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
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 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 707TH MEETING 4
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5
(ACRS) 6
+ + + + +
7 OPEN SESSION 8
+ + + + +
9 WEDNESDAY 10 JULY 12, 2023 11
+ + + + +
12 13 The Advisory Committee met via hybrid In-14 Person and Video-Teleconference, at 8:30 a.m. EDT, Joy 15 L. Rempe, Chairman, presiding.
16 17 COMMITTEE MEMBERS:
18 JOY L. REMPE, Chairman 19 WALTER L. KIRCHNER, Vice Chairman 20 DAVID A. PETTI, Member-at-Large 21 RONALD G. BALLINGER, Member 22 CHARLES H. BROWN, JR., Member 23 VICKI M. BIER, Member 24 VESNA B. DIMITRIJEVIC, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 GREGORY H. HALNON, Member 1
JOSE MARCH-LEUBA, Member 2
ROBERT P. MARTIN, Member 3
THOMAS E. ROBERTS, Member 4
MATTHEW W. SUNSERI, Member 5
6 ACRS CONSULTANT:
7 DENNIS BLEY 8
10 DESIGNATED FEDERAL OFFICIAL:
11 KENT HOWARD 12 13 ALSO PRESENT:
14 BUCK BARNER, Framatome 15 KURT CRYTZER, EPRI 16 LOIS M. JAMES, NRR 17 KENNETH GEELHOOD, NRR 18 KEVIN HELLER, NRR 19 SCOTT KREPEL, NRR 20 RYAN JOYCE, SNC 21 JOHN LAMB, NRR 22 JOHN LEHNING, NRR 23 MICHAEL MARKLEY, NRR 24 ALAN McGINNIS, Framatome 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
3 CHARLEY PEABODY, NRR 1
RADU POMIRLEANU, Westinghouse 2
DAVID RAHN, NRR 3
JIM SMITH, NMSS 4
GREGORY SUBER, NRR 5
MIKE WERNER, Westinghouse 6
BRANDON WISE, NRR 7
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 CONTENTS 1
PAGE 2
I.
Opening Remarks by the ACRS Chairman 3
- 1. Opening Statement 5
4
- 2. Agenda and Items of Current Interest 6
5 6
II.
EPRI Data Validation Topical Report 7
- 1. Remarks from Subcommittee Chair 9
8
- 2. Presentations and Discussions with EPRI 9
Representatives and NRC Staff 12 10 11 III.
Vogtle License Amendment Request (LAR) on 12 Loading Lead Test Assemblies (LTAs) with 13 Increased Enrichment 14
- 1. Remarks from Subcommittee Chair 32 15
- 2. Presentations and Discussions with 16 Southern Nuclear Representatives and 17 NRC Staff 34 18 19 IV.
ARITA-ARTEMIS/RELAP Integrated Transient 20 Analysis Methodology Topical Report 21
- 1. Remarks from Subcommittee Chair 62 22
- 2. Presentations and Discussions with 23 Framatome Representatives and NRC 24 Staff 65 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
5 P-R-O-C-E-E-D-I-N-G-S 1
8:30 a.m.
2 CHAIR REMPE: Good morning. It's 8:30 3
here on the East Coast. This meeting will now come to 4
order. This is the first day of the 707th Meeting of 5
the Advisory Committee on Reactor Safeguards.
6 I'm Joy Rempe, Chairman of the ACRS.
7 Other members in attendance are Ron Ballinger, Vicki 8
Bier, Charles Brown, Vesna Dimitrijevic, Greg Halnon, 9
Walt Kirchner, Jose March-Leuba, Bob Martin, Dave 10 Petti, Thomas Roberts, and Matthew Sunseri.
11 We do have a quorum. Today the committee 12 is meeting in-person and virtually. The ACRS was 13 established by the Atomic Energy Act and is governed 14 by the Federal Advisory Committee Act.
