ML23221A041
| ML23221A041 | |
| Person / Time | |
|---|---|
| Issue date: | 07/12/2023 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-2470 | |
| Download: ML23221A041 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Docket Number: (n/a)
Location: teleconference Date: Wednesday, July 12, 2023 Work Order No.: NRC-2470 Pages 1-102 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (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.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 707TH MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 OPEN SESSION 9 + + + + +
10 WEDNESDAY 11 JULY 12, 2023 12 + + + + +
13 14 The Advisory Committee met via hybrid In-15 Person and Video-Teleconference, at 8:30 a.m. EDT, Joy 16 L. Rempe, Chairman, presiding.
17 18 COMMITTEE MEMBERS:
19 JOY L. REMPE, Chairman 20 WALTER L. KIRCHNER, Vice Chairman 21 DAVID A. PETTI, Member-at-Large 22 RONALD G. BALLINGER, Member 23 CHARLES H. BROWN, JR., Member 24 VICKI M. BIER, Member 25 VESNA B. DIMITRIJEVIC, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 1 GREGORY H. HALNON, Member 2 JOSE MARCH-LEUBA, Member 3 ROBERT P. MARTIN, Member 4 THOMAS E. ROBERTS, Member 5 MATTHEW W. SUNSERI, Member 6
7 ACRS CONSULTANT:
8 DENNIS BLEY 9 STEPHEN SCHULTZ 10 11 DESIGNATED FEDERAL OFFICIAL:
12 KENT HOWARD 13 14 ALSO PRESENT:
15 BUCK BARNER, Framatome 16 KURT CRYTZER, EPRI 17 LOIS M. JAMES, NRR 18 KENNETH GEELHOOD, NRR 19 KEVIN HELLER, NRR 20 SCOTT KREPEL, NRR 21 RYAN JOYCE, SNC 22 JOHN LAMB, NRR 23 JOHN LEHNING, NRR 24 MICHAEL MARKLEY, NRR 25 ALAN McGINNIS, Framatome NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 1 CHARLEY PEABODY, NRR 2 RADU POMIRLEANU, Westinghouse 3 DAVID RAHN, NRR 4 JIM SMITH, NMSS 5 GREGORY SUBER, NRR 6 MIKE WERNER, Westinghouse 7 BRANDON WISE, NRR 8 KENT WOOD, NRR 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 1 CONTENTS 2 PAGE 3 I. Opening Remarks by the ACRS Chairman 4 1. Opening Statement 5 5 2. Agenda and Items of Current Interest 6 6
7 II. EPRI Data Validation Topical Report 8 1. Remarks from Subcommittee Chair 9 9 2. Presentations and Discussions with EPRI 10 Representatives and NRC Staff 12 11 12 III. Vogtle License Amendment Request (LAR) on 13 Loading Lead Test Assemblies (LTAs) with 14 Increased Enrichment 15 1. Remarks from Subcommittee Chair 32 16 2. Presentations and Discussions with 17 Southern Nuclear Representatives and 18 NRC Staff 34 19 20 IV. ARITA-ARTEMIS/RELAP Integrated Transient 21 Analysis Methodology Topical Report 22 1. Remarks from Subcommittee Chair 62 23 2. Presentations and Discussions with 24 Framatome Representatives and NRC 25 Staff 65 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 1 P-R-O-C-E-E-D-I-N-G-S 2 8:30 a.m.
3 CHAIR REMPE: Good morning. It's 8:30 4 here on the East Coast. This meeting will now come to 5 order. This is the first day of the 707th Meeting of 6 the Advisory Committee on Reactor Safeguards.
7 I'm Joy Rempe, Chairman of the ACRS.
8 Other members in attendance are Ron Ballinger, Vicki 9 Bier, Charles Brown, Vesna Dimitrijevic, Greg Halnon, 10 Walt Kirchner, Jose March-Leuba, Bob Martin, Dave 11 Petti, Thomas Roberts, and Matthew Sunseri.
12 We do have a quorum. Today the committee 13 is meeting in-person and virtually. The ACRS was 14 established by the Atomic Energy Act and is governed 15 by the Federal Advisory Committee Act.
16 The ACRS section of the US NRC public 17 website provides information about the history of this 18 committee and documents such as our charter, bylaws, 19 Federal Register notices for meetings, letter reports, 20 and transcripts of all full and subcommittee meetings, 21 including all slides presented at the meetings. The 22 committee provides its advice on safety matters to the 23 Commission through its publicly available letter 24 reports.
25 The Federal Register notice announcing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
6 1 this meeting was published on June 21, 2023. This 2 announcement provided a meeting agenda, as well as 3 instructions for interested parties to submit written 4 documents or request opportunities to address the 5 committee. The Designated Federal Officer for today's 6 meeting is Mr. Kent Howard.
7 The communications channel has been opened 8 to allow members of the public to monitor the open 9 portions of the meeting. The ACRS is inviting members 10 of the public to use the MS Teams link to view slides 11 and other discussion materials during these open 12 sessions.
13 The MS Teams link information was placed 14 in the Federal Register notice and agenda on the ACRS 15 public website. We have received no written comments 16 or requests to make oral statements from members of 17 the public regarding today's session.
18 Periodically, the meeting will be open to 19 accept comments from participants listening to our 20 meetings. Written comments may be forwarded to Mr.
21 Kent Howard, today's Designated Federal Officer.
22 During today's meeting, the committee will 23 consider the following topics: EPRI Data Validation 24 Topical Report, Vogtle License Amendment Request on 25 Loading Lead Test Assemblies with Increased NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 1 Enrichment, and ARITA-ARTEMIS/RELAP Integrated 2 Transient Analysis Methodology Topical Report.
3 Note the portions of the EPRI and ARITA-4 ARTEMIS sessions may be closed as stated in the 5 agenda. I also want to note that the topic of 6 LANCR02, Lattice Physics Model Description Licensing 7 Topical Report, that was scheduled for tomorrow's 8 session will be discussed during our Friday planning 9 and procedures meeting.
10 A transcript of the open portion of 11 today's meeting is being kept. It's requested that 12 speakers identify themselves and speak with sufficient 13 clarity and volume so they can be readily heard.
14 Additionally, participants should mute themselves when 15 not speaking.
16 This morning I do have an item of note.
17 I want to recognize our two newest members, Bob Martin 18 and Tom Roberts, who are joining us for their first 19 full committee meeting open session.
20 Bob is a career nuclear safety specialist 21 through employment with BWXT, Babcock & Wilcox, 22 Framatome, and Siemens Power Corporation. Dr. Martin 23 was responsible for the development and regulatory 24 defense of several evaluation methodologies for the 25 design and safety of conventional and advanced nuclear NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 1 power plants.
2 Among these are the first applications of 3 best-estimate plus uncertainty methods to design-basis 4 LOCA for both fuel and containment evaluations and for 5 demonstrating plant resiliency to beyond-design-basis 6 severe accidents.
7 His experience includes employment at the 8 Idaho National Laboratory, where he worked in the area 9 of thermal system fluid modeling, and being the lead 10 editor and contributor on the testing of design-basis 11 accident analysis methods for light-water nuclear 12 power plants.
13 Tom Roberts has more than 40 years' 14 experience in the field of nuclear reactor systems and 15 safety. He spent 36 years as an engineer and 16 engineering manager at the Naval Nuclear Propulsion 17 Program Headquarters, working various roles in the 18 Instrumentation and Controls Division, and then 19 completing his career with 12 years as the Director of 20 Reactor Safety and Analysis.
21 Since his retirement from the Naval 22 Nuclear Power Program, Mr. Roberts has served as a 23 subject matter expert in programs for advanced reactor 24 development, including consulting and reactor safety, 25 and IMC for a transportable micro-reactor program and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 1 for a nuclear thermal rocket reactor concept.
2 I do hope that members and others will 3 join me in welcoming Rob and Tom, and in thanking 4 Alicia, Andrea, and Sandra for helping them through 5 the onboarding process.
6 At this time, I want to ask other members 7 if they have any opening remarks. Not hearing any, 8 I'd like to ask Member March-Leuba to lead us through 9 our first topic for today's meeting.
10 Jose?
11 MEMBER MARCH-LEUBA: Our first topic is 12 the EPRI Data Validation Topical Report, which we 13 listened to during the thermal-hydraulics subcommittee 14 on June 7th. We are going to have only an open 15 session. If required, we can always create a closed 16 link and discuss.
17 So please try not to mention any numbers, 18 which does tend to be proprietary. Other than that, 19 keep it general. If necessary we can close the 20 session, but I hope we don't.
21 I believe the staff is going to present.
22 Greg or Scott, are you going to present? Greg is 23 going to. Go for it.
24 MR. SUBER: Good morning and thank you.
25 I'd like to welcome the two new members of the ACRS.
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10 1 I'm glad that you got here so you can understand the 2 term hotter than July, as you will experience over the 3 next three days.
4 Once again, I'd like to thank the 5 committee for giving us the opportunity to discuss the 6 EPRI Technical Report for the use of data validation 7 and reconciliation methods for measurement uncertainty 8 recapture.
9 This topical report is a first-of-a-kind 10 topical report. It proves that it has the potential 11 to increase power for the US Nuclear Reactor fleet.
12 It's also a first of a kind in that it is an approval 13 of concept that is not typically the subject of a 14 topical report.
15 Therefore, the staff has conducted a very 16 thorough and in-depth technical review using 17 contractors and support from Sandia Labs, which 18 enhance the NRC Staff's capabilities.
19 EPRI, its members, and consultations do 20 their part in submitting a detailed topical report and 21 responded to a request for additional information in 22 a timely manner.
23 As a result of everybody's dedication and 24 efforts, the staff's safety evaluation for the topic 25 was issued on May the 6th. The staff concludes that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 1 reasonable assurance has been provided for the DBR 2 method.
3 It can be used to determine, one, core 4 thermal power, and also core thermal power 5 uncertainty. And thus, this report can be used by 6 licensees for requests for measurement uncertainty 7 power operations.
8 The NRC staff and EPRI are ready to 9 discuss the safety evaluation in detail in this 10 meeting. With that, I turn it back to the Chair.
11 MEMBER MARCH-LEUBA: Thanks, Greg. Lois 12 James is going to be our presenter; is that correct?
13 MR. SUBER: Yes, she is.
14 MEMBER MARCH-LEUBA: So Lois, we're going 15 to give you the microphone in just a moment. I'll 16 just give you a heads up that some members were not 17 present during the subcommittee and they have 18 questions about a topic that Josh Kaizer presented 19 very eloquently to the committee, which was the way he 20 introduced the risk-informed methodology to his 21 review.
22 Either you address it when you feel it's 23 proper or we will ask you the question at the end of 24 the presentation. I'm just giving you a heads up. So 25 Lois, go ahead.
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12 1 MS. JAMES: Thank you, Chair Rempe, Member 2 March-Leuba, and other members of the ACRS. As 3 stated, we are here today to discuss the staff's 4 review of EPRI, Electric Power Research Institute's 5 report on the use of data validation and 6 reconciliation for DBR methods for measurement 7 uncertainty recapture or MUR.
8 The staff has acknowledged that this is a 9 first-of-a-kind report and has the potential to have 10 a large impact on the nuclear industry.
11 As mentioned, my name is Lois James. I am 12 the project manager for the staff's review. In the 13 room or on the phone we have Scott Krepel, Nuclear 14 Systems Performance Branch Chief, and David Rahn, one 15 of the technical reviewers. We also have several 16 members of EPRI and their supporting staff to answer 17 any questions that the members may have.
18 The agenda for today, we will be providing 19 a short history of the staff's review. We'll mention 20 the purpose of the technical report. We'll discuss 21 what DBR is. We'll also talk about how DBR impacts or 22 is used for core thermal power. We will mention the 23 review scope limitations and then we will discuss the 24 staff's evaluation conclusions.
25 I'm not going to mention every item on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 1 project history. As you can see from the two slides, 2 we have provided the ML numbers. We thought this was 3 important to provide key dates and ADAMS reference 4 numbers for everyone, for members of the public and 5 for the committee.
6 You can see that the staff's review 7 started in early 2001. We brought Sandia Labs on to 8 support the staff's review during 2001. During 2002 9 and 2003, the staff issued the RAIs and conducted its 10 review.
11 In May of 2003, we issued our draft safety 12 evaluation for proprietary review. We shared that 13 with the committee. That's all the project history 14 that we were going to mention today.
15 MEMBER MARCH-LEUBA: Lois, this is Jose.
16 Can I ask you a question now?
17 MS. JAMES: Sure.
18 MEMBER MARCH-LEUBA: Since this is the 19 full committee, in my opinion, this is more for the 20 benefit of the public than for our questions. We 21 drilled you guys during subcommittee.
22 I wanted to ask you, I see that the 23 project history is two and a half years long from 24 issuance of the topical report and issuance of the 25 SER. Can you explain, was there any kind of problems NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 1 that need to be resolved with the RAIs? Or it's 2 simply lack of resources on the part of the staff?
3 Because two and half years looks like a long time for 4 a topical report.
5 MS. JAMES: Yes, sir. When we received 6 the project initially, we decided that it was such a 7 unique project and dealt very strongly on statistical 8 methods that we chose to bring in and contract Sandia 9 to support the staff.
10 That took longer than anticipated. So a 11 good bit of 2021 was actually procuring the contract 12 from the start. That's not an easy task to go from 13 start to finish. That takes at least six to nine 14 months to get a contract in place.
15 And then when we finally started the 16 review, EPRI and its contractor support provided a lot 17 of information in the topical report. The first set 18 of RAIs we issued was to determine what exactly they 19 wanted us to review and approve because they had 20 provided a lot of examples. And they had provided 21 some code information and some method information.
22 So the first set of RAIs that we issued in 23 2002 after we got the contract in place was first to 24 determine really what we were supposed to review and 25 what we were supposed to document in the SE. And so NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 1 the actual second round of RAIs, phase B, those were 2 kind of the technical questions that we had. So 3 that's why the review took a little longer than 4 anticipated.
5 MEMBER MARCH-LEUBA: Thank you. I want 6 you to understand that, first, anything you hear here 7 until we write the letter are individual members' 8 opinions. We always say that, but we need to keep on 9 saying it.
10 If I summarize what I heard you say, the 11 reason it took two and a half years for review was 12 resource location and the lack of proper 13 communications at the beginning of the project, 14 misunderstanding of the scope.
15 In my opinion, and it's my opinion, I 16 found when I was doing the work on your side of the 17 table, audits helped with the misunderstandings very 18 much. You have an audit with the applicant and you 19 determine the scope real well.
20 I hope management keeps getting lessons 21 learned and we speed up this process because it's in 22 the benefit of everybody. Thank you. Keep going. It 23 was just an observation.
24 MS. JAMES: Understood. The staff has 25 developed internal lessons learned. I try to do that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 1 in every one of my projects so that we can go forward, 2 and do things more efficiently and effectively in the 3 future.
4 MEMBER MARCH-LEUBA: Yes. I'm not wasting 5 time. This is the beginning of the meeting so we 6 still have to -- you always have a pre-submittal 7 meeting with the applicant when you get a flavor of 8 what's going to be submitted. But once the technical 9 responsible staff get assigned to the project, in my 10 opinion, having an audit speeds up everything. It 11 really does. Okay.
12 MS. JAMES: What is data validation and 13 reconciliation? It's a statistical analysis of 14 multiple plant measurements, an aggregate to provide 15 an accurate core thermal power.
16 Of interest to us in this topical report 17 was that DBR can reduce uncertainties associated with 18 the core thermal power and allow plants to operate 19 closer to the approved thermal power without reducing 20 the safety margin.
21 Using the data validation and 22 reconciliation can also help plants reduce single-23 failure vulnerabilities, using more instrumentation to 24 determine core thermal power. It can also improve 25 condition monitoring and condition-based maintenance NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 1 by using the data points collected by the plant 2 equipment. This can be used by the system engineers 3 to better monitor their plants.
4 We would like to acknowledge that DBR has 5 been used in US and European nuclear power industries 6 since 1999. It has been used to assess the target 7 cycle thermal performance, balance of plant feedwater 8 flow metering, and accuracy of the plant metrics.
9 It is also being used, it is my 10 understanding, right now in Europe to increase power 11 output. So that's where the US market is headed.
12 The purpose of the technical report was to 13 describe the process for using the mathematical data 14 validation and reconciliation for specifically 15 monitoring core thermal power. And then use this DBR 16 to proceed into the measurement uncertainty power 17 upgrade or measurement of certain recapture power 18 upgrades.
19 Anything impacting core thermal power has 20 potential to impact safety. So the staff conducted an 21 in-depth review that was discussed in detail with the 22 ACRS subcommittee during the closed portion of the 23 meeting in June.
