needed to support safe operation and allow for the possibility of demonstrating that no human presence is necessary.
oThe rule may also prescribe minimal requirements that must be met to not use licensed operators at all.
- The rule may ensure that advanced reactor DID approaches do not rely exclusively upon a single operational feature or rely excessively upon human actions.
26
White Paper Considerations White Paper alignment with 10 CFR 53 (contd)
- The rule may account for the possibility of load-following where the load changes themselves are controlled externally from a grid control center.
- The rule will need to require that sufficient information be submitted to facilitate reviews as outlined within these goals.
Examples of such information may include the following:
oThe Concept of Operations for the design; oFunctional Requirements Analyses describing the features, systems, and human actions relied upon for safety; oA staffing plan, with supporting HFE-based analyses; oA SAT-based training program for relevant personnel.
27
White Paper Considerations Solutions: Scalable HFE Reviews
- The NRC staff has initiated work under contract with Brookhaven National Laboratory (BNL) to develop a method for scaling the scope and depth of HFE reviews for advanced reactors.
- The objective of this effort is to enable the staff to readily adjust the focus and level of staff HFE review efforts based upon factors such as risk insights and the unique characteristics of the design or facility operation.
- In the interim, the NRC staff also has the ability to adjust the scope of a NUREG-0711 HFE review on a case-by-case basis should a given license application warrant a reduction in the scope of an HFE-area technical review.
28
White Paper Considerations Solutions: Staffing Facilities Without Need for Licensed Operators To justify not using licensed operators, the applicant must demonstrate that adequate protection of the public health and safety will exist in the absence of any operator action for preventing or mitigating accidents. The following are examples of criteria that could potentially be used for assessing the acceptability of an advanced reactor design operating without using any licensed operators:
- 1. The accident analysis must demonstrate that radiological consequence criteria will be met without reliance on human actions for event mitigation, DID, or safe shutdown.
29
White Paper Considerations Solutions: Staffing Facilities Without Need for Licensed Operators (Contd)
- 2. Safety of the design should rely upon inherent safety characteristics. Absent an operator presence, the absolute reliability of inherent safety characteristics would be key.
- 3. If not fully autonomous, the design should have sufficient autonomy to support safety without human action. If human action is needed for startup, it may be appropriate to:
- a. have a licensed operator conduct the reactor startup; or
- b. demonstrate the safety analyses bound all postulated errors by a non-licensed operator during a reactor startup (warranted because a non-licensed startup operators abilities would not provide the NRC staff with the same degree of assurance as those of a licensed operator).
30
White Paper Considerations Solutions: Staffing Facilities Without Need for Licensed Operators (Contd)
- 4. License conditions could be established for the facility so that those administrative responsibilities with safety implications (e.g., Tech Spec compliance) that would otherwise have been allocated to licensed operators are reassigned (e.g., to a designated facility manager position).
- 5. For the STA position, the staff would need to engage with the Commission on a proposed departure from policy should an applicant propose a staffing plan that does not include on-shift engineering expertise. A key consideration would likely be the applicants ability to demonstrate that the results of staffing-related analyses remain adequate in the absence of the on-shift engineering expertise provided by an STA.
31
White Paper Considerations Solutions: Scalable Approach to Operator Licensing Requirements
- A flexible process that advanced reactor vendors and licensees could use to develop an operator licensing exam program for their sites might consist of the following:
- 1. Job Task Analyses to identify knowledge, skills, and abilities related to the facilitys licensed operator role.
- 2. Training and evaluation methods would be selected using a SAT process, including determining exam composition.
- 3. A vendor or licensee would pilot the proposed exam.
- 4. Exams would be reviewed and administered by the NRC.
oA potential option would be for vendors and licensees to also administer their own license examinations.
32
White Paper Considerations Solutions: Concept of Operations
- There is currently no regulation requiring applicants to provide a Concept of Operations as part of applications.
- New designs will likely conceive of radically different Concepts of Operations for which the staff may have little or no prior understanding. Therefore, there may be a need to explicitly make the Concept of Operations a part of the content of applications under the proposed Part 53 rule.