15 The ACRS section of the US NRC public 16 website provides information about the history of this 17 committee and documents such as our charter, bylaws, 18 Federal Register notices for meetings, letter reports, 19 and transcripts of all full and subcommittee meetings, 20 including all slides presented at the meetings. The 21 committee provides its advice on safety matters to the 22 Commission through its publicly available letter 23 reports.
24 The Federal Register notice announcing 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
6 this meeting was published on June 21, 2023. This 1
announcement provided a meeting agenda, as well as 2
instructions for interested parties to submit written 3
documents or request opportunities to address the 4
committee. The Designated Federal Officer for today's 5
meeting is Mr. Kent Howard.
6 The communications channel has been opened 7
to allow members of the public to monitor the open 8
portions of the meeting. The ACRS is inviting members 9
of the public to use the MS Teams link to view slides 10 and other discussion materials during these open 11 sessions.
12 The MS Teams link information was placed 13 in the Federal Register notice and agenda on the ACRS 14 public website. We have received no written comments 15 or requests to make oral statements from members of 16 the public regarding today's session.
17 Periodically, the meeting will be open to 18 accept comments from participants listening to our 19 meetings. Written comments may be forwarded to Mr.
20 Kent Howard, today's Designated Federal Officer.
21 During today's meeting, the committee will 22 consider the following topics: EPRI Data Validation 23 Topical Report, Vogtle License Amendment Request on 24 Loading Lead Test Assemblies with Increased 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
7 Enrichment, and ARITA-ARTEMIS/RELAP Integrated 1
Transient Analysis Methodology Topical Report.
2 Note the portions of the EPRI and ARITA-3 ARTEMIS sessions may be closed as stated in the 4
agenda. I also want to note that the topic of 5
LANCR02, Lattice Physics Model Description Licensing 6
Topical Report, that was scheduled for tomorrow's 7
session will be discussed during our Friday planning 8
and procedures meeting.
9 A transcript of the open portion of 10 today's meeting is being kept. It's requested that 11 speakers identify themselves and speak with sufficient 12 clarity and volume so they can be readily heard.
13 Additionally, participants should mute themselves when 14 not speaking.
15 This morning I do have an item of note.
16 I want to recognize our two newest members, Bob Martin 17 and Tom Roberts, who are joining us for their first 18 full committee meeting open session.
19 Bob is a career nuclear safety specialist 20 through employment with BWXT, Babcock & Wilcox, 21 Framatome, and Siemens Power Corporation. Dr. Martin 22 was responsible for the development and regulatory 23 defense of several evaluation methodologies for the 24 design and safety of conventional and advanced nuclear 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
8 power plants.
1 Among these are the first applications of 2
best-estimate plus uncertainty methods to design-basis 3
LOCA for both fuel and containment evaluations and for 4
demonstrating plant resiliency to beyond-design-basis 5
severe accidents.
6 His experience includes employment at the 7
Idaho National Laboratory, where he worked in the area 8
of thermal system fluid modeling, and being the lead 9
editor and contributor on the testing of design-basis 10 accident analysis methods for light-water nuclear 11 power plants.
12 Tom Roberts has more than 40 years' 13 experience in the field of nuclear reactor systems and 14 safety. He spent 36 years as an engineer and 15 engineering manager at the Naval Nuclear Propulsion 16 Program Headquarters, working various roles in the 17 Instrumentation and Controls Division, and then 18 completing his career with 12 years as the Director of 19 Reactor Safety and Analysis.
20 Since his retirement from the Naval 21 Nuclear Power Program, Mr. Roberts has served as a 22 subject matter expert in programs for advanced reactor 23 development, including consulting and reactor safety, 24 and IMC for a transportable micro-reactor program and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 for a nuclear thermal rocket reactor concept.
1 I do hope that members and others will 2
join me in welcoming Rob and Tom, and in thanking 3
Alicia, Andrea, and Sandra for helping them through 4
the onboarding process.
5 At this time, I want to ask other members 6
if they have any opening remarks. Not hearing any, 7
I'd like to ask Member March-Leuba to lead us through 8
our first topic for today's meeting.