24 Currently, core thermal nuclear power is 25 calculated based on feedwater flow measurements.
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18 1 There are inaccuracies in the direct measurement of 2 feedwater flow that have resulted in lost generation 3 and potential overpower conditions. In the past 4 ultrasonic flow measurement devices, UFMs, have been 5 used to gain measurement accuracy. But as technology 6 improves, more can be done.
7 Data validation and reconciliation, as I 8 stated, uses statistical analysis of multiple plant 9 measurements to provide an accurate reading of core 10 thermal power. The DBR can then be used to reduce 11 uncertainties in the core thermal power and thus allow 12 licensees to produce more power via a measurement 13 uncertainty power upgrade.
14 So we expect after approval of this 15 topical report that we will begin to see the 16 industry's interest in measurement uncertainty 17 recapture upgrades.
18 We wanted to mention that this review, the 19 safety evaluation was not a review of any software, 20 logic flow, or numerical method implemented by any 21 particular vendor or licensee. It is a review of the 22 concepts of DBR and steps needed to provide the model 23 used by the plants to estimate feedwater flow and core 24 thermal power need and uncertainty.
25 Specific evaluations of software, logic NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 1 flow, and numerical methods would be performed via a 2 license amendment request or an application for a 3 power upgrade.
4 MEMBER MARCH-LEUBA: Lois, this is Jose 5 again. As I told you earlier, this would be a very 6 good point at which you can introduce the risk-7 informed methodology or approach that the staff used 8 to review these power upgrades. Could you give us a 9 hint about how we did it, a summary, a high-level 10 summary?
11 MS. JAMES: Yes.
12 MR. RAHN: If you'd like, Lois -- this is 13 David Rahn -- I could help with that.
14 MS. JAMES: That would be great. Thank 15 you.
16 MR. RAHN: Okay. Thank you.
17 Chairman Rempe and Member March-Leuba, the 18 staff was concerned initially that there was quite a 19 bit of new mathematical and statistics that were used 20 in the concepts that were provided in the DBR 21 methodology.
22 The process of coming up with a reconciled 23 mean and a reconciled variance associated with the 24 measurements from all the different parameters that 25 are going to be used to help reduce the overall NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 1 uncertainty for feedwater flow measurement was 2 complicated.
3 So we wanted to look at it to say, okay, 4 even if we're way off base and we're not understanding 5 it properly, how far off could it possibly be? How 6 bad could we be to come up with a computation of core 7 thermal power?
8 To do that we used our regular risk-9 informed processes, which is to look at what we call 10 the risk triplet. What can go wrong, how likely is 11 it, and what are the consequences? In addition, we 12 looked to see if any of those questions we answered 13 incorrectly, what's the residual risk in that?
14 So essentially, we parsed the uncertainty 15 measurements into what could go wrong if we made a 16 mistake in determining the reconciled mean. And also, 17 what could go wrong if we determined the reconciled 18 variance with an incorrect measurement.
19 By looking at that, we found that there's 20 a history of operating the plant already. And we know 21 what the ballpark should have been for those 22 measurements. So if we're off, we look to see how far 23 off could they really be.
24 In our analysis we determined that even if 25 we were off two percent, which is what was our initial NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 1 look at it, it looked to say that it's not going to be 2 an adverse problem on the operation of the core. We 3 already have an allowance for that two percent.
4 But let's say we were even to double that.
5 We looked to see what the possible implications are of 6 four percent, for example, being off. And we looked 7 to see what conservatisms we already have in either 8 transient analysis or steady-state operations, 9 operating with a starting point of four percent.
10 We found that the conservatisms that we 11 already have built into our analysis are fairly 12 substantial, and that we could live with the risk of 13 as much as four percent uncertainty. So essentially, 14 the risk analysis that was performed was a what-if 15 type scenario. We tried to put bounds on it.
16 The only thing we could not do is 17 determine the likelihood. We had no basis for 18 establishing what would be the likelihood. So we 19 focused on what would be the worst-case consequences.
20 MEMBER ROBERTS: David, this is Tom 21 Roberts. I have a couple of questions on what you 22 just presented.
23 One, you said that the analysis method is 24 complicated. I guess my experience is most analysis 25 methods are complicated. And you end up having to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 1 carefully go through all of the assumptions and make 2 sure that the methods you're using are appropriate for 3 what it is you're modeling.
4 I guess I'm not seeing how this DBR is any 5 different than any other analysis in terms of the 6 residual uncertainty just because you choose to use 7 that method compared to some other method.
8 If you used square-root accommodation of 9 uncertainties for a typical analysis, for example, you 10 would still have to note that those individual terms 11 are all random, independent, and possibly 12 undistributed, all those kinds of things that are 13 important for any kind of analysis.
14 So I'm just trying to figure out why this 15 particular analysis. That's what prompts this thought 16 process.
17 MS. JAMES: David?
18 MR. KREPEL: This is Scott. This is Scott 19 Krepel speaking through a sign language interpreter, 20 if I may. I would like to take an attempt to answer 21 this.
22 Josh Kaizer is one of my staff, but 23 unfortunately he is unable to be with us today. He's 24 on a plane somewhere, I want to say over the Atlantic.
25 He is on his way back from Australia. So I will do my NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 1 best to answer your question.
2 I think one thing that David left out here 3 in the discussion when it comes to risk evaluation is 4 that the risk evaluation itself uses risk-informed 5 scope of review.
6 They did still do the review of the DBR 7 methodology and looked at a lot of the different 8 aspects of the methodology to make sure that they 9 understood appropriately and had an appropriate 10 establishment of certain criteria, so to speak, to 11 make sure that all of the factors that you just 12 mentioned are addressed. The risk evaluation itself 13 is limited in scope in terms of that review as 14 appropriate with the risks.
15 MEMBER ROBERTS: I'm wondering if this 16 thought process would apply to any analysis. It seems 17 like any analysis has the same potential pitfalls that 18 you've identified on this DBR approach.
19 MR. KREPEL: This is Scott again speaking 20 through a sign language interpreter. Sure. You could 21 apply that same approach with any analysis or 22 methodology. In fact, we typically do, but we don't 23 always document it as clearly as Josh Kaizer did. And 24 I think that's a great model for how people could do 25 this type of thing in the future.
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24 1 MEMBER ROBERTS: I think you're probably 2 right. That leads to my next question, which is the 3 conclusion that the four percent, I think you said, 4 overall error is accommodated by the margins that are 5 in other analyses.
6 I look at some of the analyses, like the 7 best-estimate plus uncertainty analysis method, for 8 example, has a 95 percent confidence requirement. I 9 always thought that that was because it was 10 essentially already risk-informed.
11 And 95 percent is good enough because of 12 the very low likelihood of the event itself and the 13 relatively remaining residual amount of defense-in-14 depth that even if you had the event fail, there is 15 still some left. So that's already risk-informed.
16 This now eats into the 95 percent. And 17 I'm wondering if you've thought about that. And the 18 four percent, how much does that eat into 95 percent 19 on an accident or a DNBR 95/95 criteria, something 20 like that? Is this essentially double-spending some 21 of the risk-informed judgement is where my question is 22 going.
23 MR. KREPEL: This is Scott again. I think 24 that's a good comment and a good point. I would say 25 that Josh did look at the consequences in his safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 1 evaluation. And I believe you might have already 2 taken a look at that safety evaluation.
3 He discusses various things that might 4 happen if you were to go to 40 percent, a higher power 5 level than estimated, like increased oxidization, for 6 example, for the fuel during normal operation. It's 7 possible there's some loss of margin to the operating 8 limits, but Josh also pointed out that typically you 9 would have margin in the design limit. That is a 10 worst-case scenario.
11 The staff typically, intuitively at least, 12 thinks that is much smaller than typical for that 13 error. Or making a mistake at 40 percent, for 14 example. So the risk would be considered acceptable.
15 The licensees typically do have some margin there in 16 their design limit.
17 MEMBER ROBERTS: Maybe not a fair question 18 at this point, but how high would it be before you 19 would start to worry? Would seven percent be a 20 problem? Would ten percent be a problem? Would 20 21 percent be a problem? Where does the thought process 22 start to break down?
23 MR. KREPEL: No, that's not a fair 24 question. To be honest, there is some engineering 25 judgement there that's involved in that determination.
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26 1 And so I can't really answer that question myself 2 alone on the spot.
3 MS. JAMES: This Lois James. I would say 4 one of the things that made this unique was we're 5 essentially approving a concept. We know that there 6 are different methods being used at different plants.
7 We know that vendors have different 8 methods for calculating and they have different 9 equipment. So we know we're going to have to go into 10 more detail of those, of some of that when we do 11 individual reviews that get submitted.
12 MEMBER ROBERTS: Thank you. I think I 13 understood that. So if I summarize, what I think I'm 14 hearing is, one, you are treating this as if this is 15 a new analysis method that's expected to be done 16 properly. And I believe your safety evaluation has 17 some conditions in it that help you ensure that the 18 applicant is applying it properly.
19 It's mathematically rigorous. I think 20 your SE would agree with that, that it's a valid 21 approach to use. And you don't expect it to introduce 22 any significant error uncertainty at all. But you 23 also do the side study to say, okay, what if we're 24 wrong?
25 MS. JAMES: Yes.
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27 1 MEMBER ROBERTS: And based on judgement at 2 what's probably four percent of a very, very high 3 estimate of how bad it could be, you conclude it's 4 probably okay. So I think I understand. I appreciate 5 it. Thank you.
6 MEMBER MARCH-LEUBA: Any more questions 7 from members or comments from the staff?
8 MS. JAMES: I should apologize. David 9 Rahn is having a little trouble with his Teams. It 10 booted him out when he was speaking. He apologizes.
11 He's unable to get back in at the moment.
12 MEMBER MARCH-LEUBA: We have a good sound 13 on our end. Okay. Lois, you can continue with the 14 conclusions.
15 MS. JAMES: Based on the staff's review, 16 we looked at the risk assessment of the DBR results.
17 We looked at previous treatment of similar models and 18 simulations.
19 We looked at previous evaluations of 20 nuclear power plant processes measurement uncertainty.
21 We looked at the understanding of the DBR method and 22 previous treatment of calculations of the feedwater 23 flow and its uncertainty.
24 And based on all of this, we concluded 25 that there is reasonable assurance that the DBR method NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 1 as described in the topical report can be used to 2 determine the core thermal power and the core thermal 3 power uncertainty. Thus, all DBR conditions and 4 limitations have been satisfied.
5 So that's all we had anticipated 6 discussing this morning. Are there any other 7 questions or comments? And I'm going to see if I can 8 get David --
9 VICE CHAIR KIRCHNER: Lois, this is Walt 10 Kirchner. Tom had asked already what I was going to 11 ask at the conclusion.
12 What would you be expecting to see if an 13 applicant comes in with an LAR to upgrade the power?
14 More precision or improved instrumentation on feed 15 order to reduce the uncertainty in that particular 16 parameter or do you think they'll just say, well, the 17 staff looked at four percent.
18 They've kind of shown their hand. And now 19 we'll look at testing the staff to see if we can get 20 a -- I'll pick a number -- three percent upgrade in 21 power, How will you use this when in a practical 22 sense an applicant comes in with a power upgrade 23 application?
24 MS. JAMES: Well, the --
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29 1 for? You won't be looking for it on the reduced 2 margin.
3 MS. JAMES: No.
4 VICE CHAIR KIRCHNER: And things like 5 DNBR. So how in practice do you anticipate this being 6 used with an application?
7 MS. JAMES: We have already rolled out 8 this to MURs. And we understand that that's 9 associated with the LEFMs. That's kind of the 10 starting point. So they're going to use this in 11 conjunction with that guidance on how to do -- well, 12 I guess it's a reg guide on how to do the MURs.
13 And then we would expect them to come in 14 with their calculation. How are they going to 15 calculate the uncertainty? What's their computer 16 program? What's their modeling?
17 Since David is not in, I don't know if any 18 EPRI person would kind of like to step in and make the 19 comment at this point of anything further. We expect 20 a lot of the computational models, the uncertainties, 21 how it's going to be used, how it's not going to be 22 used.
23 VICE CHAIR KIRCHNER: Is there a need, 24 Lois, in your opinion for further guidance from the 25 staff for implementation?
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30 1 MS. JAMES: We have not determined whether 2 we need additional guidance, but that's definitely not 3 out of the realm because we've done it, as I 4 mentioned, for the LEFMs. Our focus has been to get 5 the concept approved and out the door next month.
6 MR. KREPEL: This is Scott, if I could 7 jump in there as well, again speaking through a sign 8 language interpreter here. I just want to give a 9 reminder.
10 In the safety evaluation there are ten, if 11 I recall correctly, ten criteria. Those criteria will 12 be looked at in more detail during the actual review 13 because those are viewed as fundamental to whether or 14 not the methodology is acceptable.
15 MEMBER MARCH-LEUBA: I believe EPRI would 16 like to make a comment.
17 MR. CRYTZER: Hello.
18 MEMBER MARCH-LEUBA: Speak loudly.
19 MR. CRYTZER: Okay. This is Kurt Crytzer 20 with EPRI. The way that we had instructed it was to 21 capture the BDI 2048 method, which is a statistical 22 method, and not put it into any particular software.
23 The software we used would have to be compliant with 24 what BDI 2048 would accept.
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31 1 very much similar to the leading edge flow meters 2 where a correction factor would be applied to the 3 feedwater flow to recapture some of the uncertainty 4 measurement.
5 MEMBER MARCH-LEUBA: Since we have you 6 there, for the record, I believe the correction we're 7 talking about is 15 megawatts electric over 1,000. So 8 it's like 0.1 percent. Would the staff take the risk 9 of operation to say, we were completely wrong but we 10 would expect the correction to be 0.1, correct?
11 MR. CRYTZER: Yes.
12 PARTICIPANT: Can you say your name again 13 clearly for the court reporter?
14 MR. CRYTZER: Kurt Crytzer.
15 PARTICIPANT: That's good for context.
16 MEMBER MARCH-LEUBA: Any more comments or 17 questions from members or the staff?
18 Hearing none, I'm going to open it up in 19 case there's a member of the public that wants to 20 place a comment on the record. Please do so now.
21 Hearing none, I return the meeting to you, 22 Ms. Chair.
23 CHAIR REMPE: Thank you, Jose.
24 At this time, I would note to the court 25 reporter we're going to go off the record. We'd like NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 1 you to come back at 1:00 p.m. East Coast time.
2 (Whereupon, the above-entitled matter went 3 off the record at 9:06 a.m. and resumed at 1:00 p.m.)
4 CHAIR REMPE: Okay. It's 1:00 p.m. on the 5 East Coast. I'd like to ask Member Ballinger to take 6 us through our second topic for this meeting.
7 Ron?
8 MEMBER BALLINGER: Thank you, Madam 9 Chairman. We had a meeting on this topic for the 10 Vogtle LTA at our subcommittee meeting in June where 11 Southern Nuclear and Westinghouse presented an 12 exhaustive, I would say, very thorough presentation of 13 what they claim to do and an analysis that was 14 required.
15 The staff presented their analysis. Their 16 presentations today are a subset of those 17 presentations. And I think the staff would like to 18 say something initially.
19 MR. MARKLEY: Yes. This is Mike Markley.
20 I'm Chief of Licensing for the Division of Operating 21 Reactor Licensing for the Vogtle site, Units 1 and 2.
22 I'd like to thank you all for helping and 23 reviewing this first-of-a-kind -- it is very much a 24 first-of-a-kind review for us in the fact that it's 25 rated in five percent enriched uranium-235 of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 1 accident-tolerant fuel assemblies. And it's 2 applicable only to the current burn-up limits the 3 staff has approved for Vogtle Units 1 and 2.
4 You're going to hear the highlights today 5 regarding our safety evaluation. The staff expects 6 future submittals from the industry requesting higher 7 enrichments and higher burn-ups for longer operating 8 cycles.
9 We have been having pre-licensing of the 10 other licensees and expecting other licensees to 11 submit a license amendment request to transition to 12 24-month cycles in the next few months. We appreciate 13 the subcommittee's questions and comments during its 14 June 21st meeting.
15 For today's briefing, the subcommittee 16 requests that the presentation discusses why NRC has 17 confidence in technical issues to resolve properly, 18 balancing engineering judgement and risk-informed, and 19 look at the long game of batch loading. We are 20 prepared to do that today.
21 The staff welcomes an ACRS letter. We 22 thank you for the opportunity to present and talk with 23 you today. We'll do our best to answer questions.
24 Thank you. I'll turn it over to Southern 25 Company staff.
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34 1 MEMBER BALLINGER: I think Southern 2 Company's on.
3 MR. MARKLEY: Yes.
4 MR. JOYCE: Thank you. This is Ryan 5 Joyce. I'm the Licensing Manager at Southern Nuclear.
6 First of all, I'd like to thank the ACRS 7 for their consideration of this important initiative 8 that benefits not just SNC but the entire industry as 9 we move to higher enriched fuels. It ultimately will 10 help ensure the safety, reliability, and economics of 11 nuclear power plants, ensuring nuclear power is a 12 viable energy source for many years into the future.