- A description of the Concept of Operations will help the NRC staff to avoid confusion, understand and confirm to what extent a design relies on the humans for safe operation, determine the appropriate scope of the staff review, and reduce the need for Requests for Additional Information.
33
White Paper Considerations Solutions: Staffing Analyses
- It may be appropriate for applicants to propose their own alternative staffing models. At a minimum, an HFE-based staffing analysis of sufficient scope and depth to allow for the NRC staff to adequately assess the acceptability of the proposed staffing levels would be needed.
- Alternative staffing models for advanced reactor applicants could be informed by the existing process of NUREG-1791.
- It may also be appropriate for the Part 53 rule to provide a prescriptive staffing model as an option for applicants that prefer not to conduct the staffing analyses needed to support an alternative, flexible staffing model.
34
White Paper Considerations Solutions: HFE Programs
- Applications are likely to need to contain specific information related to an HFE program and the related analyses (e.g.,
designs of control room HSIs or proposals for alternative staffing models would be expected to be informed by HFE principles).
- Part 53 may require advanced reactor applications to address the incorporation of state-of-the-art HFE principles more comprehensively than existing regulations require at present.
An advanced reactor HFE program should be adequate to ensure that humans can perform the full range of tasks necessary to ensure the continued availability of plant-specific safety functions; this may also extend to maintenance and testing activities related to plant safety functions.
35
Next Steps The objectives are being addressed using the approach below Draft Part 53 Rule Language Regulatory Guidance
- Prescriptive criteria for not
- Technology-Inclusive OL - efforts using licensed operators to create an ISG are underway via NRR/RES working group;
- Content of Applications; to national lab support is being address inclusion of the pursued.
Concept of Operations and
- Flexible Staffing - efforts to the HFE-related information create an ISG being coordinated needed to for advanced in NRR and RES.
reactor applications.
- Scalable HFE Reviews - efforts to
- HFE and Operating License create ISG underway via (OL) Integration - an overall NRR/RES working group with objective in development of national lab support (BNL).
both rule language and guidance. 36
White Paper Considerations Final Thoughts
- Draft concepts in white paper meant to solicit feedback on key areas of advanced reactor operations. Final scope and guidance outcome will be determined in coming months.
- Well-defined and unambiguous criteria is critical for a performance-based, graded approach related to Operations.
Leveraging results of existing methodologies as part of an application (such as Licensing Modernization Project (LMP),
maximum hypothetical accident (MHA), deterministic insights, probabilistic risk assessment (PRA) insights, etc.) will be explored.
- White Paper concepts are intended for future 10 CFR 53 applicants, however, NRC may use the concepts described to inform proposed exemptions from Part 50/52 requirements for near-term applicants. 37
Next Steps White Paper: Risk-Informed and Performance-Based Human-System Operation Considerations for Advanced Reactors*
- Additional stakeholder interactions will follow in coming months. Industry feedback expected July 2021 timeframe.
- The NRC staff are evaluating resources/schedule to identify what areas of the guidance to prioritize.
- Initial discussions with national labs to assist with expertise
- White Paper guidance in final form will support development of proposed 10 CFR 53 rule language and regulatory guidance.
- Final form of guidance is still being evaluated. For example:
oInterim Staff Guidance oRegulatory Guide 38
- Released March 25, 2021 (ADAMS Accession No. ML21069A003)
Thank Questions/Comments?
You!