9 Jose?
10 MEMBER MARCH-LEUBA: Our first topic is 11 the EPRI Data Validation Topical Report, which we 12 listened to during the thermal-hydraulics subcommittee 13 on June 7th. We are going to have only an open 14 session. If required, we can always create a closed 15 link and discuss.
16 So please try not to mention any numbers, 17 which does tend to be proprietary. Other than that, 18 keep it general. If necessary we can close the 19 session, but I hope we don't.
20 I believe the staff is going to present.
21 Greg or Scott, are you going to present? Greg is 22 going to. Go for it.
23 MR. SUBER: Good morning and thank you.
24 I'd like to welcome the two new members of the ACRS.
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
10 I'm glad that you got here so you can understand the 1
term hotter than July, as you will experience over the 2
next three days.
3 Once again, I'd like to thank the 4
committee for giving us the opportunity to discuss the 5
EPRI Technical Report for the use of data validation 6
and reconciliation methods for measurement uncertainty 7
recapture.
8 This topical report is a first-of-a-kind 9
topical report. It proves that it has the potential 10 to increase power for the US Nuclear Reactor fleet.
11 It's also a first of a kind in that it is an approval 12 of concept that is not typically the subject of a 13 topical report.
14 Therefore, the staff has conducted a very 15 thorough and in-depth technical review using 16 contractors and support from Sandia Labs, which 17 enhance the NRC Staff's capabilities.
18 EPRI, its members, and consultations do 19 their part in submitting a detailed topical report and 20 responded to a request for additional information in 21 a timely manner.
22 As a result of everybody's dedication and 23 efforts, the staff's safety evaluation for the topic 24 was issued on May the 6th. The staff concludes that 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
11 reasonable assurance has been provided for the DBR 1
method.
2 It can be used to determine, one, core 3
thermal
- power, and also core thermal power 4
uncertainty. And thus, this report can be used by 5
licensees for requests for measurement uncertainty 6
power operations.
7 The NRC staff and EPRI are ready to 8
discuss the safety evaluation in detail in this 9
meeting. With that, I turn it back to the Chair.
10 MEMBER MARCH-LEUBA: Thanks, Greg. Lois 11 James is going to be our presenter; is that correct?
12 MR. SUBER: Yes, she is.
13 MEMBER MARCH-LEUBA: So Lois, we're going 14 to give you the microphone in just a moment. I'll 15 just give you a heads up that some members were not 16 present during the subcommittee and they have 17 questions about a topic that Josh Kaizer presented 18 very eloquently to the committee, which was the way he 19 introduced the risk-informed methodology to his 20 review.
21 Either you address it when you feel it's 22 proper or we will ask you the question at the end of 23 the presentation. I'm just giving you a heads up. So 24 Lois, go ahead.
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
12 MS. JAMES: Thank you, Chair Rempe, Member 1
March-Leuba, and other members of the ACRS. As 2
stated, we are here today to discuss the staff's 3
review of EPRI, Electric Power Research Institute's 4
report on the use of data validation and 5
reconciliation for DBR methods for measurement 6
uncertainty recapture or MUR.
7 The staff has acknowledged that this is a 8
first-of-a-kind report and has the potential to have 9
a large impact on the nuclear industry.
10 As mentioned, my name is Lois James. I am 11 the project manager for the staff's review. In the 12 room or on the phone we have Scott Krepel, Nuclear 13 Systems Performance Branch Chief, and David Rahn, one 14 of the technical reviewers. We also have several 15 members of EPRI and their supporting staff to answer 16 any questions that the members may have.
17 The agenda for today, we will be providing 18 a short history of the staff's review. We'll mention 19 the purpose of the technical report. We'll discuss 20 what DBR is. We'll also talk about how DBR impacts or 21 is used for core thermal power. We will mention the 22 review scope limitations and then we will discuss the 23 staff's evaluation conclusions.