13 The agenda items I'll be discussing will 14 be the LTA program review, request exemptions, summary 15 of testing for adopting AXIOM cladding, and the 16 various analyses that were performed that ultimately 17 will demonstrate the due diligence performed to ensure 18 the LTAs will operate safely and within the analyzed 19 limits.
20 The goal of the program is to irradiate 21 higher enriched fuels in a commercial reactor and 22 generate data in support of future license 23 applications. Although it's a limited number of fuel 24 rods, this will allow us to exercise the regulatory 25 process and work through various issues associated NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 1 with ultimately loading higher enriched fuel 2 assemblies beyond five weight percent, and work 3 through logistical, regulatory, and legal challenges 4 that will come up as we go to this higher enriched 5 fuel.
6 A second objective is to obtain additional 7 data for accident-tolerant and advanced fuel 8 materials, fuel pellet and cladding materials.
9 A future goal of this, which is outside 10 the scope of this amendment, is to ultimately increase 11 the license burn-up limit and go to higher burn-up 12 fuels. Again, that's outside of this specific 13 amendment but that is the end in mind, to ultimately 14 allow the application of higher enriched, higher burn-15 up fuel.
16 For the LTA program, we have four LTAs.
17 Each LTA will contain four fuel rods with up to six 18 weight percent U-235. All LTA rods will have AXIOM 19 cladding. All but one LTA will have chromium coating.
20 About half the rods, 136 for LTA including higher 21 enriched rods, will have doped ADOPT pellets whereas 22 the other half, roughly 128, will be IFBA rods.
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36 1 will have this higher enriched fuel.
2 The license amendment request requested to 3 change Tech Specs 4.2.1, 3.7.18, and 4.3 regarding 4 these LTAs. Tech Spec 4.2.1 already allows test 5 assemblies that have not completed representative 6 testing to be in non-limiting core regions. In a few 7 slides from now, I'll discuss why we do not believe 8 this non-limiting requirement was met and why we felt 9 the need to explicitly revise Tech Spec 4.2.1.
10 Tech Spec 3.7.18 and 4.3 require an 11 assembly enrichment of less than five percent. Due to 12 the small number of fuel rods above five percent, so 13 only four out of 264 per assembly, this requirement is 14 still met. In other words, the average assembly 15 enrichment is less than five percent.
16 However, we felt it prudent to remove any 17 kind of regulatory uncertainty associated with whether 18 or not we were meeting this requirement. And as I 19 previously mentioned, we wanted to make sure we 20 exercised the regulatory process to lay a road map and 21 ultimately load fuel assemblies greater than five 22 percent enrichment.
23 As will be discussed in a few slides, as 24 part of this amendment we revised our facility 25 operating license to remove an exemption we had to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 1 70.24 as part of our change in licensing basis, which 2 will be discussed in a few slides.
3 As part of this, we requested an exemption 4 to 50.46 and Appendix K for AXIOM cladding. The 5 exemption request will apply to the application of 10 6 CFR 50.46 and 10 CFR 50 Appendix K of regulations to 7 the LTA design, utilizing AXIOM cladding in Vogtle 8 Unit 2.
9 In conjunction with the 17 percent maximum 10 local oxidation acceptance criteria prescribed by 11 50.46, a more restrictive criteria was assessed 12 consistent with the data presented in the AXIOM 13 topical reports.
14 For regulatory clarity, SNC decided to 15 adopt a newer 50.68 regulation to replace the older 16 70.24 regulation that was described in our facility 17 operating license. We felt that adopting 50.68 18 provided a clean regulatory foundation for moving 19 forward with high-enriched fuel assemblies.
20 Moving to 50.68 necessitated a requested 21 exemption, 50.68(b)(7), to allow the LTAs to have five 22 percent enriched fuel rods. Similar to our tech spec 23 requirements, 50.68 refers to an assembly enrichment 24 of less than five percent.
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38 1 meet this on an assembly average, a weight average 2 level, but we felt it prudent to once again request an 3 exemption in lieu of any regulatory uncertainty based 4 on the limited number of higher enriched rods.
5 The technical justification, the intent of 6 50.68 is to include inadvertent criticality. We 7 demonstrate through our very thorough analyses and 8 very conservative analyses that inadvertent 9 criticality well be precluded based on the 10 restrictions we have in place.
11 The NRC approved for the LTAs to include 12 their own criticality analysis. Adherence to these 13 analyses fulfill the 50.68 requirements. The 14 placement is the new storage racks, which is 15 administratively controlled.
16 AXIOM cladding, ADOPT pellets, and 17 chromium-coated Optimized ZIRLO cladding have all been 18 used in US PWR reactors, similar to Vogtle. In 19 addition, the AXIOM cladding in the topical reports 20 has been interviewed by the NRC and the ARCs.
21 The only novel feature without US 22 operating experience associated with these four LTAs 23 is a very limited number of fuel rods with greater 24 than five-percent enrichment. And again, this 25 represents a very small percentage of the overall NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 1 reactor core.
2 There's about 51,000 rods in the reactor 3 core, so 16 rods out of the 51,000. So a very small 4 percentage of the overall reactor core.
5 MEMBER MARCH-LEUBA: And remind me again 6 from the subcommittee, some of these rods are in 7 limiting positions. Or they're in positions that 8 we're not limiting until -- can you explain that?
9 MR. JOYCE: If I understand the question 10 about limiting, some of these will have a highest, and 11 as mentioned in a couple of slides, will be in --
12 MEMBER MARCH-LEUBA: I'll wait. Yes.
13 MR. JOYCE: Any other questions?
14 VICE CHAIR KIRCHNER: For the record, 15 could you quickly review how many rods with those 16 claddings and composition were tested?
17 MR. JOYCE: For Millstone, unit three.
18 The licensee stated there will be up to eight re-test 19 assemblies containing fuel rods fabricated with AXIOM 20 cladding inserted into the core for Millstone's team.
21 That was from May 2017.
22 Byron Unit 2 from an April 2019 amendment 23 requested to insert two LTAs designed by Westinghouse.
24 The LTAs were based on a vintage optimized fuel 25 assembly design. The licensee proposed to insert up NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
40 1 to 20 test rods of three different types between the 2 two LTAs. The rods contained a mixture of three 3 materials: uranium, fuel pellets ADOPT, and coated 4 optimizer of a low cladding.
5 VICE CHAIR KIRCHNER: So there's a good 6 database there from that LTA set of experiments?
7 MR. JOYCE: I would say we believe there 8 is a good database that supports those ADOPT, AXIOM, 9 and Permian cladding. I don't know if Westinghouse 10 has anything to add to that.
11 MR. SMITH: Ryan, this is Jim. We agree.
12 The basis for what we were doing is the other 13 applications.
14 MR. JOYCE: So to your question earlier, 15 the LTAs will have the highest linear heat generation 16 rate or local peaking for portions of the cycle, both 17 at steady state and transient conditions.
18 Nonetheless, the technical specification and limits 19 will continue to be met.
20 There are no additions to the reference 21 Tech Spec 5.6.5 needed for these LTAs as the current 22 methods are used to evaluate them.
23 MEMBER MARCH-LEUBA: Will the five percent 24 enrichment rods lean? What's the power inside the 25 bundle for those five-percent enrichment positions?
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41 1 MR. JOYCE: Radu, can you understand that 2 question?
3 MR. POMIRLEANU: Could you repeat it?
4 Could you repeat the question, please?
5 MEMBER MARCH-LEUBA: There are four bins 6 in the fuel element, an enrichment at about five 7 percent. What is the relative bin power for those 8 bins? Is it much higher than average, about average, 9 or lower than average?
10 MR. POMIRLEANU: It is not significantly 11 different from the leading bins in any given cycle.
12 MEMBER MARCH-LEUBA: Okay. That's good 13 enough. It's about the same. Thank you.
14 MR. POMIRLEANU: Yes. Thank you.
15 MR. JOYCE: You'll see in the slides.
16 It'll be within the FDA bin factor limits that are 17 already in the core report.
18 MR. POMIRLEANU: It's in the core report.
19 This Radu Pomirleanu from Westinghouse.
20 MR. JOYCE: For LOCA, the existing large-21 brick LOCA and small-brick LOCA analysis of record for 22 Vogtle are representative of the LTAs. The fuel is 23 negligibly impacted by the presence of the LTAs. The 24 50.46 acceptance criteria continues to be met.
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42 1 analysis that depends only on parameters are not 2 impacted by the LTAs as these don't impact core 3 average heat transfer characteristics, decay heat, or 4 initial core energy. The events that depend on local 5 effects were evaluated for potential effects to the 6 LTAs.
7 It was determined that there were no 8 impacts to the codes or methods. Any impact due to 9 LTA features is offset by system margins. While the 10 LTAs may lead the core in factors, they are placed in 11 non-limiting locations with respect to analysis.
12 Ultimately, fuel-specific criteria applicable to each 13 accident continues to be met. There is no impact to 14 the source or consequences. The announcement remains 15 bounding.
16 For fuel rod performance, ADOPT and AXIOM 17 were explicitly modeled to PAD5. Premium coated 18 benefits for corrosion were conservatively not 19 included. For fuel rod design, there is no impact to 20 existing DNBR margin.
21 For core physics, the chromium coating and 22 ADOPT fuel pellets were explicitly modeled. There is 23 no impact to neutronics modeling for the fuel rods 24 about five weight percent. Core monitoring with 25 Beacon is ineffective.
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43 1 VICE CHAIR KIRCHNER: Can you put that in 2 context of looking ahead to what you want to do next, 3 where you go to higher enrichment? Higher average 4 fuel assembly will have a lot higher enrichment if you 5 put more rods in, obviously, that are six percent.
6 What you're saying here is that for these 7 analyses of record, fuel assembly isn't really -- the 8 margins that you have in a conservative analysis of 9 record are greater than the impact that you can 10 calculate as a result of putting just four rods in?
11 MR. JOYCE: You're speaking regarding the 12 core physics?
13 VICE CHAIR KIRCHNER: Physics and thermal 14 hydraulics.
15 MR. JOYCE: Radu, do you want to take the 16 lead on addressing that question?
17 MR. POMIRLEANU: Sure. This is Radu 18 Pomirleanu from Westinghouse. First of all, I'd like 19 to point out that there will be other licensing 20 submittals that will address publications with an 21 increasing number of higher enrichment fuel and higher 22 burn-ups. But yes, we expect a wider range of impacts 23 that will have to be addressed separately beyond the 24 scope of this submittal.
25 VICE CHAIR KIRCHNER: What I was trying to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 1 infer simply on your behalf is that was only four rods 2 in a bundle. The impact that it would calculate on 3 your core physics parameters and your thermal-4 hydraulics is really minimal. When you take the next 5 step, there will obviously be an impact on all these 6 evaluations.
7 MR. POMIRLEANU: Yes. That is correct.
8 MR. JOYCE: Any additional comments or 9 questions on slide 10?
10 For the criticality analysis, these 11 assemblies are only requested to be used in the unit 12 to prevent fuel pool. Unit 1 spent fuel pool storage 13 is prohibited.
14 The law addresses increased enrichment, 15 use of ADOPT pellets, and use of chromium-coated AXIOM 16 cladding with regard to storage criticality. Current 17 NRC-approved codes were applied to address LTA 18 storage.
19 For LTA storage not requiring the new fuel 20 storage ranks in Unit 2 2-out-of-4 checkerboard fuel 21 pool storage, a direct analysis was performed. The 22 new fuel storage rack analysis demonstrates a margin 23 of limits including dry, fully flooded, maximum, 24 moderation considerations. The Unit 2 storage racks 25 in Unit 2 2-out-of-4 checkerboard are confirmed for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
45 1 the reload process.
2 For spent fuel pool storage, the current 3 burn-up limit for maximum enriched fuel assemblies is 4 about four gigawatt-days per MTU. That's the current 5 limit for about a five percent enriched fuel assembly.
6 To provide significant conservativeness 7 into this limit, a burn-up limit of 64 gigawatt-days 8 per MTU was selected for the LTA storage. So a margin 9 increase of 24 gigawatt-days per MTU. This is greater 10 than the eight percent margin.
11 The 64 gigawatt-days per MTU was selected 12 by additional storage options. Should the LTAs be 13 approved for operation, the burn-up rate is 64 14 gigawatt-days per MTU. In other words, 64 is beyond 15 our burn-up limits. We cannot go to 64 without a LAR.
16 But if we do ultimately go to a higher burn-up with 17 these four LTAs in the third cycle, we can store with 18 the other storage option.
19 MEMBER BALLINGER: This is Ron Ballinger.
20 You used the word if, but it's more like when.
21 MR. JOYCE. Yes, when/if.
22 MEMBER BALLINGER: Okay.
23 MEMBER HALNON: This is Greg. The Unit 1 24 spent fuel, is that prohibited because of lack of 25 analysis or a failed analysis?
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46 1 MR. JOYCE: I'll let Mike Werner answer 2 this.
3 MR. WERNER: Mike Werner from 4 Westinghouse. Essentially, the two pools have 5 different racks so we only analyze those in the Unit 6 2.
7 MEMBER HALNON: So it's lack of analysis?
8 MR. WERNER: Correct.
9 MEMBER HALNON: Talk to me about 10 physically. Can you physically get a Unit 2 assembly 11 into the Unit 1 spent fuel pool?
12 PARTICIPANT: This is Matt. We're doing 13 this for Unit 2 because that analysis is specific to 14 Unit 2.
15 MEMBER HALNON: There's a gate between 16 them?
17 PARTICIPANT: That's correct.
18 CHAIR REMPE: Excuse me. Someone on the 19 line has their microphone open and it's hard for us in 20 the room to hear what the conversation is. So please 21 check your mics on your computers.
22 Please re-answer again, Matt. Sorry to 23 interrupt.
24 PARTICIPANT: No problem.
25 Yes, that's correct. It's a shared pool.
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47 1 They do have a physical gate between the two pools.
2 However, they can transfer fuels between them if 3 needed. As I stated before, this was more for an 4 analysis convenience sake to keep it specific to the 5 Unit 2 pool and not do the two.
6 MEMBER HALNON: Is it a routine or has 7 Unit 2 fuel gone to Unit 1?
8 PARTICIPANT: They have. There are cases 9 in a redesign scenario where we can pull from Unit 2 10 fuel to load Unit 1, but that's on a case-by-case 11 basis.
12 MEMBER HALNON: How do you control that?
13 Is that a typical type of --
14 PARTICIPANT: It is. It's controlled via 15 the procedures and processes.
16 MEMBER HALNON: The fuel accounting 17 process?
18 PARTICIPANT: That's correct.
19 MEMBER HALNON: And that's done by the 20 fuel engineering?
21 PARTICIPANT: That's correct.
22 MEMBER HALNON: Okay.
23 MR. JOYCE: When we analyze the old Yankee 24 fuel storage racks, the analyses are very different 25 with regard to the filters in the two fuel racks.
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48 1 MEMBER HALNON: Okay. You get rid of all 2 the borax and all that stuff, there's nothing that 3 you've got to worry about degradation in the Unit 2 4 fuel pool racks?
5 MR. WERNER: Mike Werner again. We're not 6 crediting any of the older filters.
7 MEMBER HALNON: Okay. Crediting any boron 8 in the pool as well?
9 MR. WERNER: Well, the boron is still the 10 boron credit.
11 MEMBER HALNON: But you aren't crediting 12 that?
13 MR. WERNER: Right.
14 MEMBER HALNON: Okay. Thanks.
15 MR. WERNER: The analysis that's currently 16 in place credits that. We've just built upon it.
17 We're not changing it.
18 MEMBER HALNON: Okay. They're credited 19 the same amount so one can't dilute the other one when 20 you open the gate?
21 MR. WERNER: I'm not sure what the --
22 MEMBER HALNON: I'd be interested in 23 understanding what physically can go wrong that can 24 cause the boron concentration to change. Typically, 25 if it was my plan, I would credit the same amount of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
49 1 boron concentration in both pools so one can't dilute 2 the other.
3 MR. WERNER: I believe that the spec limit 4 is the same for both. It's 2,300, I believe.
5 MEMBER HALNON: Okay. That would be my 6 guess but I didn't want to guess.
7 MR. JOYCE: We can confirm that, but 8 what's listed is Unit 1 and Unit 2 -- we have combined 9 tech specs with some different requirements for the 10 Unit 1 and Unit 2 spent fuel pools. Where that 11 requirement is listed is common to both units.