Jesse Seymour, NRR/DRO/IOLB/HFT Jesse.Seymour@nrc.gov
Human-System Operation Considerations Discussion 40
General Provisions and Definitions - Subpart A 41
Part 53 General Layout
- Subpart A, General Provisions
- Subpart B, Technology-Inclusive Safety Objectives
- Subpart C, Design and Analysis
- Subpart D, Siting Requirements
- Subpart E, Construction and Manufacturing Requirements
- Subpart F, Requirements for Operation
- Subpart G, Decommissioning Requirements
- Subpart H, Applications for Licenses, Certifications and Approvals
- Subpart I, Maintaining and Revising Licensing Basis Information
- Subpart J, Reporting and Administrative Requirements 42
10 CFR Part 53 Subpart A Layout
- § 53.010 - Scope
- § 53.020 - Definitions
- § 53.040 - Written Communications
- § 53.050 - Deliberate Misconduct
- § 53.060 - Employee Protection
- § 53.070 - Completeness and Accuracy of Information
- § 53.080 - Specific Exemptions
- § 53.090 - Combining Licenses; Elimination of Repetition
- § 53.100 - Jurisdictional Limits
- § 53.110 - Attacks and Destructive Acts
- § 53.120 - Information Collection Requirements: Office of Management and Budget Approval 43
Subpart A - § 53.020 Definitions
- Advanced nuclear plant o Advanced nuclear plant [or facility] means a utilization facility consisting of one or more advanced nuclear reactors [as defined in NEIMA] and associated co-located support facilities, which may include one or more reactor modules, [using nuclear fission, nuclear fusion, or accelerator-driven reactor technologies] that are used for producing power for commercial electric or other commercial purposes. The advanced nuclear plant includes the collection of sites, buildings, radionuclide sources, and structures, systems, and components for which a license is being sought under this part.
- Definition of advanced nuclear reactor (NEIMA) o a nuclear fission or fusion reactor, including a prototype plant (as defined in sections 50.2 and 52.1 of title 10, Code of Federal Regulations (as in effect on the date of enactment of this Act)), with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of this Act, including improvements such as (A) additional inherent safety features; (B) significantly lower levelized cost of electricity; (C) lower waste yields; (D) greater fuel utilization; (E) enhanced reliability; (F) increased proliferation resistance; (G) increased thermal efficiency; or (H) ability to integrate into electric and nonelectric applications.
- SECY-20-0032 o The staff interprets NEIMAs definition of an advanced nuclear reactor, which states that such a reactor will have significant improvements compared to commercial nuclear reactors under construction as of January 14, 2019, as excluding Generation III+ designs from the definition because the AP1000 reactors were under construction at the time of NEIMAs enactment.
44
Subpart A - § 53.020 Definitions
- Consensus code or standard o 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.
45
Subpart A - § 53.020 Definitions
- End state o means the set of conditions at the end of an event sequence that characterizes the impact of the sequence on the plant or resulting releases of radionuclides to the environment. In most probabilistic risk assessments, end states typically include success states (i.e., those states with negligible impact) and release categories.
- Event sequence o means a postulated initiating event defined for a set of initial plant conditions followed by system, safety function, and operator successes or failures, and terminating in a specified end state depending on the system, safety function, and operator successes and failures (e.g., prevention of release of radioactive material or release in one of the reactor-specific release categories). An event sequence may include many unique variations of events (e.g., minimal cut sets) that are similar in terms of how they impact the performance of safety functions along the event sequence.
46
Subpart A - § 53.020 Definitions
- Normal plant operation or normal operation o means operations that are expected to occur during planned operations or shutdown of the reactor.
- Licensing basis events (LBEs)
- mean a collection of event sequences considered in the design and licensing of the advanced nuclear plant. LBEs are unplanned events and include AOOs, unlikely event sequences, very unlikely event sequences, and DBAs.
- Design basis accidents (DBAs)
- mean postulated event sequences that are used to set functional design criteria and performance objectives for the design of safety-related structures, systems, and components. DBAs are a type of licensing basis event and are based on the capabilities and reliabilities of safety-related structures, systems, and components needed to mitigate and prevent event sequences, respectively.
- Anticipated operational occurrences (AOOs) o mean anticipated event sequences expected to occur one or more times during the life of a nuclear power plant. An event sequences with a mean frequency of 1x10-2/plant-year and greater is an AOO. AOOs take into account the expected responses of all SSCs within the plant, regardless of safety classification. AOOs are a type of licensing basis event. [Based, in part, on Appendix A to part 50.]
47
Subpart A - § 53.020 Definitions
- Unlikely event sequences o mean event sequences that have estimated frequencies below the frequency of AOOs. Unlikely event sequences are a subset of LBEs. [For example, within the licensing modernization project, this would equate to design basis events with a frequency range of between 1x10-4 and 1x10-2 per plant year with an accounting for uncertainties.]
- Very unlikely event sequences o mean event sequences that have estimated frequencies well below the frequency of events expected to occur in the life of an advanced nuclear plant. Very unlikely event sequences are a subset of LBEs. [For example, within the licensing modernization project, this would equate to beyond design basis events with a frequency range of between 5x10-7 and 1x10-4 per plant year with an accounting for uncertainties.]