24 I'm not going to mention every item on the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 project history. As you can see from the two slides, 1
we have provided the ML numbers. We thought this was 2
important to provide key dates and ADAMS reference 3
numbers for everyone, for members of the public and 4
for the committee.
5 You can see that the staff's review 6
started in early 2001. We brought Sandia Labs on to 7
support the staff's review during 2001. During 2002 8
and 2003, the staff issued the RAIs and conducted its 9
review.
10 In May of 2003, we issued our draft safety 11 evaluation for proprietary review. We shared that 12 with the committee. That's all the project history 13 that we were going to mention today.
14 MEMBER MARCH-LEUBA: Lois, this is Jose.
15 Can I ask you a question now?
16 MS. JAMES: Sure.
17 MEMBER MARCH-LEUBA: Since this is the 18 full committee, in my opinion, this is more for the 19 benefit of the public than for our questions. We 20 drilled you guys during subcommittee.
21 I wanted to ask you, I see that the 22 project history is two and a half years long from 23 issuance of the topical report and issuance of the 24 SER. Can you explain, was there any kind of problems 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
14 that need to be resolved with the RAIs? Or it's 1
simply lack of resources on the part of the staff?
2 Because two and half years looks like a long time for 3
a topical report.
4 MS. JAMES: Yes, sir. When we received 5
the project initially, we decided that it was such a 6
unique project and dealt very strongly on statistical 7
methods that we chose to bring in and contract Sandia 8
to support the staff.
9 That took longer than anticipated. So a 10 good bit of 2021 was actually procuring the contract 11 from the start. That's not an easy task to go from 12 start to finish. That takes at least six to nine 13 months to get a contract in place.
14 And then when we finally started the 15 review, EPRI and its contractor support provided a lot 16 of information in the topical report. The first set 17 of RAIs we issued was to determine what exactly they 18 wanted us to review and approve because they had 19 provided a lot of examples. And they had provided 20 some code information and some method information.
21 So the first set of RAIs that we issued in 22 2002 after we got the contract in place was first to 23 determine really what we were supposed to review and 24 what we were supposed to document in the SE. And so 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 the actual second round of RAIs, phase B, those were 1
kind of the technical questions that we had. So 2
that's why the review took a little longer than 3
anticipated.
4 MEMBER MARCH-LEUBA: Thank you. I want 5
you to understand that, first, anything you hear here 6
until we write the letter are individual members' 7
opinions. We always say that, but we need to keep on 8
saying it.
9 If I summarize what I heard you say, the 10 reason it took two and a half years for review was 11 resource location and the lack of proper 12 communications at the beginning of the project, 13 misunderstanding of the scope.
14 In my opinion, and it's my opinion, I 15 found when I was doing the work on your side of the 16 table, audits helped with the misunderstandings very 17 much. You have an audit with the applicant and you 18 determine the scope real well.
19 I hope management keeps getting lessons 20 learned and we speed up this process because it's in 21 the benefit of everybody. Thank you. Keep going. It 22 was just an observation.
23 MS. JAMES: Understood. The staff has 24 developed internal lessons learned. I try to do that 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
16 in every one of my projects so that we can go forward, 1
and do things more efficiently and effectively in the 2
future.
3 MEMBER MARCH-LEUBA: Yes. I'm not wasting 4
time. This is the beginning of the meeting so we 5
still have to -- you always have a pre-submittal 6
meeting with the applicant when you get a flavor of 7
what's going to be submitted. But once the technical 8
responsible staff get assigned to the project, in my 9
opinion, having an audit speeds up everything. It 10 really does. Okay.
11 MS. JAMES: What is data validation and 12 reconciliation? It's a statistical analysis of 13 multiple plant measurements, an aggregate to provide 14 an accurate core thermal power.
15 Of interest to us in this topical report 16 was that DBR can reduce uncertainties associated with 17 the core thermal power and allow plants to operate 18 closer to the approved thermal power without reducing 19 the safety margin.