12 MEMBER HALNON: Since I'm elaborating, 13 I'll ask the question, will the operator see anything?
14 Will they see anything during operation or is it 15 thermocouples, SPMDs, or anything in the core?
16 PARTICIPANT: I don't think we -- Radu, 17 I'll defer to you, but my personal answer to this is 18 they shouldn't see anything leave with four bins per 19 assembly. Even if they're in an instrumented 20 location, there could be some minor differences in 21 mapping when we take those in rack engineering.
22 But in terms of operation, I think what 23 we're expecting should be negligible and probably 24 would be within the amount which we typically see in 25 operation.
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50 1 MEMBER HALNON: Okay. Thanks.
2 PARTICIPANT: Radu, do you have anything 3 additional to add to that?
4 MR. POMIRLEANU: No.
5 PARTICIPANT: Thank you.
6 MR. JOYCE: That concludes my 7 presentation. Are there any additional questions or 8 discussions?
9 MEMBER BALLINGER: Okay. I don't know if 10 we have our consultants on the line. I don't know 11 who's online. If not, then thank you very much for 12 your presentation.
13 Now we should transfer to the staff.
14 CHAIR REMPE: I don't see any consultants 15 on the line.
16 MEMBER BALLINGER: Yes. I thought Steve 17 would be online. He was this morning.
18 MR. SCHULTZ: No questions or comments, 19 Ron.
20 MR. BLEY: Dennis is here too. No 21 questions either.
22 CHAIR REMPE: I didn't see you on the 23 list. Sorry.
24 MEMBER BALLINGER: Okay. Thanks again.
25 We just need to change out. The seat is still warm.
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51 1 There we go.
2 MR. LAMB: My name is John Lamb. I work 3 NRC's Division of Operating Reactor Licensing at the 4 Office of Nuclear Reactor Regulation, NRR. I'll 5 describe the licensing actions that SNC has requested.
6 Charley Peabody, who is virtual and works 7 for the Nuclear Systems Performance Branch, SFSB, in 8 the Division of Safety Systems, DSS and NRR, will 9 discuss the updated final safety analysis report, the 10 FSAR, Chapter 15, accident analyses, the loss of LOCA 11 accident analyses and the non-LOCA accident analyses.
12 Brandon Wise from SFNB, DSS, and NRR will 13 discuss the code analysis and fuel rod design. Kent 14 Wood from SFNB, DSS, and NRR will discuss the fuel 15 handling and storage. Mike Markley, who you heard 16 through the introductions earlier, is the Branch Chief 17 from Doral and NRR and will provide the conclusion.
18 SNC has requested four licensing actions.
19 One is a license amendment request and three are 20 exemptions. The license amendment request is to 21 revise the license condition 2D and four technical 22 specifications.
23 The proposed change in license condition 24 2D is to delete a 1986 exemption to Title 10 of the 25 Code of Federal Regulations, 10 CFR Section 72.4, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 1 which is criticality analysis requirement and 2 criticality accident requirements, as SNC is 3 voluntarily adopting 10 CFR 50.68.
4 ADOPT, AXIOM, Prime ZIRLO, and Optimized 5 ZIRLO are trademarks or registered trademarks of 6 Westinghouse Electric Company.
7 The three tech specs that SNC is proposing 8 is to revise, one, Tech Spec 3.7.18, which is fuel 9 assembly storage in the fuel storage pool. The second 10 one is Tech Spec 4.2.1, fuel assemblies. And the 11 third one is Tech Spec 4.3, fuel storage.
12 Technical Specification 3.7.18 refers to 13 fuel storage in a fuel storage pool. Therefore, the 14 tech spec note is added to the fuel storage for the 15 accident-tolerant fuel, ATF, lead test assemblies, 16 LTAs, to meet the Tech Spec 4.3.
17 Tech Spec 4.2.1 allows ZIRLO, Zircaloy, 18 and Optimized ZIRLO only. Therefore, the tech spec 19 change is needed for the insulation of the ATF LTAs.
20 In addition, Tech Spec 4.2.1 states a 21 limited number of lead test assemblies that have not 22 completed representative testing may be placed in non-23 limiting core regions. Therefore, the tech spec 24 change is needed to allow --
25 MR. PEABODY: Excuse me, John. Can you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
53 1 advance the slides so they can see what you're talking 2 about? Thank you.
3 MR. LAMB: There are a limited number of 4 lead test assemblies that have not been completed.
5 Representative testing may be placed in non-limiting 6 regions. Therefore, a tech spec change is needed to 7 allow SNC to place the four ATF LTAs in limiting core 8 locations in Vogtle Unit 2.
9 Technical Specification 4.3 allows up to 10 five weight percent uranium-235. Therefore, the tech 11 spec is needed for storage of the ATF LTAs.
12 The first exemption is 10 CFR 50.46. The 13 second one is 10 CFR 50 Appendix K. Those proposed 14 exemptions are needed to allow the use of AXIOM 15 cladding. The third exemption is 10 CFR 50.68(b)(7),
16 which allows a greater-than-five weight percent 17 uranium-235.
18 I'm going to turn it over to Charley 19 Peabody to discuss the FSAR Chapter 15, accident 20 analyses.
21 Charley, are you there?
22 MR. PEABODY: Yes. Thanks, John. Can you 23 advance to the next slide? All right.
24 So the accident analyses, most of the 25 accidents were addressed with one of the points that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 1 was mentioned in the previous discussion. The overall 2 enrichment only increases negligibly by adding the 3 four LTAs when you are addressing core-wide effects.
4 That didn't really change any of the dose 5 consequences for accidents that are evaluated from a 6 core-wide standpoint. There were a few accidents that 7 were evaluated that do have local effects, locked 8 rotor and RCPs. Actually, there were a couple of 9 others.
10 The only one that ended up still being 11 limiting was the rod injection analysis. So the tech 12 specs are going to reflect that the LTA core locations 13 will remain appropriately limiting for rod injection 14 accidents for LTA utilization.
15 That's all on this. Are there any 16 questions on the accident analysis?
17 VICE CHAIR KIRCHNER: Charley, this is 18 Walt Kirchner. When you say limiting, the LTAs cannot 19 go into a control rod position?
20 MR. PEABODY: That's correct.
21 VICE CHAIR KIRCHNER: Is that what you're 22 saying in plain English?
23 MR. PEABODY: Yes.
24 VICE CHAIR KIRCHNER: Okay, just for the 25 record.
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55 1 MR. WISE: This is Brandon Wise for the 2 staff. I just want to correct that they can go in rod 3 locations but not limiting rod locations.
4 VICE CHAIR KIRCHNER: Okay. That is 5 determined by previous analysis of record? How do you 6 determine a priori which are limiting and which 7 aren't?
8 MR. WISE: I believe SNC and Westinghouse 9 have determined which locations are limiting with 10 respect to control rod location. I believe they tend 11 to be the same locations for most reloads, so they 12 won't be in those locations. And I'm sure there will 13 be some confirmatory analysis to confirm that.
14 MR. PEABODY: Yes. It also would depend 15 on the individual rod worth. I know rods towards the 16 center of the core have more rod worth than control 17 rod locations on the periphery.
18 MR. WISE: All right. I am Brandon Wise 19 with the NRC's Nuclear Methods and Fuel Analysis 20 Branch. I did the review for the code analysis and 21 fuel rod design.
22 For the most part, the codes used by SNC 23 and Westinghouse for the analysis of the LTAs are 24 mostly applicable to the LTAs with the exception of 25 some enrichment limits. Given the limited number of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 1 higher enriched rods and the amount of enrichment that 2 they have, the overall enrichment of the assembly 3 still remains below five percent, and in a way still 4 remains in the range of applicability for codes such 5 as PARAGON and NEXUS.
6 Overall, there's very little change in the 7 neutronic performance resulting from the higher 8 enriched rods. I'd expect more of a change from the 9 ADOPT fuel pellets, which are being justified by not 10 being in limiting positions with respect to control 11 rod injection due to the potential for more severe 12 control rod injection accidents due to the increased 13 density of the ADOPT fuel pellets.
14 As far as thermal-hydraulic codes go, 15 there's basically no impact as a result of enrichment 16 or ADOPT fuel pellets. There are chromium-coated 17 rods. Although one rod is chromium-coated, we would 18 expect there to be safety enhancement as a result of 19 the chromium coating, but they're not crediting that 20 enhancement.
21 We did examine some potential detrimental 22 effects of the chromium coating and determined that 23 there's no significant loss of margin associated with 24 any potential detrimental effects that were outlined 25 in the ATF-ISG.
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57 1 So overall, we expect the AXIOM cladding 2 to perform as well as the Optimized ZIRLO cladding, 3 which is what's in the co-resident fuel. And there's 4 no credit for the chromium-coated cladding.
5 Next slide, please.
6 I'll go into a little bit more detail on 7 each of the four LTA characteristics, the first being 8 the AXIOM cladding. It's a zirconium alloy cladding 9 that is expected to demonstrate better in reactor 10 performance compared to Optimized ZIRLO.
11 For the sake of this application, we 12 assumed it performs as well as Optimized ZIRLO.
13 Therefore, it's on par with the rest of the co-14 resident fuel.
15 The chromium coating is a thin chromium 16 coating. It has corrosion resistance, enhanced 17 corrosion resistance compared to Zircaloy cladding.
18 There's no impact to the thermal-hydraulic analysis.
19 The coating is extremely thin and thermal-20 hydraulic analysis doesn't have enough resolution to 21 even capture the reduction in flow area that would 22 result from the chromium coating. And of course, 23 there's no benefit taken for the chromium coating.
24 For each assembly there's four enriched 25 rods enriched to six percent. There's minimal impact NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 1 to neutronic performance as a result because the 2 overall assembly enrichment is going up by a 3 negligible amount.
4 There's actually a much higher increase in 5 fuel content associated with the ADOPT fuel pellets, 6 which are a slightly higher density. And they are 7 coated with chromium and alumina. Several 8 characteristics are different from standard UO2 9 pellets. That enhances the performance of those ADOPT 10 pellets.
11 One of the concerns the staff had with the 12 ADOPT fuel pellets was the increased fuel content and 13 the potential for more severe reactivity in accidents 14 such as control rod injection. This was dispositioned 15 by limiting the locations in which the LTAs can be 16 stored in the core to non-limiting locations with 17 respect to control rod injection.
18 Next slide, please.
19 Any questions for fuel rod design or 20 coating analysis? Okay. I'll hand the presentation 21 over to Kent Wood, who will discuss fuel handling and 22 storage.
23 MR. WOOD: Good afternoon. My name is 24 Kent Wood. I'm here to do fuel handling and storage.
25 We've done this several times. You're getting four NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
59 1 rods in higher enrichment to six percent, which adds 2 a very small amount of material. The theoretical 3 increase in density actually has much more visible 4 material to the fuel assemblies.
5 For the analysis, they compared the 6 analysis. They used applied engineering judgement to 7 compare this analysis to the analysis of record and 8 then added copious amounts of margin.
9 In particular, for the new fuel storage 10 and for the spent fuel storage for the 2-out-4, which 11 is the fresh fuel assembly storage in the spent fuel 12 pool, they credited IFBA, which is not in the analysis 13 of record for the new fuel storage or spent fuel 14 storage. So that provides a lot of margin.
15 As Southern said, they really can't get to 16 -- they also did an analysis where they can credit 17 burn-up, but they can't get that burn-up they're 18 crediting of 64. It will add another LAR, but I guess 19 they won't have to talk to me that time.
20 So we looked at that. That provides 21 margin. We looked at their analysis of record. They 22 have a copious amount of margins for that 64 gigawatt-23 days for the burned fuel.
24 For the LTR they did for their accident, 25 which was a multiple misloading where they modeled all NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
60 1 of the LTAs together, they did an analysis. They got 2 acceptable results, but it didn't have the copious 3 margins that the others did. But they also didn't 4 explicitly credit the IFBA, so I took that into 5 account. I said they're going to have a margin there 6 as well so we don't need to go any further.
7 Based on that, we have reasonable 8 assurance that this license amendment and the LTAs 9 will meet 10 CFR 50.68(b)(2) and (b)(3), which are the 10 new fuel storage, fresh fuel high and dry, fuel rod 11 moderation, and flooded conditions. And then also 12 (b)(4), which is the spent fuel pool, which is 13 considered a fully flooded application.
14 So we think that that's going to be that.
15 It's reasonable to have an exception to 10 CFR 16 50.68(b)(7), which is the enrichment limit. That 17 concludes my presentation.
18 MR. MARKLEY: Mike Markley again. I'm 19 Branch Chief of Licensing for Vogtle. We appreciate 20 the feedback that we've received from the subcommittee 21 and each of the members. This is a first of a kind 22 for us.
23 We know that we're going to be getting 24 batch loads of higher burn-up and higher enrichment, 25 and they're probably coming faster. So again, we want NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
61 1 to come to the ACRS early and often, get the feedback 2 from you, and take that into consideration as we go.
3 The NRC staff determined for this request 4 that there's reasonable assurance that the health and 5 safety of the public will not be endangered by 6 allowing SNC to use the four ATF LTAs in limiting 7 locations without completion of representative testing 8 for up to two cycles, except for the locations where 9 the LTAs may not be placed in core regions that have 10 been shown to be limiting with respect to control rod 11 analysis.
12 Again, we would appreciate a letter. We 13 don't need one to proceed, but we really do value the 14 ACRS' feedback. Thank you.
15 MEMBER BALLINGER: Thank you.
16 Questions from the members or consultants?
17 Hearing no others, thank you once again, both Southern 18 Nuclear and the staff, for a good presentation.
19 I think we're back to you, ma'am.
20 CHAIR REMPE: We need to give public 21 comment.
22 MEMBER BALLINGER: Sorry about that. Are 23 there any people out in the public that would like to 24 make a comment? If that's true, please state your 25 name and make your comment.
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62 1 Hearing none, back to you, Madam Chair.
2 CHAIR REMPE: Okay. This time I will 3 accept. Thank you. I believe you do have a letter.
4 Do you want to read it in?
5 MEMBER BALLINGER: Yes.
6 CHAIR REMPE: We're going to go off the 7 record.
8 (Whereupon, the above-entitled matter went 9 off the record at 1:43 p.m. and resumed at 3:29 p.m.)
10 CHAIR REMPE: Okay. It's about 3:30 here 11 on the East Coast. I'm going to turn the meeting back 12 over to Member March-Leuba to lead us through our 13 third topic for today.
14 MEMBER MARCH-LEUBA: So the topic right 15 now is ARITA from Framatome, which is their transient 16 analysis methodology for essentially everything but 17 LOCA and logic.
18 We covered this topic in our subcommittee 19 meeting on June 22nd. And without further ado, Greg 20 is going to give us some remarks from the staff.
21 MR. SUBER: Thank you. Good afternoon.
22 My name is Gregory Suber. I am the Deputy Director 23 for the Division of Operating Reactor Licensing in the 24 Office of Nuclear Reactor Regulations.
25 I'd like to thank the full committee for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
63 1 the opportunity to discuss the staff's review of the 2 Framatome topical report. As previously stated, the 3 staff did present their findings in the ACRS 4 subcommittee meeting in June. We will be providing a 5 high-level overview of the findings and conclusions 6 today.
7 As mentioned in the opening remarks to the 8 subcommittee, this is an effort that culminated over 9 a four-year period. We appreciate Framatome's efforts 10 in working with the staff and resolving some very 11 difficult and complex issues.
12 The staff also appreciates the ACRS review 13 of this topical report and the safety evaluation. The 14 staff plans to issue the final safety evaluation 15 either later this month or hopefully early in August.
16 I look forward to your comments. Thank you.
17 MEMBER MARCH-LEUBA: Thanks, Greg.
18 From Framatome, Allan?
19 MR. McGINNIS: Hi. I'm Allan McGinnis, 20 Licensing Manager for Framatome. I just want to take 21 a second to thank the ACRS members for their time 22 today and listening to the information that's going to 23 be provided on our ARITA topical report and the NRC's 24 draft safety evaluation.
25 ARITA is a methodology that is a key part NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
64 1 of Framatome's efforts to upgrade our current PWR 2 analysis methodologies with more sophisticated 3 methods. As such, ARITA is going to be part of the 4 basis and platform for further development in advanced 5 fuel management and accident-tolerant fuel.
6 Framatome has a great deal of resources 7 invested in ARITA. We are very anxious to get the 8 final safety evaluation issued so that we can 9 implement the methodology and its benefits for the 10 industry. Thank you.
11 MEMBER MARCH-LEUBA: I wanted to remind 12 everybody that this is an open session. So if we want 13 to discuss questions, don't mention numbers and we'll 14 stay safe.
15 MR. McGINNIS: I also want to thank the 16 NRC for their efforts. They've put a lot of effort 17 in, especially in these last several months to get the 18 safety evaluation out and work with us on some last-19 minute issues. We appreciate that.
20 I'm going to go ahead and pass it on to 21 Buck Barner, who has been the lead on this project for 22 the majority of the development and review.