- The frequency ranges were incorrect in the publicly-released preliminary proposed rule language. The NRC staff have corrected the error in the definitions above and will update the definitions in a future iteration of subpart A, General Provisions.
48
Subpart A - § 53.020 Definitions
- Safety-related (SR) o means those SSCs and human actions that warrant special treatment and are relied upon to demonstrate compliance with the safety criteria in § 53.210(b).
- Non-safety related but safety significant (NSRSS) o means those SSCs and human actions that warrant special treatment and are not safety-related but are relied on to achieve defense-in-depth or perform risk-significant functions.
- Non-safety significant (NSS) o means those SSCs not warranting special treatment, are not safety-related, and are not relied on to achieve adequate defense-in-depth or to perform risk-significant functions.
49
Subpart A - § 53.020 Definitions
- Special treatment o means those requirements, such as measures taken to satisfy functional design criteria, quality assurance, and programmatic controls, that provide assurance that certain SSCs will provide defense-in-depth or perform risk-significant functions and that provide confidence that the SSCs will perform under the service conditions and with the reliability assumed in the analysis performed in accordance with § 53.450 to provide reasonable assurance of meeting the safety criteria in § 53.210(b) and § 53.220(b).
- Defense in depth o means inclusion of multiple independent and redundant layers of defense in the design of a facility and its operating procedures to compensate for potential human and mechanical failures so that no single layer of defense, no matter how robust, is exclusively relied upon. Defense-in-depth includes, but is not limited to, the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures.
50
Subpart A - § 53.020 Definitions
- Design features o means the active and passive structures, systems, or components and inherent characteristics of those structures, systems or components that contribute to limiting the total effective dose equivalent to individual members of the public during normal operations and prevent or mitigate the consequences of unplanned events.
- Inherent characteristic
- means an attribute of a design feature that has such a high degree of certainty in its performance that uncertainties need not be quantified.
- Functional design criteria o means requirements for the performance of SSCs. For safety-related SSCs, these criteria define requirements necessary to demonstrate compliance with first tier safety criteria in § 53.210(b).
For non-safety-related but safety-significant SSCs, these criteria define requirements necessary to meet the second tier safety criteria in § 53.220(b).
51
Subpart A - § 53.040 - § 53.120 Other Administrative Requirements
- Subpart A consistent with Part 50 and currently includes bracketed references to existing requirements in Parts 50, 52, etc.
- Intending to develop Part 53 with largely no pointers to Parts 50 and 52; this will require copying and pasting the Part 50 or 52 language into Part 53 instead of using pointers.
- Subpart A will include many pointers to other sections of Part 53 that will be added in future iterations.
52
Subpart A - Stakeholder Feedback
- Advanced nuclear plant o Allow broad applicability for advanced reactors under Part 53.
- Design basis accident o Clarify the relationship between DBAs and unlikely event sequences and very unlikely event sequences.
- Consensus codes and standards o Questioned use of must in definition and asked if term includes general engineering standards, such as IEEE.
- Defense in depth o Update subpart B to match the definition in subpart A because the definition excludes the clause that no single feature should be exclusively relied upon.
- Safe, stable end state o Clarify if this term requires complete shutdown or demonstration of no further inventory release.
- Safety related, Non-safety related, Non-safety related but safety significant o Clarify these terms in a way that reduces overlap.