20 Using the data validation and 21 reconciliation can also help plants reduce single-22 failure vulnerabilities, using more instrumentation to 23 determine core thermal power. It can also improve 24 condition monitoring and condition-based maintenance 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
17 by using the data points collected by the plant 1
equipment. This can be used by the system engineers 2
to better monitor their plants.
3 We would like to acknowledge that DBR has 4
been used in US and European nuclear power industries 5
since 1999. It has been used to assess the target 6
cycle thermal performance, balance of plant feedwater 7
flow metering, and accuracy of the plant metrics.
8 It is also being used, it is my 9
understanding, right now in Europe to increase power 10 output. So that's where the US market is headed.
11 The purpose of the technical report was to 12 describe the process for using the mathematical data 13 validation and reconciliation for specifically 14 monitoring core thermal power. And then use this DBR 15 to proceed into the measurement uncertainty power 16 upgrade or measurement of certain recapture power 17 upgrades.
18 Anything impacting core thermal power has 19 potential to impact safety. So the staff conducted an 20 in-depth review that was discussed in detail with the 21 ACRS subcommittee during the closed portion of the 22 meeting in June.
23 Currently, core thermal nuclear power is 24 calculated based on feedwater flow measurements.
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
18 There are inaccuracies in the direct measurement of 1
feedwater flow that have resulted in lost generation 2
and potential overpower conditions. In the past 3
ultrasonic flow measurement devices, UFMs, have been 4
used to gain measurement accuracy. But as technology 5
improves, more can be done.
6 Data validation and reconciliation, as I 7
stated, uses statistical analysis of multiple plant 8
measurements to provide an accurate reading of core 9
thermal power. The DBR can then be used to reduce 10 uncertainties in the core thermal power and thus allow 11 licensees to produce more power via a measurement 12 uncertainty power upgrade.
13 So we expect after approval of this 14 topical report that we will begin to see the 15 industry's interest in measurement uncertainty 16 recapture upgrades.
17 We wanted to mention that this review, the 18 safety evaluation was not a review of any software, 19 logic flow, or numerical method implemented by any 20 particular vendor or licensee. It is a review of the 21 concepts of DBR and steps needed to provide the model 22 used by the plants to estimate feedwater flow and core 23 thermal power need and uncertainty.
24 Specific evaluations of software, logic 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
19 flow, and numerical methods would be performed via a 1
license amendment request or an application for a 2
power upgrade.
3 MEMBER MARCH-LEUBA: Lois, this is Jose 4
again. As I told you earlier, this would be a very 5
good point at which you can introduce the risk-6 informed methodology or approach that the staff used 7
to review these power upgrades. Could you give us a 8
hint about how we did it, a summary, a high-level 9
summary?
10 MS. JAMES: Yes.
11 MR. RAHN: If you'd like, Lois -- this is 12 David Rahn -- I could help with that.
13 MS. JAMES: That would be great. Thank 14 you.
15 MR. RAHN: Okay. Thank you.
16 Chairman Rempe and Member March-Leuba, the 17 staff was concerned initially that there was quite a 18 bit of new mathematical and statistics that were used 19 in the concepts that were provided in the DBR 20 methodology.
21 The process of coming up with a reconciled 22 mean and a reconciled variance associated with the 23 measurements from all the different parameters that 24 are going to be used to help reduce the overall 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
20 uncertainty for feedwater flow measurement was 1
complicated.
2 So we wanted to look at it to say, okay, 3
even if we're way off base and we're not understanding 4
it properly, how far off could it possibly be? How 5
bad could we be to come up with a computation of core 6
thermal power?
7 To do that we used our regular risk-8 informed processes, which is to look at what we call 9
the risk triplet. What can go wrong, how likely is 10 it, and what are the consequences? In addition, we 11 looked to see if any of those questions we answered 12 incorrectly, what's the residual risk in that?
13 So essentially, we parsed the uncertainty 14 measurements into what could go wrong if we made a 15 mistake in determining the reconciled mean. And also, 16 what could go wrong if we determined the reconciled 17 variance with an incorrect measurement.