23 MR. BARNER: Thank you. This is Buck 24 Barner. I'll be presenting today on ARITA.
25 I want to first off say thank you and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 1 welcome. I'm excited to be here to talk about ARITA 2 in this full ACRS meeting. I appreciate your 3 willingness to be here and listen to us.
4 We're excited about ARITA. We're excited 5 about what it brings to the industry. I'm looking 6 forward to getting it implemented for the first time 7 and excited to hear that we're moving forward with the 8 SE. With that being said, we'll go ahead and move on 9 now into the presentation.
10 So to start off, we'll start with an 11 overview and go into some background and history of 12 the topical. We'll look at the approval request and 13 the range for applicability, describe the evaluation 14 model, and finalize with a summary.
15 Next slide.
16 So what is ARITA? ARITA stands for 17 ARTEMIS/RELAP Integrated Transient Analysis 18 Methodology. As has been said, it is a non-LOCA 19 method. It covers the Chapter 15 non-LOCA events 20 except for the control rod injection.
21 It is a non-parametric approach, which is 22 a novel approach for non-LOCA methodologies, and does 23 employ a Monte Carlo sampling approach. It was 24 developed using the guidance in SRP 15.0.2. It does 25 have some elements of the in-depth process, but was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
66 1 generally followed using the 15.0.2 guidance.
2 Along with the standard non-LOCA 3 methodology that's in there, we did address several 4 additional topics. We'll talk about the motivation 5 for that in future slides. That will also include 6 methodologies for mixed-core power distribution 7 control set points and fuel assembly reconstitution.
8 Next slide.
9 To give you some background and history, 10 there's a little time line across the top here. As we 11 move through the slides, we'll progress from left to 12 right, starting up there with some advanced codes of 13 method development.
14 In the 2006 time frame is when we began 15 development of a new set of advanced PWR codes and 16 methods, really with a focus on our neutronics solver 17 ARCADIA, the core TH solver COBRA-FLX, and fuel 18 performance code GALILEO. Around the same time in the 19 industry, there was also a push to replace legacy 20 methods, which was fortuitous and in good timing with 21 what we were doing internally.
22 So with those two things in mind, we moved 23 forward with a goal with several aspects in mind.
24 That was to ensure that we're using our state-of-the-25 art modeling. We wanted to use our best codes and our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
67 1 best capabilities that we could. We wanted to 2 implement the best practices we have from decades of 3 our experience in all three of our regions -- the US, 4 French, and German -- and really take advantage of 5 that experience that we have.
6 We wanted to simplify the topical reports 7 and interdependencies to reduce complexity. So we 8 wanted to take what had traditionally been in several 9 smaller topical reports and combine them into a 10 single, consistent topical report. And we also wanted 11 to do all this to facilitate development for future 12 method development and continuing innovations.
13 After these first three codes were 14 approved, our first methodology topical area was 15 approved in 2020. And now today we're here talking 16 about ARITA, which represents that final realization 17 of our objective to bring innovation and improvement 18 to the industry through our advanced codes and 19 methods.
20 Next slide.
21 So a brief time line of the developments.
22 We began working on ARITA and having pre-submittal 23 meetings February 2015. We submitted the topical 24 report in August 2018. After that time, we moved into 25 the licensing and review phase.
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68 1 We had the first set of RAIs on December 2 '19. I won't go into the details. We had a lot of 3 interaction with the NRC through RAIs and resolutions 4 to those, audits back and forth, and ultimately over 5 the last five years ended up with the draft SE, which 6 was provided in April of 2023.
7 So we've been committed to this topical, 8 committed to the industry to make sure that we get 9 this thing through. We want to see it through, and 10 we're excited to hear that the SE is at the final 11 steps.
12 Next slide.
13 Moving forward, what does this mean and 14 what does it bring to the industry? Now that we have 15 advanced modeling of actual plant behavior, it helps 16 us understand plant response and safety margins 17 better. With this understanding, it actually helps us 18 focus in areas where safety is really most important 19 and ensuring compliance with the regulations.
20 So events where safety is really being 21 challenged is where we're able to focus and really 22 understand things better, have a better understanding 23 of wherever margins are, and ensure that the plants 24 remain safe.
25 Along with safety, it also helps us bring NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
69 1 new value to the industry, things like addressing 2 regulatory changes. Reg Guide 1.236, it actually 3 addresses through area but any additional regulatory 4 changes, we believe ARITA sets us up well for 5 addressing any sort of new regulatory changes that may 6 be coming.
7 Also core design and optimization, and 8 then it also helps us move forward with advanced 9 initiatives like advanced fuel management with 10 increased enrichment and high burn-up. So all this 11 allows us with our higher fidelity simulation to 12 provide increased understanding and confidence in our 13 plant safety.
14 Next slide.
15 So range of applicability, we're looking 16 for a Chapter 15 methodology excluding control rod 17 injection. As I mentioned before, we're also looking 18 for applicability to our mixed-core methods, methods 19 to analyze the power density limiting condition of 20 operations and the core safety limit lines, power 21 distribution control, and fuel assembly 22 reconstitution.
23 It's generally applicable to Westinghouse 24 pressurized water reactor designs, as well as CE 25 pressurized water reactor designs. It does use NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 1 approved critical heat force correlations that are not 2 applicable to the correlations in the topical itself, 3 but it will use approved critical heat force 4 correlations.
5 It's mostly within the range of applicable 6 constituent codes that are used within the 7 methodology, which include ARTEMIS/RELAP, S-RELAP5, 8 COBRA-FLX, and GALILEO.
9 Furthermore about the evaluation model, we 10 talked about the codes. ARTEMIS is our 3D nodal 11 simulator code, previously approved in ANP-10297.
12 That's the ARCADIA topical report, which also includes 13 the APOLLO2-A code. That was originally submitted and 14 approved in 2013 and a supplement in 2018.
15 COBRA-FLX is the subchannel thermal-16 hydraulics code, which was approved in 2010. GALILEO 17 is the fuel performance code. That was approved in 18 2020. And S-RELAP5 is our system thermal-hydraulics 19 code that was previously applied in EMF-2310, 20 originally in 2004 and again in 2011.
21 So using these codes, we have developed 22 what we call a three-evaluation-model variance, which 23 we'll refer to as the coupled system thermal-hydraulic 24 and neutronics model, the 0D thermal-hydraulics model, 25 and the static core evaluation model. We'll get into NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
71 1 more detail on that in the next slides.
2 I'll review the coupling and the 3 evaluation model variance. First we'll talk about the 4 coupled EM. I have an illustration here on the left-5 hand side describing that, with ARTEMIS on the left 6 and S-RELAP5 on the right.
7 Here we have ARTEMIS performing the 3D 8 core neutronics modeling and simulation, with S-RELAP5 9 providing the system thermal-hydraulics simulation, 10 with RELAP passing the mass flow and temperature and 11 core outlet pressure to ARTEMIS, and ARTEMIS passing 12 the core power back to S-RELAP5 through the sector 13 heating rates.
14 That also can work with boron 15 concentrations if they're important to the specific 16 event. We regard it as an optional transfer, only if 17 it's needed.
18 This is what's used in the coupled EM 19 evaluation. It does this in a transient mode for the 20 Chapter 15 transient analyses. There's a time-21 dependent multi-physics solution there. This model is 22 used for cases where SAFDLs are being evaluated and 23 non-SAFDLs.
24 MEMBER MARCH-LEUBA: If I understand 25 correctly, ARTEMIS is the core simulation that has the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 1 3D core inside the core. RELAP is used for the 2 environmental conditions, simulators, all that.
3 MR. BARNER: Correct.
4 MEMBER MARCH-LEUBA: I don't want to call 5 it the balance of plant, but it's kind of like the 6 balance of plant. Let's call it the balance of the 7 core.
8 MR. BARNER: Balance of the core.
9 MEMBER MARCH-LEUBA: Yes. Okay.
10 MR. BARNER: The second one is what we 11 call the 0DEM. This is similar to what we've seen in 12 legacy methods where data is generated in the 13 neutronics simulator, or ARTEMIS in this case, which 14 is provided to the simulator in S-RELAP5.
15 And S-RELAP5 is there to run stand-alone.
16 Note that this is only used for non-SAFDL events, 17 things like overpressure, secondary overpressure.
18 MEMBER MARCH-LEUBA: So this is when 19 you're trying to provide figures that are mostly the 20 primary boundary? Overpressure, temperatures, things 21 like that, not core performance?
22 MR. BARNER: Correct. And then finally, 23 we have the stack EM, which is Artemis used in stand-24 alone mode. That's used for events where the system 25 thermal-hydraulics is needed.
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73 1 So for static events such as a statically 2 misaligned rod or a mis-load analysis, then the 3 ARTEMIS core model is used. And that's also used for 4 SAFDLs. You'll see that whenever a SAFDL is being 5 evaluated, the ARTEMIS 3D model is what's being used 6 to evaluate the standards.
7 A little bit of information on the 8 statistical approach. Again, it's a Monte Carlo 9 sampling approach. If you feel it better manages the 10 complexity than the couple of non-LOCA transient 11 analyses, there's a lot of moving parts to this. It 12 allows us to sample that appropriately.
13 It uses a non-parametric approach based on 14 Wilks method. It uses ordered statistics to make a 15 95/95 statement on the figures and merits.
16 For cases where there's events where 17 multiple figures and merits are required for a single 18 event, it does account for that and has an ability to 19 handle multiple figures and merits from a statistical 20 standpoint.
21 I'll also note that some parameters are 22 still biased. This is a statistical method. There 23 are places where we have bias and moved things to be 24 conservative and be more consistent with the safety 25 analysis and not so much a best estimate. So we do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
74 1 try to be realistic, and also provide additional 2 margins and conservatisms through biasing some 3 parameters.
4 Because we are using a non-parametric 5 approach, we don't necessarily need to develop full 6 distribution results. We can use a smaller number of 7 realizations, say 59 cases, to make that actual 8 statistical statement. And this is similar to what's 9 been done in previous approaches I've presented to the 10 ACRS and have been approved by the NRC before.
11 If you look over here on the left, this is 12 just an illustration of a typical transient where you 13 may see 59 realizations of a transient across time.
14 And then you're able to use this to help with your 15 statistical statement. This is the type of output 16 that you might get from this.
17 MEMBER MARCH-LEUBA: Can you give me an 18 example of when you use bias in parameters as opposed 19 to sample uncertainty?
20 The way I understand it is when actually 21 obtaining the certainty is too complicated -- one 22 example you gave us during the subcommittee meeting 23 was the position of the control rod when you have an 24 asymmetric -- give us an example of when you bought 25 it.
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75 1 MR. BARNER: Peaking would be a place 2 where we would use a bias. So if you're radial 3 peaking, you'll use a bias in that case.
4 MEMBER MARCH-LEUBA: So you're using a 5 limiting radial distribution?
6 MR. BARNER: We'll pick it up to the tech 7 spec.
8 MEMBER MARCH-LEUBA: Instead of searching 9 through your core for the words, conditions, you just 10 place it on tech spec?
11 MR. BARNER: The details, I would have to 12 have a full session. But yes, that's the general.
13 MEMBER MARCH-LEUBA: Okay. Thank you.
14 MEMBER ROBERTS: What do your outlet 15 distributions look like? You talked about that in the 16 opening session.
17 Would you say bifurcations or something 18 kind of fishy happening past the 95 percent point in 19 the distributions or do they look pretty much normal?
20 MR. BARNER: We have not seen anything 21 like that, not something we would look at we saw 22 something in distribution that looked strange, we 23 would investigate it. We have not seen that in 24 anything that we've done yet.
25 MEMBER ROBERTS: Is there anything in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76 1 methodology that looks for that?
2 MR. BARNER: There are some things we 3 built in for certain parameters where we needed to 4 look at certain areas, but that will probably be 5 talked about more in closed soon. We do that as the 6 methodology --
7 MEMBER MARCH-LEUBA: If you really want 8 the answer to the question, we can sit up a clothes 9 station in five minutes. So don't strain yourself.
10 Just do not give any proprietary questions here. We 11 will select the proper venue.
12 MR. BARNER: So I just want to summarize 13 ARITA. It is a 3D coupled statistical, non-LOCA 14 methodology applicable to see plants.
15 It represents a metered milestone in our 16 commitment to state-of-the-art modeling and it 17 incorporates decades of art industry experience, sets 18 us up, sets the foundation for future development 19 moving forward. Ultimately, the insurance confidence 20 and plant safety compliance with all regulations and 21 requirements is higher fidelity modeling.
22 I think that's all we have.
23 MEMBER HALNON: Just real quick, earlier 24 in the presentation you talked about enhanced 25 marbling, innovation, all these great things that you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 1 did. Being a layman at this kind of stuff, could you 2 just give me one or two things that are innovative and 3 enhanced in this?
4 You talked about Monte Carlo. I've heard 5 that around forever. You talked about code coupling 6 and I've heard that forever. So what is it that 7 you've spent your last decade on?
8 MR. BARNER: Good question. So yes and 9 yes. It's both a statistical approach and the core 10 coupling, as well as the 3D transient.
11 The ARTEMIS simulator gives us the real 12 feedback. The fact that we have that coupling 13 feedback is what really gives us the ability to see 14 things.
15 The Monte Carlo sampling is around that to 16 take advantage of that coupling and be sure we're not 17 breaking the coupling by doing things that the codes 18 themselves wouldn't allow by taking unrealistic 19 conditions and combining them with the other.
20 So when we were able to do that and couple 21 them all together, that's what really took us to the 22 next level. You said coupling is nothing new and the 23 statistical piece is nothing new, but really it was 24 the combining of those two pieces together --
25 MEMBER HALNON: So more power in smarter NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 1 brains putting it together and tying the knowledge 2 together so that it all works together?
3 MR. BARNER: Yes.
4 MEMBER HALNON: All right. Thanks.
5 MEMBER MARCH-LEUBA: Any more questions 6 from the members? Then let's switch with the staff.
7 I want you to introduce yourselves and 8 speak clearly into the microphone.
9 MR. GEELHOOD: I'll go ahead and get 10 started. I'm Ken Geelhood. I'm the Project Manager 11 for Reports. I work in NRR and I'm Project Manager.
12 MR. HELLER: Hi. I'm Kevin Heller. I 13 work in the Nuclear Methods and Fuel Analysis Branch.
14 I'm obviously one of the reviewers on ARITA.
15 MR. LEHNING: And John Lehning. I'm also 16 a technical reviewer in the Nuclear Methods and Fuel 17 Analysis Branch.
18 MR. HELLER: All right. Let's launch into 19 it then. Again, I'm Kevin Heller. With me here is 20 John Lehning. It's our pleasure to be here in front 21 of the full committee to present our review of the 22 ARITA topical report. We certainly want to thank you 23 for your time.
24 As I noted on the slide here, I just 25 wanted to quickly point out Pacific National Northwest NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 1 Laboratory served as our consultant on this review and 2 provided input that the staff used as a basis for the 3 draft safety evaluation.
4 Next slide, please.
5 Okay. I think Framatome did a good job 6 talking about the evaluation model and what that is.
7 So really, the only thing I want to point out on this 8 slide is that the staff's review of ARITA really did 9 focus more so on the unique aspects of the 10 methodology, so primarily the calculational procedure 11 and the uncertainty treatment. I've got a couple of 12 slides on that.
13 MEMBER MARCH-LEUBA: Is it the case that 14 most methodologies have already been approved? The 15 individual components, codes were already approved?
16 MR. HELLER: Yes, that would be correct.
17 A large portion of these codes, yes.
18 MEMBER MARCH-LEUBA: And you leveraged 19 that for approval? You didn't have to redo it?
20 MR. HELLER: Exactly. We'll see that.
21 Next slide, please.
22 Okay. To talk a little bit about the 23 regulatory requirements and guidance, the full list of 24 the regulations and guidance that the staff used can 25 be found within the safety evaluation. The list here NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
80 1 includes the key requirements and guidance.
2 In particular I wanted to point out 3 Chapter 15 of the Standard Review Plan, which has 4 guidance on how the staff ought to review evaluation 5 models. And of course, the evaluation model 6 development and assessment process or MDAP, which is 7 in Regulatory Guide 1.203, that was used to structure 8 parts of the safety evaluation in order to ensure it 9 was a comprehensive assessment of the ARITA evaluation 10 model.
11 With that, I think we can go to the next 12 slide.
13 Okay. So a little bit of quick overview 14 here on the review history. ARITA, the review can 15 actually be broken up into four phases.
16 So if you actually proceed forward one 17 more time? There we go. We can see the four review 18 phases.
19 I really want to point out that ARITA was 20 one of the most complex, challenging, and intensive 21 reviews that I think either of us has really been a 22 part of in our time at the NRC, not only because there 23 were a number of first-of-a-kind issues but just given 24 the complexity of the subject material, which in the 25 past might have spanned several different topical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 1 reports.