53
Subpart A - General Provisions Discussion 54
MEETING BREAK Meeting to resume in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 55
Subpart F - Operational Objectives 56
Part 53 General Layout
- Subpart A, General Provisions
- Subpart B, Technology-Inclusive Safety Objectives
- Subpart C, Design and Analysis
- Subpart D, Siting Requirements
- Subpart E, Construction and Manufacturing Requirements
- Subpart F, Requirements for Operation
- Operational Objectives
- Subpart G, Decommissioning Requirements
- Subpart H, Applications for Licenses, Certifications and Approvals
- Subpart I, Maintaining and Revising Licensing Basis Information
- Subpart J, Reporting and Administrative Requirements 57
Subpart F - § 53.700 Operational Objectives
- Licensee must:
o Define structures, systems, and components (SSCs) o Maintain capabilities and reliabilities of SSCs o Ensure plant personnel have adequate knowledge and skills to perform their assigned duties to support safety functions o Implement plant programs sufficient to ensure the performance of identified safety functions 58
Subpart F - § 53.710 Transition to Operation
- Prepare a transition plan from construction to operations o Demonstrate the SR and NSRSS SSCs are appropriately constructed and capable to perform o Plant personnel are appropriately licensed and trained to perform safety functions o Programs, procedures and controls are implemented to support the safety functions 59
Subpart F - § 53.720 Maintaining Capabilities
- Capabilities and reliability of SSCs, when combined with associated programmatic controls and human actions, provide reasonable assurance that the safety criteria defined in §§ 53.210(b) and 53.220(b) will be met.
- Paragraph (a) defines controls for SR SSCs (technical specifications).
- Paragraph (b) defines controls for NSRSS SSCs (reliability assurance and other special treatment).
60
Subpart F - § 53.720(a)
Technical Specifications (TS)
- TS required to define conditions or limitations on SSCs to fulfill safety functions (§ 53.230) and first tier safety criteria (§ 53.210(b))
- Inventories of radioactive materials
- Operating limits
- For each SSC classified as safety related
- Limiting Conditions for Operation
- Surveillance Requirements
- Design Attributes
- Administrative Controls
- Decommissioning
- First iteration does not include:
- Safety limits or associated limiting safety system settings
- Criteria for limiting conditions for operation
- Some stated preferences to use deterministic approaches may be better addressed within Part 50.
61
Subpart F - § 53.720(b)
Special Treatment of NSRSS SSCs
- Configurations and special treatments for NSRSS SSCs ensure capabilities, availabilities, and reliabilities to satisfy second tier safety criteria (§ 53.210(b)).
- Controls must:
- Identify authorities and processes for configuration changes
- Describe means by which special treatments for each NSRSS SSC will be provided and maintained
- Controls for NSRSS SSCs needed to implement a performance-based approach used to gain operational flexibilities and as part of methods that include replacing the single-failure criterion with a probabilistic (reliability) approach.
- Deterministic approaches with different supporting analysis, safety classification schemes, and design approaches (e.g., inclusion of the single failure criterion) may be better addressed within Part 50.
62
Subpart F - § 53.730 Maintenance, Repair and Inspection a) Develop a program to maintain and repair SR and Safety Significant SSCs.
b) Take appropriate corrective action when NSRSS SSCs do not meet special treatment requirements or performance goals.
c) Evaluate performance and preventive maintenance activities every 24 months.
d) Conduct risk-informed assessment of the impact and scope of any maintenance activities.
63
Subpart F - § 53.740 Design Control
- Assess the potential for adverse effects on safety, security, emergency preparedness (EP), operations, or other items related to plant safety during the design process and before implementing design or operational changes.
o Physical modifications, procedural changes, operator actions, maintenance activities, system reconfigurations, access modifications or restrictions, changes to the emergency plan and security plan or their implementation.
o Establish measures for the identification and control of interfaces among plant activities.
64
Subpart F, Operational Objectives -
Stakeholder Feedback
- In general, reduce the duplication of requirements for quality assurance, testing, and review language for interactions across sections.
- Clarify safety objectives (§ 53.200), operational objectives (§ 53.700), and programs (§ 53.800) to avoid scope creep in future licensing reviews.
- Allow automated plant surveillance and testing.
65
Subpart F - Operational Objectives Discussion 66
Subpart F - Programs 67
Part 53 General Layout
- Subpart A, General Provisions
- Subpart B, Technology-Inclusive Safety Objectives
- Subpart C, Design and Analysis
- Subpart D, Siting Requirements
- Subpart E, Construction and Manufacturing Requirements
- Subpart F, Requirements for Operation
- Programs
- Subpart G, Decommissioning Requirements
- Subpart H, Applications for Licenses, Certifications and Approvals
- Subpart I, Maintaining and Revising Licensing Basis Information
- Subpart J, Reporting and Administrative Requirements 68
Subpart F - § 53.800 Programs
- Programs must be provided for each advanced nuclear plant such that, when combined with associated design features and human actions, the plant will satisfy the first and second tier safety criteria defined in §§ 53.210 and 53.220.