18 By looking at that, we found that there's 19 a history of operating the plant already. And we know 20 what the ballpark should have been for those 21 measurements. So if we're off, we look to see how far 22 off could they really be.
23 In our analysis we determined that even if 24 we were off two percent, which is what was our initial 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
21 look at it, it looked to say that it's not going to be 1
an adverse problem on the operation of the core. We 2
already have an allowance for that two percent.
3 But let's say we were even to double that.
4 We looked to see what the possible implications are of 5
four percent, for example, being off. And we looked 6
to see what conservatisms we already have in either 7
transient analysis or steady-state operations, 8
operating with a starting point of four percent.
9 We found that the conservatisms that we 10 already have built into our analysis are fairly 11 substantial, and that we could live with the risk of 12 as much as four percent uncertainty. So essentially, 13 the risk analysis that was performed was a what-if 14 type scenario. We tried to put bounds on it.
15 The only thing we could not do is 16 determine the likelihood. We had no basis for 17 establishing what would be the likelihood. So we 18 focused on what would be the worst-case consequences.
19 MEMBER ROBERTS: David, this is Tom 20 Roberts. I have a couple of questions on what you 21 just presented.
22 One, you said that the analysis method is 23 complicated. I guess my experience is most analysis 24 methods are complicated. And you end up having to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 carefully go through all of the assumptions and make 1
sure that the methods you're using are appropriate for 2
what it is you're modeling.
3 I guess I'm not seeing how this DBR is any 4
different than any other analysis in terms of the 5
residual uncertainty just because you choose to use 6
that method compared to some other method.
7 If you used square-root accommodation of 8
uncertainties for a typical analysis, for example, you 9
would still have to note that those individual terms 10 are all
- random, independent, and possibly 11 undistributed, all those kinds of things that are 12 important for any kind of analysis.
13 So I'm just trying to figure out why this 14 particular analysis. That's what prompts this thought 15 process.
16 MS. JAMES: David?
17 MR. KREPEL: This is Scott. This is Scott 18 Krepel speaking through a sign language interpreter, 19 if I may. I would like to take an attempt to answer 20 this.
21 Josh Kaizer is one of my staff, but 22 unfortunately he is unable to be with us today. He's 23 on a plane somewhere, I want to say over the Atlantic.
24 He is on his way back from Australia. So I will do my 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 best to answer your question.
1 I think one thing that David left out here 2
in the discussion when it comes to risk evaluation is 3
that the risk evaluation itself uses risk-informed 4
scope of review.
5 They did still do the review of the DBR 6
methodology and looked at a lot of the different 7
aspects of the methodology to make sure that they 8
understood appropriately and had an appropriate 9
establishment of certain criteria, so to speak, to 10 make sure that all of the factors that you just 11 mentioned are addressed. The risk evaluation itself 12 is limited in scope in terms of that review as 13 appropriate with the risks.
14 MEMBER ROBERTS: I'm wondering if this 15 thought process would apply to any analysis. It seems 16 like any analysis has the same potential pitfalls that 17 you've identified on this DBR approach.
18 MR. KREPEL: This is Scott again speaking 19 through a sign language interpreter. Sure. You could 20 apply that same approach with any analysis or 21 methodology. In fact, we typically do, but we don't 22 always document it as clearly as Josh Kaizer did. And 23 I think that's a great model for how people could do 24 this type of thing in the future.
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
24 MEMBER ROBERTS: I think you're probably 1
right. That leads to my next question, which is the 2
conclusion that the four percent, I think you said, 3
overall error is accommodated by the margins that are 4
in other analyses.
5 I look at some of the analyses, like the 6
best-estimate plus uncertainty analysis method, for 7
example, has a 95 percent confidence requirement. I 8
always thought that that was because it was 9
essentially already risk-informed.