2 We're seeing a lot of it really all in one 3 here. A wide range of in-scope event responses, 4 physical phenomena to be understood, limitations, and 5 available data for doing non-LOCA statistical 6 analyses, and the intricacies of the calculational 7 procedure that Framatome proposed.
8 One of the main challenges which probably 9 contributed to this is that to some measure, the sheer 10 amount of information that was needed to be reviewed 11 for this particular topical report. It's clear there 12 was a lot of work and a lot of information that went 13 into this.
14 Starting initially with beginning the 15 review, we really didn't have enough information to 16 begin drafting the safety evaluation or really know 17 which way some of the identified issues would actually 18 be resolved. So it took a bit of time to resolve a 19 number of the RAIs. Many major issues would actually 20 not end up getting resolved until a couple of years or 21 maybe a little bit later within the review effort.
22 So to kind of put that in a little bit of 23 perspective, the total volume of RAI responses was 24 basically equivalent to another topical report.
25 There's an additional 800 pages of information that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 1 was provided on top of the original submittal for 2 about 1,600 pages in total.
3 What I want to point out at the last phase 4 there, the draft SE phase, during that final phase 5 there was not a lot of interaction that the staff had 6 with Framatome in that particular portion because the 7 staff was focused on completing the draft safety 8 evaluation.
9 There were a handful of new limitations 10 and conditions that were identified to close out open 11 issues that Framatome's final RAI responses didn't 12 resolve, but we left off the review at Framatome's 13 request given the needed safety evaluation date.
14 So all parties understood that there were 15 some issues that would probably have to get resolved 16 through additional limitations and conditions. And 17 that's why you can see a number of the limitations and 18 conditions afford alternative approaches or 19 justifications.
20 All right. Next slide, please.
21 Okay. So talking about the calculational 22 procedures, what I'm going to try to do is high-level 23 talk about what it is that we looked at and why we 24 found it to be acceptable. When it comes to the 25 calculational procedure, the key review issue that the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
83 1 staff identified was associated with assessing the 2 ARITA methodology against regulatory guidance 3 concerning suitably conservative initial conditions.
4 When applying this guidance, the staff 5 considered what is the objective of the ARITA 6 methodology with respect to its intended statistical 7 statement and how is that going to be achieved.
8 Ultimately, the tech specs and the safety analyses 9 need to be in alignment.
10 In consideration of the staff's concern 11 was that the as-proposed method would not provide 12 conservative results for plant operation throughout 13 the allowed operating domain. So during the course of 14 the review, through interaction with the NRC staff and 15 through RAI responses, Framatome modified the 16 calculational procedure to establish suitably 17 conservative event definitions.
18 The revised approach is really 19 characterized -- I've got a couple of bullets here.
20 Achieving a 95/95 or a 95 percent probability with 95 21 percent confidence tolerance limits over the allowed 22 operating domain. Utilizing conservative sampling for 23 highly influential parameters. Really, it's kind of 24 a compromise between a traditional approach and a 25 best-estimate sampling approach.
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84 1 MEMBER MARCH-LEUBA: We talked in the 2 subcommittee about this best-estimate plus 3 uncertainty. We can eventually agree that this 4 approach is best-estimate plus uncertainty plus 5 biases, biases being conservative initial conditions, 6 conservative power distributions. Do you agree on 7 that?
8 MR. HELLER: Yes, I would agree with that.
9 There were a number of biases that were utilized for 10 conservatisms. I'll actually speak to that a little 11 bit on a future slide.
12 MEMBER MARCH-LEUBA: If you state it in 13 writing, we'll talk about whether calling it biases or 14 penalty factors. What do you prefer?
15 To a mathematician, bias means a lot to 16 you. But as a member of the public, I don't know what 17 -- a bias is something psychological. We'll have a 18 spirited discussion later.
19 MR. HELLER: It sounds like an interesting 20 philosophical discussion.
21 MEMBER MARCH-LEUBA: It is the same thing, 22 right?
23 MR. HELLER: I think effectively, as far 24 as mechanistically at the end of the day, yes.
25 So when it comes to the calculational NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 1 procedure, it was through the establishment of this 2 conservative event definition that the staff 3 ultimately found the calculational procedure to be 4 acceptable.
5 Next slide, please.
6 All right. So talking about the treatment 7 of uncertainty. NRC staff utilized the guidance 8 provided in Chapter 15.0.2 of the Standard Review Plan 9 to review ARITA's methodology treatment of 10 uncertainty. And per this guidance analysis, it 11 should address all important sources of uncertainty.
12 When we're dealing with a statistic-based 13 uncertainty method, one of the major challenges is 14 developing sufficient data to support the uncertainty 15 distributions for a wide array of phenomenon and 16 processes. With that in mind, the NRC staff's review 17 of the ARITA methodology, its uncertainty treatment 18 focused on three areas.
19 The first area was the uncertainty and the 20 input parameters, specifically the distributions and 21 their technical basis. Staff assessed the uncertainty 22 treatment for all the input parameters. As modified 23 by RAIs and limitations and conditions, the staff 24 found the treatment of those uncertainties to be 25 acceptable.
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86 1 The second area was the uncertainty in the 2 calculated figures of merit, specifically the 3 statistical basis for the method to derive the 4 uncertainties in those figures of merit. Framatome 5 proposed approaches to determine the uncertainties in 6 those figures of merit on univariate and multi-varied 7 conditions.
8 Staff assessed those approaches and found 9 them acceptable because they appropriately determined 10 a 95/95 tolerance limit. We actually discussed the 11 details of this within the subcommittee meeting.
12 And then lastly, ARITA's overall 13 statistical calculational procedure, specifically 14 whether the procedure assures fidelity of the results.
15 The staff reviewed the process proposed by Framatome 16 and found it acceptable because it's a statistically 17 rigorous approach that assures 95/95 tolerance limits 18 are appropriately calculated.
19 And it appropriately avoids biasing of the 20 procedure itself based on knowledge of the 21 realizations, which could degrade tolerance limits.
22 So it avoids that.
23 Next slide.
24 MEMBER ROBERTS: I guess I could ask you 25 the same question I asked Framatome. With non-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
87 1 parametric statistics, it's possible to have the 95 2 percent, which is kind of good because you're actually 3 getting more information about the potential worst-4 case scenarios that could happen. Sometimes depending 5 on what you have, they could be very bad.
6 In this case, I'm guessing because you've 7 got so many biased parameters, there aren't that many 8 parameters left that could vary and conspire to come 9 up with some sort of localized minimum or maximum. I 10 was wondering if that's what you'd look for.
11 MR. LEHNING: This John Lehning from the 12 staff. That was a concern that we brought up in the 13 early part of the review.
14 Where I think we ended up, as you alluded 15 to in the question, by focusing in on, as Kevin 16 mentioned, some of the highly influential parameters 17 and some of the biases that Framatome had input into 18 its method, we didn't see evidence of that. And we 19 don't believe there's a reason to think that there 20 would be some special amount of bifurcation or 21 spreading of this distribution beyond what's typical.
22 And we think, in fact, it might be less of 23 an issue for this method because of the biases as 24 compared to some other methods out there that may 25 sample things even more broadly without some the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 1 biases and special focus on some of these highly 2 influential parameters.
3 MEMBER ROBERTS: Okay. Thank you.
4 MR. HELLER: All right. Next slide.
5 Okay, so validation assessment. In 6 practical terms, assessment and validation of an 7 evaluation model involves checking the results of its 8 computations against test facility measurements for 9 reference data to provide insight into the credibility 10 of the model's predictive capability for similar 11 postulated conditions.
12 Framatome formulated the bulk of that 13 assessment for the ARITA methodology in terms of the 14 constituent codes, which we already mentioned. There 15 was some interval validation comparison for those 16 constituent codes, but there was also a lot of code or 17 model-based validation provided.
18 It was often cited by past review efforts.
19 That frequently made use of measured data and 20 experimental data. So staff examined those past 21 reviews in consideration of the intended application 22 of the ARITA methodology and concluded they largely 23 support the ARITA methodology.
24 In cases where there was limited 25 validation for codes or models, staff concluded that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
89 1 some of the conservatisms that were present in the 2 methodology were limitations and conditions acceptably 3 compensated for this. Therefore, the staff concluded 4 that the constituent codes have sufficient 5 capabilities for modeling the phenomena that are 6 relevant to the set of events within the scope of 7 ARITA.
8 Talking about then the evaluation model 9 variance themselves, for the 0D and static evaluation 10 model variance, staff found those to be very similar 11 to stand-alone treatments. And therefore, ultimately 12 found those to be acceptable.
13 For the coupled evaluation model variant, 14 the staff ended up assessing the information exchange 15 between the coupled codes. And also Framatome 16 provided integral assessments. Those results were 17 found to be acceptable.
18 So the NRC staff ultimately then has 19 reasonable assurance that the ARITA methodology has 20 adequate capability of modeling Westinghouse and 21 combustion engineering PWRs for the applicable events.
22 For supplementary evaluation model 23 features, there's a series of what we refer to as 24 supplemental evaluation model features for ARITA.
25 These can serve as additional functionalities that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
90 1 Framatome has incorporated into the methodology.
2 For the boron dilution, the mixed core 3 treatment, and for fuel assembly reconstitution, staff 4 found these were acceptable when consideration was 5 given to code conservatisms in some limitations and 6 conditions.
7 For the set points and the power 8 distribution control, the staff ultimately found that 9 these were evolutionary updates to existing methods, 10 retaining the overall existing approaches. And 11 therefore, the staff also found them to be acceptable.
12 Next slide, please.
13 At the conclusion of drafting the safety 14 evaluation, staff imposed a total of 28 limitations 15 and conditions. These can be broadly broken down as 16 seen in the table here.
17 So there's about 16 of them that, really, 18 they're there to assure appropriate treatment of 19 uncertainties or the application of a statistical 20 process. There's approximately ten that are there to 21 assure appropriate application of the ARITA 22 methodology.
23 And then there are two to assure that 24 conservatisms which were cited as justification for 25 the methodology are representative of future fuel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 1 designs in existing plants.
2 The staff found that these limitations and 3 conditions were necessary in order to ensure the 4 acceptability of the methodology based on a full 5 review of the material that was contained in the 6 docketed materials that Framatome submitted.
7 The last point then. After receiving the 8 draft safety evaluation for proprietary review, 9 Framatome did inform the NRC staff it intends to 10 pursue resolution of some of the concerns underlying 11 two of the limitations and conditions, 18 and 19, in 12 a future regulatory review process.
13 We've had some dialogue as to possible 14 avenues that could be pursued through that as far as 15 the vehicles. For example, an update or addition to 16 a topical report, inclusion within a plant license 17 amendment request, or another avenue. We anticipate 18 hearing from Framatome proposing a pathway in the near 19 future for that.
20 MEMBER MARCH-LEUBA: Can you give us a 21 high-level description of limitation conditions 18 and 22 19?
23 MR. HELLER: I was trying to do it off the 24 top of my head, but we do actually have a back-up 25 slide. Slide 10, I believe. It's actually after the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 1 conclusion.
2 MEMBER MARCH-LEUBA: I don't have it on 3 mine.
4 MR. HELLER: You don't have it?
5 You should have it there. There we go.
6 Okay.
7 MEMBER MARCH-LEUBA: We will have it if 8 you put it on the record.
9 MR. HELLER: Yes.
10 MEMBER MARCH-LEUBA: If it's on the 11 record, you have given us a copy.
12 MR. HELLER: Yes. So limitation and 13 condition 18 deals with the axial peaking factor.
14 Basically, what it says is licensees shall justify the 15 approach that Framatome proposed for treating its 16 uncertainty. Or they could conservatively estimate it 17 via what's iterated within the limitation and 18 condition, or they could justify another approach.
19 Really where this is coming from is the 20 staff couldn't conclude from the material that was 21 provided that Framatome's proposed approach directly 22 followed.
23 MEMBER MARCH-LEUBA: Couldn't we have 24 followed with a bias? That's typically for LOCA, 25 right, or not?
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93 1 MR. LEHNING: I think that in essence is 2 similar to what we've found is being done here. So 3 yes, there is a delicate balance here in terms of 4 that.
5 We did try to take what we thought was 6 somewhat of a bounding approach, especially on 19 7 because of the limitations and some of the data that 8 underlies that item. So I think that's probably what 9 I can say here in this session. We could go into --
10 MEMBER ROBERTS: I have a simple question.
11 So you had 28 total limitations and conditions. I 12 imagine some of those were legacy LNCs from the 13 previous reviews because certainly you always have 14 limitations and conditions.
15 Were some resolved from the past? Did a 16 number of them just not apply? How many are new?
17 MR. LEHNING: I think they basically are 18 new in general. And now the existing -- I think what 19 you're referring to is there's previous safety 20 evaluations that the staff has written on these other 21 methods.
22 We have words in our safety evaluation 23 that basically, the previous safety evaluations also 24 have to be respected. So we didn't incorporate any of 25 those existing limitations and conditions into our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
94 1 safety evaluation. The understanding is that those 2 methods would be used.
3 MEMBER ROBERTS: So these are kind of in 4 addition to the legacy applications. Okay. So 5 really, in a sense there's more than 28. There's 28 6 plus the old ones too. Okay. Those ones don't get 7 captured again in your SE?
8 MR. LEHNING: That's true. The way that 9 the methods work, there probably are some -- for 10 instance, if we were to look back at -- I think in 11 particular, the ones that are the clearest are the 12 ARTEMIS and GALILEO because those codes would be used 13 in the same manner basically that they're being used 14 in those code specifics or the topical reports that we 15 reviewed.
16 The S-RELAP5 might be a little bit 17 different in the sense of the realistic LOCA topical 18 report. There are some aspects of that that may or 19 may not necessarily -- some of the limitations might 20 trivially be satisfied because they're not really 21 applicable to non-LOCA analysis.
22 I'd say again, for the 28 items that are 23 in the safety evaluation for ARITA, a lot of those 24 deal with things that are specific to ARITA in terms 25 of the uncertainty distributions and the calculation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
95 1 procedure for that evaluation model. So there is not 2 a repeat of existing safety evaluation limitations and 3 conditions into this safety evaluation we've written 4 for ARITA.
5 MEMBER ROBERTS: Okay.
6 MR. HELLER: So for limitation and 7 condition 19, this is somewhat a similar situation in 8 that there was a proposed approach for treating the 9 uncertainty with the axial peaking factor for a 10 specific event. The staff could not conclude from the 11 information that was provided before the review ended 12 that the approach was appropriate.
13 MEMBER MARCH-LEUBA: The solution was the 14 staff proposed what they considered to be a 15 conservative treatment of the uncertainties, which 16 Framatome considers to be too conservative; is that 17 correct?
18 MR. HELLER: I think that would be a fair 19 assessment.
20 MEMBER MARCH-LEUBA: Probably the solution 21 would be the first LAR to be followed by a supplement.
22 That seems like a path forward, but that's my opinion.
23 MR. HELLER: So really, the last thing I 24 wanted to point out with this is -- you can see it 25 there at the bottom of the slide -- the introduction NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
96 1 of these limitations and conditions, they reflect only 2 where we ended the review based on the docketed 3 material. There was a cut-off date and it was 4 understood that there might be some things that end up 5 resulting. And these were two of them.
6 MEMBER MARCH-LEUBA: Okay. Thank you. Go 7 back to your conclusions.
8 VICE CHAIR KIRCHNER: It begs a different 9 question, a more generic question on limitations and 10 conditions. How many of them are based on physics and 11 how many of them are based on events? In other words, 12 the methods of the first order don't know it's a 13 streamline break.
14 You put those boundary conditions on the 15 system in the analysis but is it -- do you see where 16 I'm going with this? It sounds like a lot of 17 limitations and conditions on top of those that 18 already exist. So are they mainly because of 19 inadequacies and treating physical events or just the 20 range that the code is being applied to?
21 MR. LEHNING: Maybe I'll take a first 22 crack at that. I don't think a lot of them have to do 23 with basically physics or perceived inadequacies in 24 the way that the physics of the code works.
25 I think we feel pretty comfortable with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 1 those codes and the models. And any of the issues 2 with those would have been identified in these 3 previous reviews that we just discussed in response to 4 a previous question.
5 So I'd say the majority, if not all, of 6 these limitations and conditions really go to how the 7 calculation ought to be done and whether we agreed 8 with the specific way that Framatome said we're going 9 to -- maybe assumptions, inputs, or other things like 10 that much more so.
11 MR. HELLER: I would simply follow that up 12 with -- I guess I would ultimately end up reiterating 13 to a certain extent but using different words.
14 Alluding back to what Framatome said, the novel 15 aspects to this, these constituent codes are being 16 applied in a particular manner.