69
Subpart F - § 53.810 Radiation Protection
- Implement a radiation protection program to limit occupational exposure in accordance with Part 20.
- Limit exposure to the public.
o Develop procedures and remedial actions in an Offsite Dose Calculations Manual (ODCM).
o ODCM Define methodology used in the calculation.
Contain radioactive effluent controls and environmental monitoring activities.
Annual Radiological Environmental Operating and Radioactive Effluent Release Reports.
70
Subpart F - § 53.820 Emergency Preparedness
- Develop and implement an EP Program for operations that is commensurate with the risks posed by the licensing basis events as analyzed in accordance with § 53.450.
o Staff is developing preliminary proposed rule language for Part 53 in coordination with activities related to the Emergency Preparedness for Small Modular Reactors and Other New Technologies rulemaking.
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Subpart F - § 53.830 Security Programs
- Develop and implement security programs:
o Information security o Physical security o Cyber security o Access authorization o Material control and accounting 72
Subpart F - § 53.840 Quality Assurance (QA)
- Develop and execute a QA program:
o Define duties and responsibilities for QA of SSCs o Written QA manual o Written procedures Qualified personnel Procurement Handling, shipping and storage of materials Testing and inspection Corrective action Document and configuration control Design control Record keeping Auditing o Document results of QA activities 73
Subpart F - § 53.850 Integrity Assessment Program (IAP)
- Develop and implement an IAP.
- Monitor, evaluate and manage:
o Effects of aging on SR and NSRSS SSCs whose failure could affect performance of safety functions.
o Cyclic and transient loads are maintained within applicable design limits.
o Degradation related to chemical interactions, operating temperatures, irradiation, and other environmental factors to ensure the capabilities and reliabilities of SSCs satisfy the functional design criteria of §§ 53.410(b) and 53.420(b).
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Subpart F - § 53.860 Fire Protection
- Develop and implement a fire protection plan:
o Identify responsible parties and authorities o Outline plans for fire protection, detection, suppression capability, and limitation of damage o Administrative controls, personnel requirements, and suppression activities o Means to limit damage to SR and NSRSS SSCs
- Specific features of program:
o SR and NSRSS SSCs designed to minimize effect of fires o Use noncombustible and fire-resistant materials wherever practical in facility o Appropriate capacity and capability o Design such that inadvertent operation does not impair SR SSCs 75
Subpart F - § 53.870 Inservice Testing (IST) and Inservice Inspection (ISI)
- Develop programs for ISI and IST:
o ISI and IST includes codes and standards supplemented by risk insights o Testing and frequency done to maintain reliability of SSCs o Documented procedures
- Perform baseline inspections prior to starting operations:
o Determine benchmarks o Develop acceptance criteria o Results provided to plant manager and determine need for corrective action 76
Subpart F - § 53.880 Criticality Safety Program
- The program must address the requirements in 10 CFR 70.24 for maintaining a monitoring system capable of detecting a criticality, having emergency procedures, and providing radiation protection for plant workers.
A topic for discussion is whether the alternatives to 10 CFR 70.24 provided in 10 CFR 50.68, Criticality accident requirements, are appropriate and useful in Part 53.
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Subpart F - § 53.890 Facility Safety Program (FSP)
- Establish an FSP using a risk-informed, performance-based process to proactively identify new or revised hazards and performance issues.
- Routinely evaluate potential hazards, operating experience, human actions, and programmatic controls.
- Consider measures to mitigate or eliminate the resulting risks.
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Subpart F - § 53.892 FSP Performance Criteria
- Take measures to protect public health and minimize danger to life or property as may be reasonably achieved when considering costs.
o Assess risk reduction measures related to radionuclide release during normal operation.
o Assess risk reduction measures for contributors to the overall cumulative risk from unplanned events.
- Certified designs/manufacturing licenses must also use change control from Subpart H.
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Subpart F - § 53.894 FSP Plan
- FSP must use written plan and address:
o Scope of facilities covered o How FSP will be implemented o How personnel will be trained in FSP o Risk-informed hazard management program o Technology assessment program o Internal facility safety program assessment
- Note that staff is looking at the possibility that some of the administrative details in the first iteration language might be addressed within guidance documents.