10 And 95 percent is good enough because of 11 the very low likelihood of the event itself and the 12 relatively remaining residual amount of defense-in-13 depth that even if you had the event fail, there is 14 still some left. So that's already risk-informed.
15 This now eats into the 95 percent. And 16 I'm wondering if you've thought about that. And the 17 four percent, how much does that eat into 95 percent 18 on an accident or a DNBR 95/95 criteria, something 19 like that? Is this essentially double-spending some 20 of the risk-informed judgement is where my question is 21 going.
22 MR. KREPEL: This is Scott again. I think 23 that's a good comment and a good point. I would say 24 that Josh did look at the consequences in his safety 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
25 evaluation. And I believe you might have already 1
taken a look at that safety evaluation.
2 He discusses various things that might 3
happen if you were to go to 40 percent, a higher power 4
level than estimated, like increased oxidization, for 5
example, for the fuel during normal operation. It's 6
possible there's some loss of margin to the operating 7
limits, but Josh also pointed out that typically you 8
would have margin in the design limit. That is a 9
worst-case scenario.
10 The staff typically, intuitively at least, 11 thinks that is much smaller than typical for that 12 error. Or making a mistake at 40 percent, for 13 example. So the risk would be considered acceptable.
14 The licensees typically do have some margin there in 15 their design limit.
16 MEMBER ROBERTS: Maybe not a fair question 17 at this point, but how high would it be before you 18 would start to worry? Would seven percent be a 19 problem? Would ten percent be a problem? Would 20 20 percent be a problem? Where does the thought process 21 start to break down?
22 MR. KREPEL: No, that's not a fair 23 question. To be honest, there is some engineering 24 judgement there that's involved in that determination.
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26 And so I can't really answer that question myself 1
alone on the spot.
2 MS. JAMES: This Lois James. I would say 3
one of the things that made this unique was we're 4
essentially approving a concept. We know that there 5
are different methods being used at different plants.
6 We know that vendors have different 7
methods for calculating and they have different 8
equipment. So we know we're going to have to go into 9
more detail of those, of some of that when we do 10 individual reviews that get submitted.
11 MEMBER ROBERTS: Thank you. I think I 12 understood that. So if I summarize, what I think I'm 13 hearing is, one, you are treating this as if this is 14 a new analysis method that's expected to be done 15 properly. And I believe your safety evaluation has 16 some conditions in it that help you ensure that the 17 applicant is applying it properly.
18 It's mathematically rigorous. I think 19 your SE would agree with that, that it's a valid 20 approach to use. And you don't expect it to introduce 21 any significant error uncertainty at all. But you 22 also do the side study to say, okay, what if we're 23 wrong?
24 MS. JAMES: Yes.
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27 MEMBER ROBERTS: And based on judgement at 1
what's probably four percent of a very, very high 2
estimate of how bad it could be, you conclude it's 3
probably okay. So I think I understand. I appreciate 4
it. Thank you.
5 MEMBER MARCH-LEUBA: Any more questions 6
from members or comments from the staff?
7 MS. JAMES: I should apologize. David 8
Rahn is having a little trouble with his Teams. It 9
booted him out when he was speaking. He apologizes.
10 He's unable to get back in at the moment.
11 MEMBER MARCH-LEUBA: We have a good sound 12 on our end. Okay. Lois, you can continue with the 13 conclusions.
14 MS. JAMES: Based on the staff's review, 15 we looked at the risk assessment of the DBR results.
16 We looked at previous treatment of similar models and 17 simulations.
18 We looked at previous evaluations of 19 nuclear power plant processes measurement uncertainty.
20 We looked at the understanding of the DBR method and 21 previous treatment of calculations of the feedwater 22 flow and its uncertainty.
23 And based on all of this, we concluded 24 that there is reasonable assurance that the DBR method 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
28 as described in the topical report can be used to 1
determine the core thermal power and the core thermal 2
power uncertainty. Thus, all DBR conditions and 3
limitations have been satisfied.