17 So you can kind of think of it as there's 18 a wrapper around those codes. It's the nature of that 19 wrapper that -- those limitations and conditions exist 20 to ensure that it's being applied appropriately.
21 VICE CHAIR KIRCHNER: Okay. I'm testing 22 you because I didn't want to hear that it was 23 something fundamental in the neutronics or whatever.
24 It's more how it's tied together and how it's applied.
25 Okay.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
98 1 MR. HELLER: Yes. That would be the case.
2 Okay, conclusions then. Staff found that 3 the ARITA methodology is acceptable for modeling in-4 scope Chapter 15 events from the Standard Review Plan.
5 This includes all three of the ARITA evaluation model 6 variants, the associated calculational process, and 7 the statistical uncertainty methodology. The staff 8 also found that the supplementary evaluation model 9 features are acceptable.
10 MEMBER MARCH-LEUBA: Can you explain what 11 those are?
12 MR. HELLER: Let me actually go back up in 13 my notes here. You can stay on this slide.
14 MEMBER MARCH-LEUBA: Is it boron?
15 MR. HELLER: Yes. Boron dilution, the 16 setpoints approach, the fuel assembly reconstitution, 17 the mixed core treatments.
18 MEMBER MARCH-LEUBA: That's good, a good 19 example. We have memorized the Section 3.8.
20 MR. HELLER: So the NRC staff found it was 21 acceptable. And the staff's conclusions then are 22 predicated upon two things.
23 First, the ARITA methodology being used 24 within its proposed range of applicability. And that 25 licensees acceptably address the limitations and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
99 1 conditions within the safety evaluation. That's it.
2 MEMBER MARCH-LEUBA: Thank you.
3 Any more questions from the members?
4 MEMBER HALNON: I just wanted to explore 5 back on your schedule slide. You had an acceptance 6 review. Did you take it back or just not accept it?
7 You said the first acceptance review was 8 completed and then you had to supplement it. I know 9 this doesn't affect the fact that everything is 10 acceptable and you guys have done a good and thorough 11 job of getting through this, but that's a long time 12 and a lot of RAIs.
13 MR. LEHNING: Basically, what happened 14 there was that we found information during the 15 acceptance review that staff felt was not fully 16 complete, and so there was a supplement. It wasn't --
17 MEMBER HALNON: So this was if you can 18 provide this, then we can -- it wasn't extended beyond 19 the 45 days?
20 MR. LEHNING: It wasn't rejected.
21 MEMBER HALNON: Okay.
22 MR. LEHNING: There's an option to reject 23 it, accept it, or to accept it with a supplement.
24 This went down the path of accepting it with a 25 supplement.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
100 1 (Simultaneous speaking.)
2 MR. LEHNING: Correct. That's true. It's 3 at this very coarse level of detail. Is all the 4 information we need there to do the full detailed 5 review? And we found that there was not. This yellow 6 bar represents basically the time to complete that.
7 MEMBER HALNON: As you ramped up your 8 understanding of it, you can see it getting more and 9 more detailed and more involved. Did you predict it 10 was going to take that long? Or did it kind of just 11 occur because of all of the technical complexities?
12 MR. LEHNING: We did. I think we had a 13 pretty clear idea that the review schedule was going 14 to be substantially longer than the standard two-year 15 review period that we assume for topical reports, but 16 there are a lot of uncertainties.
17 For example, there were some issues that 18 even in -- I think there were three pre-submittal 19 meetings for ARITA. There were some issues that we 20 identified there very early on that continued to 21 persist through the review and maybe didn't even get 22 resolved until June 2022.
23 That was because, I think, the vendor in 24 this case had a principal idea that they were in the 25 right, that their approach was justified, and it took NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
101 1 some time before we could get to that point where 2 there was this agreement on going forward with that.
3 So yes, the exact length of time was not 4 completely clear, but we had a clear idea from the 5 very beginning. Framatome was aware of that when they 6 got into the review.
7 MEMBER HALNON: I mean, the 19 meetings 8 and all that, we've approved entire reactor plants in 9 much less time. I can only imagine the complexity and 10 how deep you guys got. It's good. I'm glad you 11 presented that because it gives us the context of just 12 how complex it was.
13 MEMBER MARCH-LEUBA: Any more questions 14 from members, even the member on the phone line? I 15 assume if you don't raise your hand, you don't want to 16 talk.
17 MEMBER DIMITRIJEVIC: No questions, Jose.
18 MEMBER MARCH-LEUBA: Thanks, Vesna.
19 So at this point, I'd like to open the 20 floor for members of the public. If somebody wants to 21 place a comment on the open transcript, please do so 22 now. I hear none.
23 Madam Chair?
24 CHAIR REMPE: Thank you.
25 At this point, I'd like to tell the court NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
102 1 reporter we're going off the record for the rest of 2 this meeting. Thanks for your support.
3 (Whereupon, the above-entitled matter went 4 off the record at 4:20 p.m.)
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
Lead Test Assembly (LTA) License Amendment Request (LAR) - 707th ACRS Meeting -
July 12-14, 2023 Ryan Joyce - SNC July 12th, 2023
Agenda
- Vogtle LTA Program Overview
- Requested Exemptions
- Summary of In-Pile Testing (ADOPT, AXIOM, Cr-coated cladding)
- Analysis 2
Vogtle LTA Program Overview
- Initial goals of program (within scope of proposed LAR):
- Irradiate higher enriched fuel in a commercial reactor to generate data in support of future licensing applications
- Obtain additional data for accident tolerant fuel (ATF) materials
- Future goal of program (outside scope of proposed LAR):
- Support licensing applications for higher burnup fuels
- Four rods in each LTA with enrichment up to 6 wt.% 235U
- AXIOM high performance fuel rod cladding (WCAP-18546-P/NP-A)
- EnCore chromium coated cladding
- ADOPT' doped fuel material for non-IFBA (Integral Fuel Burnable Absorbers) rods (WCAP-18482-P/NP-A)
- Standard (undoped) fuel material for IFBA rods ADOPT, AXIOM, BEACON, Optimized ZIRLO and EnCore are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.
3
Vogtle LTA LAR- As Supplemented on Sept. 13, 2022 and May 5, 2023
- The license amendment requested the following changes:
- Placement in limiting core regions (except for control rod ejection transients)
- Inclusion of advanced coated cladding with doped or standard fuel material, and
- Having a maximum nominal 235U enrichment of 6.0 wt.%
- TS 3.7.18 Fuel Assembly Storage in the Fuel Storage Pool, and TS 4.3 Fuel Storage will be changed to reflect LTA:
- Spent and new fuel storage restrictions
- Allowance for maximum nominal 235U enrichment of 6.0 wt.% in the New Fuel storage racks 4
- An exemption to §50.46 and §50 Appendix K was requested for use of AXIOM cladding
- A more restrictive embrittlement criterion was used in conjunction with the 17%
maximum local oxidation criterion
- AXIOM cladding topical presents the data in support of this application 5
Exemption Requests (contd)
- Licensing basis to change from §70.24 to §50.68
- Exemption is needed to §50.68(b)(7) to allow these LTAs to have greater than 5 wt.% 235U rods
- Technical Justification:
- Intent of rule (to preclude inadvertent criticality) being maintained
- Administrative controls will be in place for temporary storage of LTAs in Traveller-B containers prior to LTA placement in their designated storage locations
- Remaining §50.68(b) criticality requirements are unaffected and continue to be implemented
- New Fuel Storage Vault
- Spent Fuel Pool 6
Summary of In-Pile Testing (ADOPT, AXIOM, Cr-coated cladding)
- Millstone Unit 3 LTAs
- AXIOM cladding
- Completed 3 cycles of irradiation
- Byron Unit 2 LTAs
- ADOPT pellets and Cr-coated Optimized ZIRLO cladding
- Completed 2 cycles of irradiation
- Hot Cell PIE after 1st irradiation cycle
- Significant BWR irradiation experience for ADOPT pellets (WCAP-18482-P-A) and PWR irradiation experience for AXIOM cladding (WCAP-18546-P-A) was used to establish corresponding irradiation properties databases The primary novel feature that does not have prior commercial industry operating experience is the 4 fuel rods per LTA (out of 50,952 total fuel rods in the reactor core) with enrichment greater than 5 wt. % 235U.
7
Analysis
- LTAs will lead the core during portions of steady-state operation and during some transient conditions
- Leading the core = highest linear heat generation rate
- LTAs do not establish core operating limits
- The LTAs and co-resident fuel will continue to meet all Technical Specifications (TS) 2.1 Safety Limits, 3.1 Reactivity Control, and TS 3.2 Power Distribution Limit requirements
- The analytical methods used to determine the core operating limits will be those previously reviewed and approved by the NRC (per TS 5.6.5)
- No new methods were employed in the evaluation of the LTAs
- Where pertinent, newer approved methods were used to confirm efficacy of the current licensing basis methods
- A separate LAR will be required to go above the licensed fuel rod burnup limit (not expected until 3rd cycle of operation) 8
Analysis
- BASH evaluation model (EM) for LBLOCAs and NOTRUMP EM for SBLOCAs are acceptable for evaluating the LTAs
- LBLOCA meets all acceptance criteria per §50.46
- No impact to peak clad temperature (PCT)
- Maximum local oxidation (MLO), and core-wide oxidation (maximum hydrogen generation) meet acceptance criteria
- Demonstrated inconsequential impact associated with radiative heat transfer of Cr-coated rods
- Non-LOCA/Transient Analysis
- No impact on AOR for transients dependent on core-average effects
- Negligible impact of 4 LTAs on core-average heat transfer characteristics, decay heat, initial core stored energy
- For events dependent on local effects (SLB, Locked Rotor, Loss of Flow, RWFS, Rod Ejection)
- No impact due to LTA on approved non-LOCA codes, methods, or relevant acceptance criteria
- The LTAs will not be placed in limiting core locations for Rod Ejection 9
Analysis
- Source Term and Dose Consequences
- Utilized ORIGEN-ARP to generate core inventories for the LTAs
- Performed parametric runs across a range of inputs (power, enrichment, burnup) that is broader than expected in operation
- No impact to source term or dose consequences (AOR source term remains bounding)
- Bounding source term will be confirmed per Reload Analysis on a cycle specific basis
- Fuel Rod Performance and T/H Design
- The latest fuel performance models, PAD5 (WCAP-17642-P-A, Revision 1), ADOPT Fuel (WCAP18482PA),
AXIOM cladding (WCAP-18546-P-A), are used to explicitly model LTA features
- Improved corrosion resistance of Cr coated rods are conservatively neglected
- PAD5 approval up to 5 wt. % 235U, but was validated to 13 wt. %
- T/H - No impact to existing DNB margin (AXIOM cladding, ADOPT pellets, 6 wt. % 235U, Cr coated cladding)
- Core Physics
- Explicit modeling of Cr - cladding coating and ADOPT fuel pellets
- Negligible neutronic impact from AXIOM cladding
- No change to reload analysis methods, or the currently approved neutronic methods
- No impact to neutronic modeling for fuel rods above 5 wt.% 235U
- Few rods per fuel assembly neutron flux spectrum similar to currently operating core
- Core monitoring with BEACON Core Monitoring System is unaffected 10
Analysis
- Criticality Analysis
- Unit 1 SFP Storage prohibited
- Paragon was used for depletion calculations, while SCALE 6.2.3 was used for the rack criticality analysis.
- Storage analysis performed via direct reactivity analysis
- Storage not requiring burnup credit:
- New Fuel Storage Racks
- Demonstrated significant margin to storage limit including Dry, Fully Flooded, and Optimum moderation conditions
- Unit 2 SFP two-out-of-four storage pattern
- Demonstrated significant margin available
- Both storage analyses credit IFBA, which will be confirmed on a cycle specific basis using the reload process
- Storage requiring burnup credit:
- Unit 2 AOR all-cell storage pattern burnup limit is approximately 40 GWd/MTU
- Multiple full pool misload event also analyzed using Technical Specification soluble boron limit of 2000 ppm showed acceptable results 11
Questions/Discussion Acronyms and Terms
- ACRS: Advisory Committee on Reactor
- NRC: Nuclear Regulatory Commission Safeguards
- PCT: Peak Clad Temperature
- AOR: Analysis of Record
- PIE: Post-Irradiation Examination
- PWR: Pressurized Water Reactor
- BWR: Boiling Water Reactor
- SBLOCA: Small LOCA
- DNB: Departure from Nucleate Boiling
- SLB: Steam Line Break
- GWd: Gigawatt Days
- RWSC: Rod Withdrawal from
- IFBA: Integral Fuel Burnable Absorber Subcritical
- LAR: License Amendment Request
- SFP: Spent Fuel Pool
- LBLOCA: Large LOCA
- SNC: Southern Nuclear Company
- LOCA: Loss of Coolant Accident
- TS: Technical Specification
- LTA: Lead Test Assembly
- wt.%: weight percent
- MLO: Maximum Local Oxidation
- MTU: Metric Ton Uranium 13
BACKUP SLIDE - Example Loading Pattern LTA Assemblies S X X X S Fresh Fuel lowest enrichment S S S S Fresh Fuel mid-enrichment S X S X S Fresh Fuel highest enrichment (non-LTA)
S S Burned fuel X X X X X S S X Control Bank Location X S X X X X X S Shutdown Bank Location S S X X X X X S S S X S X S S S S S S X X X S 14
NRC Staff Presentation Accident Tolerant Fuel (ATF)
Lead Test Assemblies (LTAs)
Vogtle, Units 1 and 2 July 12, 2023
OPENING REMARKS Michael T. Markley, Branch Chief Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
LICENSING ACTIONS
- SNC requested:
- Amendments to License Condition 2.D and TSs: (1)
TS 3.7.18, Fuel Assembly Storage in the Fuel Storage Pool, (2) TS 4.2.1, Fuel Assemblies, and (3) TS 4.3, Fuel Storage.
- Three Exemptions
- 10 CFR 50.46 and 10 CFR part 50, Appendix K to allow the use of coated AXIOM cladding.
- 10 CFR 50.68(b)(7) to allow greater than 5 weight-percent U-235 3
UFSAR CHAPTER 15 ACCIDENT ANALYSIS
- Avg enrichment of the core is only increased 0.0003% by 4 LTAs resulting in negligible changes to source term or dose consequence for core wide accidents.
- Accidents with local core effects were re-evaluated using LTA data by the vendor and remained bounding with one exception
- Rod Ejection Accident remained limiting for LTAs and Tech Specs will reflect that LTA core locations will be appropriately limited during LTA utilization.
4
CODES AND METHODS
- Neutronics
- PARAGON and NEXUS
- Average Assembly enrichment is still within the range of applicability for PARAGON.
- Very little change in neutronic performance resulting from four enriched rods per assembly.
- Thermal Hydraulics
- VIPRE-W, DNB Correlations, Rod Bow Evaluation Methodology
- No impact to T-H performance or methodologies.
- AXIOM is expected to perform at least as well as Optimized ZIRLO.
- Control Rod Ejection
- The LTAs will not be placed in positions that have been shown to be limiting with respect to CRE.
5
FUEL ROD DESIGN
- AXIOM Cladding
- Zirconium alloy that is expected to demonstrate better in-reactor performance compared to Optimized ZIRLO.
- Chromium Coating
- Thin chromium coating for corrosion resistance with no impact to the thermal-hydraulic analyses.
- Increased Enrichment
- Four rods per assembly enriched up to 6 wt% U-235.
- Minimal impact to neutronic performance
- ADOPT Fuel Pellets
- Higher density pellets containing chromia and alumina 6
FUEL HANDLING & STORAGE
- Enrichment exemption: four rods per LTA.
- ADOPT increased TD actually adds more U235.
- Analysis
- LTAs compared to Vogtle New Fuel Storage Rack &
Spent Fuel Pool analyses of record.
- LTA LAR credits IFBA whereas AORs do not.
- LTA LAR did not credit IFBA for SFP accident. NRC review did.
- Reasonable assurance 10 CFR 50.68(b)(2),
(b)(3), and (b)(4) are met.
7
CONCLUSION
- The NRC staff determined that there is reasonable assurance that the health and safety of the public will not be endangered by allowing SNC the use of four ATF LTAs for up to two cycles operation in Vogtle, Unit 2.
- The NRC staff looks forward to the ACRS Letter Report.