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Subpart F - § 53.896 Review, Approval, and Retention of FSP
- FSP plan is part of the application
- NRC to review/approve FSP plan
- Will define staff process for reviewing FSP plan changes and amendments 81
Subpart F - § 53.900 Procedures and Guidelines
- Integrated set of procedures and guidelines to maintain normal operations and respond to unplanned events
- Plan must address:
o Plant operations o Maintenance under § 53.730 o Program requirements under this subpart (e.g., radiation protection, QA, Integrity Assessment) o Emergency operating procedures if human intervention required o Accident management guidelines 82
Subpart F, Programs -
Stakeholder Feedback
- Industry stakeholder feedback focused on regulatory burden o Industry stakeholders urged NRC to reduce net burden with Part 53 rule. Concern new subpart F provisions will increase burden, including transition to operations (§ 53.710), integrity assessment programs (§ 53.850),
criticality safety program (§ 53.880), and facility safety program (§ 53.892).
o Industry stakeholders concerned regulatory burden makes rule unattractive.
- NGO stakeholder disagreed with industry feedback and suggested NRC staff revert to traditional notice and comment rulemaking process.
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Subpart F: Programs Discussion 84
Discussion on Previously Released Subparts and Integration of Subparts (if time allows) 85
Stakeholder Feedback on Previously Released Subparts
- Subpart B - Technology-Inclusive Safety Requirements o Some stakeholders commented that the NRC need not regulate NSS SSCs.
o A stakeholder asked the NRC to clarify how inherent characteristics could be credited for DID.
o Industry stakeholders characterized the NRCs second iteration of subpart B as less than transformational and reiterated concerns about the two-tiered safety construct.
- Subpart C - Design and Analysis Requirements o Industry stakeholders would prefer NRC adopt a graded approach to PRA.
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Stakeholder Feedback on Previously Released Subparts
- Subpart D - Siting Requirements o Industry stakeholders expressed support for the potential to collapse the exclusion area boundary and the low population zone to the site boundary.
- Subpart E - Construction and Manufacturing Requirements o Stakeholders asked the NRC to clarify the scope and applicability of a manufacturing license (ML) as it may relate to subcontractors, transportation, the completion status of the reactor, and linkage to the combined license.
o Industry stakeholders will provide input on potential business models for ML applicants and the timeframe for these business models so that the NRC can develop appropriate and timely requirements.
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Previously Released Subparts Discussion 88
Final Discussion and Questions 89
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 90
Acronyms and Abbreviations Advisory Committee on Reactor ISG Interim staff guidance ACRS Safeguards ISI Inservice testing AEA Atomic Energy Act IST Inservice inspection Anticipated operational AOO occurrence LBE Licensing basis event BNL Brookhaven National Laboratory LMP Licensing Modernization Project CFR Code of Federal Regulations LWR Light water reactor DBA Design basis accident MHA Maximum hypothetical accident DID Defense in depth ML Manufacturing license Nuclear Energy Innovation and EP Emergency preparedness NEIMA Modernization Act FSP Facility Safety Program NGO Non-governmental organization HFE Human factors engineering U.S. Nuclear Regulatory NRC HSI Human-system interface Commission Office of Nuclear Reactor IAP Integrity Assessment Program NRR Regulation 91
Acronyms and Abbreviations Non-safety related but safety Office of Nuclear Regulatory NSRSS RES significant Research NSS Non-safety significant SAT Systems Approach to Training Nuclear Waste Policy Act of SMR Small modular reactor NWPA 1982 Offsite Dose Calculations SR Safety related ODCM Manual Structures, systems, and OL Operating license SSC components PRA Probabilistic risk assessment STA Shift technical advisor QA Quality assurance TS Technical specifications 92
Background Slides 93
First Principles See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities 94
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 95
Part 53 Rulemaking The Part 53 Rulemaking Process*
- The process depicted in this schematic is unique to the Part 53 rulemaking and varies in some ways compared to a similar A Typical Rulemaking Process schematic available on the NRCs public website.
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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, 97