8
- LTAs - Lead Test Assemblies
- NRC - U.S. Nuclear Regulatory Commission
- SNC - Southern Nuclear Operating Company
- NRR - Office of Nuclear Reactor Regulation
- DSS - Division of Safety Systems
- DORL - Division of Operating Reactor Licensing
- SNSB - Nuclear Systems Performance Branch
- SFNB - Nuclear Methods & Fuel Analysis Branch 9
ACRONYMS - Continued
- TS - Technical Specification
- CFR - Code of Federal Regulations
- UO2 - Uranium Dioxide
- MWd - Megawatt-Day
- ZrB2 - Zirconium Diboride
- IFBA - Integral Fuel Burnable Absorber
- LOCA - Loss-of-Coolant Accident
- EM - Evaluation Model
- DNBR - Departure from Nucleate Boiling Ratio 10
ACRONYMS - Continued
- PCT - Peak Cladding Temperature
- REA - Rod Ejection Accident
- TD - Theoretical Density
- NFSR - New Fuel Storage Rack
- SFP - Spent Fuel Pool
- AOR - Analysis of Record
- 2oo4 - Two out of Four Configuration
- 4oo4 - All-Cell Configuration
- B10 - Boron 10
- ppm - Parts Per Million 11
EPRI Report 3002018337, "Use of Data Validation and Reconciliation Methods for Measurement Uncertainty Recapture:
Topical Report (open session)
ACRS Full Committee Meeting Lois M. James, Senior Project Manager Office of Nuclear Reactor Regulation July 12, 2023
Agenda
- Project History
- Purpose of the Technical Report
- Data Validation and Reconciliation (DVR)
- Data Validation and Reconciliation for CTP -
MUR Relevance
- Review Scope Limitations
- Safety Evaluation Conclusion Slide # 2
Project History Date Activity ADAMS Accession No.
1/27/2021 Electric Power Research Institute (EPRI) Transmitted ML21053A027 "Use of Data Validation [DVR] and Reconciliation Methods for Measurement Uncertainty Recapture
[MUR]: Topical Report for U.S. Nuclear Regulatory Commission (NRC) review and approval 3/16/2021 NRC issued its completeness and withholding ML21110A049 Determination for EPRI DVR for MUR Topical Report 5/2/2022 NRC issued request for additional information ML22118A052 regarding regulatory review scoping issues (RAIs) 8/8/2022 EPRI submitted responses to Phase A RAIs ML22223A052 Slide # 3
Project History (cont.)
Date Activity ADAMS Accession No.
12/7/2022 NRC issued follow-up technical (Phase B) RAIs ML22341A075 3/9/2023 EPRI submitted Phase B RAI responses ML23066A242 4/13/2023 EPRI submitted supplemental information to ML23103A150 the RAI 14 Response 5/6/2023 NRC issued the draft safety evaluation for EPRI ML23088A184 proprietary review Slide # 4
Purpose of Technical Report
- EPRI Report 3002018337 describes a process for using a mathematical data validation and reconciliation (DVR) method for monitoring core thermal power and use of the methods for measurement uncertainty recapture uprates.
Slide # 5
Data Validation and Reconciliation (DVR)
- What it is
- A statistical analysis of multiple plant measurements, in aggregate, to provide accurate core thermal power
- What it can do
- Reduce uncertainties associated with the core thermal power and allow plants to operate closer to the approved core thermal power
- Reduce single failure vulnerabilities by using more instrumentation to determine core thermal power
- Improve condition monitoring and condition-based maintenance by using the data points collected by plant equipment and defined physical relationships between measurements monitor equipment performance
- History
- Used by the US and European nuclear power industry since 1999 to assess turbine cycle thermal performance, balance of plant feedwater flow metering, and accuracy of the plant calorimetric
- Used DVR to increase power output in Europe Slide # 6
Data Validation and Reconciliation for CTP - MUR Relevance Current Practice of Determining Core Thermal Power (CTP)
- Inaccuracies in the direct measurement of feedwater flow have resulted in lost generation and potential overpower conditions
- In the past, ultrasonic flow measurement devices (UFM) have been used to gain measurement accuracy, but as technology improves, more can be done Data Validation and Reconciliation (DVR)
- DVR can be used to reduce the uncertainties in the calculation of CTP and thus allow a licensee to produce more power via a measurement uncertainty uprate Slide # 7
Review Scope Limitations
- The safety evaluation
- Is not a review of any software, logic flow, or numerical methods implemented in any particular vendors packaging of the DVR methodology.
- Is a review of the concepts of DVR and the steps needed to model the plant for the purpose of estimating feedwater flow and core thermal power mean and uncertainty.
- Specific evaluation of software, logic flow, or numerical methods implemented would be performed via a license amendment request or application.
Slide # 8
Safety Evaluation Conclusion Based on the
- NRC staffs risk assessment of the DVR results
- NRC staffs previous treatment of similar models and simulations
- NRC staffs previous evaluation of nuclear power plant process measurement uncertainty
- NRC staffs understanding of the DVR methodology
- NRC staffs previous treatment of the calculation of the feedwater flow rate and its uncertainty NRC staff concludes that there is reasonable assurance that the DVR method as described in EPRI TR 3002018337 can be used to determine the core thermal power and the core thermal power uncertainty, provided all DVR conditions and limitations have been satisfied.
Slide # 9
ARITA ARTEMIS/RELAP Integrated Transient Analysis Methodology Buck Barner ACRS Full Committee, July 12th, 2023
Content
- 1. Overview
- 2. Background and History
- 3. Approval Request and Range of Applicability
- 4. Evaluation Model Description
- 5. Summary ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology 2
ACRS Full Committee , July 12th, 2023 2
Overview ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology Defines a methodology to analyze non-Loss-of-Coolant (non-LOCA) events Uses a non-parametric statistical approach to make a 95/95 statistical statement for each figure of merit (FOM) using a Monte Carlo approach Standard Review Plan (SRP) Chapter 15.0.2 was used as guidance in development of the method Addresses additional topics, such as mixed core power distribution control setpoints and fuel assembly reconstitution Excludes Control Rod Ejection (CRE) which is analyzed using AREA - ARCADIA Rod Ejection Accident Topical Report ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 3
Background and History ARITA Advanced C&M Development ARITA Development ARITA Licensing Future In 2006, Framatome began the development of a new set of advanced PWR codes ARCADIA (ANP-10297PA, Revision 0 and Supplement 1, Revision 1)
Includes the 2D cross section code APOLLO2-A and the 3D nodal code ARTEMIS COBRA-FLX (ANP-10311PA, Revision 1)
GALILEO (ANP-10323PA, Revision 1)
Around the same time (2010) there was a push in the industry to replace legacy methods The goal was to develop new methodologies that:
Use state-of-the-art modeling Implement best practices from decades of US, French and German industry experience Simplify topical report interdependences to reduce licensing complexity Facilitate future method development and innovation AREA (ANP-10338PA, Revision 1) was the first methodology topical approved.
ARITA represents the realization of Framatomes objective of bringing innovation and improved performance to the industry through advanced methods.
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 4
Background and History ARITA Advanced C&M Development ARITA Development ARITA Licensing Future Pre-submittal meetings held February 2015, June 2016, and July 2017 The ARITA Topical Report was submitted August 2018 The first set of RAIs (1-13) were transmitted to Framatome December 2019 Responses to RAIs 1-13 were transmitted to the NRC April 2020 All RAIs transmitted to Framatome April 2020 Responses to RAIs 14-92 were transmitted to the NRC June 2021 Audits and discussions between NRC and Framatome continued through April 2022 Final updated responses to all RAIs to address reviewer comments were transmitted to the NRC June 2022 The Draft SER April 2023 ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 5
Background and History ARITA Advanced C&M Development ARITA Development ARITA Licensing Future What advantage does this provide to the industry?
Enhanced modeling of the actual plant behavior leads to better understanding of plant response and the actual safety margins.
Increased understanding allows us to focus on the areas that are most important to safety and demonstrate compliance with all safety regulations and requirements Bringing value and new opportunities to the industry Address Regulatory Changes (e.g., RG 1.236)
Core Design Optimization Advanced Fuel Management (AFM) - Increased Enrichment and High Burnup Higher fidelity simulation and modeling provide increased understanding and confidence in plant safety ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 6
Approval Request and Range of Applicability Non-LOCA Chapter 15 methodology, excluding CRE Mixed Core Method Local Power Density Limiting Condition of Operation (LPD LCO) and Core Safety Limit Lines (CSLL)
Power Distribution Control (PDC)
Fuel Assembly Reconstitution Applicable to Westinghouse (2-, 3-, and 4-loop) Pressurized Water Reactor (PWR) designs and Combustion Engineering (CE) PWR designs Use of approved Critical Heat Flux (CHF) correlations Within the range of applicability of the constituent codes (ARTEMIS, S-RELAP5, COBRA-FLX, GALILEO)
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 7
Evaluation Model Description Constituent codes ARTEMIS - 3D nodal simulator code previously approved in ANP-10297 (2013 & 2018)
COBRA-FLX - Subchannel core thermal-hydraulics code previously approved in ANP-10311 (2010)
GALILEO - Fuel performance code previously approved in ANP-10323 (2020)
S-RELAP5 - System thermal-hydraulics code previously applied in EMF-2310 (2004 & 2011)
Evaluation Model (EM) Variants There are 3 EMs described in the ARITA topical:
- 1) Coupled system-thermal hydraulic and neutronics model,
- 2) 0D system thermal-hydraulic model, and
- 3) Static core evaluation model.
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 8
Evaluation Model Description Code Coupling In the Coupled EM, ARTEMIS and S-RELAP5 are coupled together to solve time-dependent multi-physics problems (Specified Acceptable Fuel Design Limits (SAFDLs) and non-SAFDL FOM)
In the 0D EM, point kinetics data generated in ARTEMIS is provided to S-RELAP5 (Non-SAFDL)
In the Static EM, ARTEMIS is used for events that do not require a system thermal-hydraulic solution (SAFDL)
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 9
Evaluation Model Description Statistical Approach Monte Carlo approach better manages the complexity of the coupled Non-LOCA transient analysis Non-parametric approach based on the Wilks method is used to make a statistical statement on the FOM.
Account for multiple FOM.
Some parameters may still be biased No necessity to develop a full distribution of results The statistical approach is used for all 3 EM variants described above.
Similar non-parametric approaches have been presented before ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 10
Summary - ARITA Is a 3D coupled statistical Non-LOCA method applicable to CE and Westinghouse plants Represents a major milestone in Framatomes commitment to state-of-the-art modeling and decades of industry experience.
Provides the foundation for future development and innovation Ensures confidence in plant safety and compliance with all regulations and requirements through high fidelity modeling ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 11
Acronyms AFM - Advanced Fuel Management LOCA - Loss of Coolant Accident AREA - ARCADIA Rod Ejection Accident LPD LCO - Local Power Density Limiting Condition of Operation ARITA - ARTEMIS/RELAP Integrated Transient Analysis Non-LOCA - non-Loss of Coolant Accident CE - Combustion Engineering NRC - U.S. Nuclear Regulatory Commission CHF - Critical Heat Flux PDC - Power Distribution Control CRE - Control Rod Ejection PWR - Pressurized Water Reactor CSLL - Core Safety Limit Lines SAFDL - Specified Acceptable Fuel Design Limits EM - Evaluation Model SRP - Standard Review Plan FOM - Figure of Merit ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 12
Thank you ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee , July 12th, 2023 13
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APOLLO-2A, ARCADIA, AREA, ARITA, ARTEMIS, COBRA-FLX, GAIA, GALILEO, M5FRAMATOME and S-RELAP5 are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries.
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NRC Staffs Review of Framatome Topical Report ANP10339P, ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology Open Presentation to the Advisory Committee on Reactor Safeguards July 12, 2023 K. Heller, U.S. NRC J. Lehning, U.S. NRC K. Geelhood, D. Richmond, T. Zipperer, B. Schmitt, D. Engel, PNNL
Introduction
- The ARITA methodology is a statistical approach for performing most Standard Review Plan (SRP)
Chapter 15 reactor safety analyses
- Not including LOCA and rod ejection
- Applicable to conventional Westinghouse and Combustion Engineering PWRs
- ARITA involves three distinct evaluation model variants
- The codes used in the ARITA methodology have been previously reviewed by the NRC staff
- NRC staffs review focused mainly on the calculational procedure and uncertainty treatments 2
Key Regulatory Requirements and Guidance
- 10 CFR 50, Appendix A, General Design Criteria, e.g.,
- GDC 10, Reactor Design
- GDC 15, Reactor Coolant System Design
- 10 CFR 50.36, Technical Specifications
- 10 CFR 50.67 or 10 CFR Part 100 Dose Limits
- Standard Review Plan, Chapter 15
- Regulatory Guide 1.203, Evaluation Model Development and Assessment Process 3
Acceptance Review History Review Supplement Mar 2019 FRAMATOME Updated Topical ANP10339P Response Response Response Response Report / Final Submitted to 13 RAIs to 37 RAIs to 16 RAIs to 26 RAIs RAI Response Aug 2018 Mar 2020 Jul 2020 Dec 2020 May 2021 Jun 2022 2018 2019 2020 2021 2022 2023 Accptnce Detailed RAI Resolution Draft SE Supplmnt Review Acceptance Batch 1 Prospective 19 Audits & Meetings Draft SE ACRS SC Review RAIs (13) L&Cs First to Resolve RAIs Apr 2023 Jun 2023 Complete Issued Discussed Nov 2020 - April 2022 Nov 2018 Dec 2019 Dec 2020 Onsite Batch 2 Audit RAIs (79)
Sept 2019 Issued NRC Apr 2020 4
Calculational Procedure
- Key review issue: Assessing ARITA methodology against regulatory guidance concerning initial conditions, for example SRP 15.0.I.6.C.ii:
- The reviewer verifies that initial conditions used in the analyses are suitably conservative
- What is objective with respect to ARITAs intended statistical statement and how to achieve it?
- Technical Specifications and safety analyses need to align
- Staffs concern: assubmitted method may not provide conservative results for plant operation throughout the allowed operating domain
- Establish a conservative event definition
- Achieve 95/95 for postulated event over the allowed operating domain
- Conservative sampling of highly influential parameters
- Compromise between traditional and bestestimate sampling approaches 5
Treatment of Uncertainty
- The ARITA methodology uses a statistical sampling approach for uncertainty quantification
- Wilks method: nonparametric estimate of 95/95
- Challenge lies in developing sufficient data to support uncertainty distributions
- NRC staffs review focused on treatment of:
- Uncertainty in input parameters
- As modified, treatments found acceptable
- Uncertainty in calculated figures of merit
- Univariate and multivariate approaches acceptable
- Overall statistical calculation procedure
- Rigorous determination of tolerance limits
- Appropriately avoids biasing that could degrade tolerance limits 6
Validation Assessment
- Constituent Codes
- SRELAP5, ARTEMIS, COBRAFLX, GALILEO
- Most validations cited past reviews
- Conservatisms and limitations and conditions account for cases/areas of limited validation
- Evaluation Model Variants
- 0D and Static
- Very similar to prior standalone approvals
- Coupled
- Assessed information exchange and the provided integral assessments
- Supplementary Evaluation Model Features
- Acceptable with appropriate conservatisms in place
- Largely evolutionary updates 7
Limitations and Conditions
- A total of 28 Limitations and Conditions
- Approximate breakdown:
- Statistical Process and/or 16 Uncertainty Treatment Methodology Application 10 Conservatism Assurance 2
- Framatome intends to address L&Cs 18 and 19 in a future review 8
Conclusions
- The NRC staff found the ARITA methodology acceptable for modeling inscope SRP Chapter 15 events, including
- all three ARITA evaluation model variants
- the associated calculational process
- the statistical uncertainty methodology
- The NRC staff found the supplementary evaluation model features described in Section 3.8 of its safety evaluation acceptable
- The staffs conclusions are predicated upon
- the ARITA methodology being used within its proposed range of applicability in Section 13.0 of ANP10339P
- licensees acceptably addressing limitations and conditions in Section 5.2 of the staffs safety evaluation 9
Limitations and Conditions 18 and 19
- Limitation and Condition 18
- Licensees shall justify Framatomes proposed approach for treating the uncertainty in the axial peaking factor (FZ), conservatively estimate it, or justify another approach
- Impetus: Staff could not conclude Framatomes proposed approach directly follows from the information presented during the review
- Limitation and Condition 19
- For the stated event, licensees shall treat the uncertainty for the axial peaking factor (FZ) for the postscram main steam line break event via the specified conservative approach or justify an alternative
- Impetus: NRC staff concluded Framatome did not adequately justify the proposed uncertainty treatment would wellcharacterize the event conditions
- Introduction of L&Cs 18 and 19 reflects only where we ended the review, based on docketed material 10
Future Framatome Plans to Address Limitations and Conditions 18 and 19
- NRC staff and Framatome have discussed potential pathways to address Limitations and Conditions 18 and 19 in a future regulatory review
- Additional technical justification may reduce or eliminate impact of these limitation and conditions
- Additional technical dialogue would facilitate timely resolution
- NRC staff will complete current review of ANP10339P and issue final SE as scheduled
- Framatome to propose pathway to address technical issues in Limitations and Conditions 18 and 19 in a separate review process
- e.g., expedited review of update/revision to the topical report, inclusion in a plant license amendment request, etc.
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