4 So that's all we had anticipated 5
discussing this morning. Are there any other 6
questions or comments? And I'm going to see if I can 7
get David --
8 VICE CHAIR KIRCHNER: Lois, this is Walt 9
Kirchner. Tom had asked already what I was going to 10 ask at the conclusion.
11 What would you be expecting to see if an 12 applicant comes in with an LAR to upgrade the power?
13 More precision or improved instrumentation on feed 14 order to reduce the uncertainty in that particular 15 parameter or do you think they'll just say, well, the 16 staff looked at four percent.
17 They've kind of shown their hand. And now 18 we'll look at testing the staff to see if we can get 19 a -- I'll pick a number -- three percent upgrade in 20 power, How will you use this when in a practical 21 sense an applicant comes in with a power upgrade 22 application?
23 MS. JAMES: Well, the --
24 VICE CHAIR KIRCHNER: What are you looking 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
29 for? You won't be looking for it on the reduced 1
margin.
2 MS. JAMES: No.
3 VICE CHAIR KIRCHNER: And things like 4
DNBR. So how in practice do you anticipate this being 5
used with an application?
6 MS. JAMES: We have already rolled out 7
this to MURs. And we understand that that's 8
associated with the LEFMs. That's kind of the 9
starting point. So they're going to use this in 10 conjunction with that guidance on how to do -- well, 11 I guess it's a reg guide on how to do the MURs.
12 And then we would expect them to come in 13 with their calculation. How are they going to 14 calculate the uncertainty? What's their computer 15 program? What's their modeling?
16 Since David is not in, I don't know if any 17 EPRI person would kind of like to step in and make the 18 comment at this point of anything further. We expect 19 a lot of the computational models, the uncertainties, 20 how it's going to be used, how it's not going to be 21 used.
22 VICE CHAIR KIRCHNER: Is there a need, 23 Lois, in your opinion for further guidance from the 24 staff for implementation?
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30 MS. JAMES: We have not determined whether 1
we need additional guidance, but that's definitely not 2
out of the realm because we've done it, as I 3
mentioned, for the LEFMs. Our focus has been to get 4
the concept approved and out the door next month.
5 MR. KREPEL: This is Scott, if I could 6
jump in there as well, again speaking through a sign 7
language interpreter here. I just want to give a 8
reminder.
9 In the safety evaluation there are ten, if 10 I recall correctly, ten criteria. Those criteria will 11 be looked at in more detail during the actual review 12 because those are viewed as fundamental to whether or 13 not the methodology is acceptable.
14 MEMBER MARCH-LEUBA: I believe EPRI would 15 like to make a comment.
16 MR. CRYTZER: Hello.
17 MEMBER MARCH-LEUBA: Speak loudly.
18 MR. CRYTZER: Okay. This is Kurt Crytzer 19 with EPRI. The way that we had instructed it was to 20 capture the BDI 2048 method, which is a statistical 21 method, and not put it into any particular software.
22 The software we used would have to be compliant with 23 what BDI 2048 would accept.
24 And so the implementation of this would be 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
31 very much similar to the leading edge flow meters 1
where a correction factor would be applied to the 2
feedwater flow to recapture some of the uncertainty 3
measurement.
4 MEMBER MARCH-LEUBA: Since we have you 5
there, for the record, I believe the correction we're 6
talking about is 15 megawatts electric over 1,000. So 7
it's like 0.1 percent. Would the staff take the risk 8
of operation to say, we were completely wrong but we 9
would expect the correction to be 0.1, correct?
10 MR. CRYTZER: Yes.
11 PARTICIPANT: Can you say your name again 12 clearly for the court reporter?
13 MR. CRYTZER: Kurt Crytzer.
14 PARTICIPANT: That's good for context.
15 MEMBER MARCH-LEUBA: Any more comments or 16 questions from members or the staff?
17 Hearing none, I'm going to open it up in 18 case there's a member of the public that wants to 19 place a comment on the record. Please do so now.
20 Hearing none, I return the meeting to you, 21 Ms. Chair.
22 CHAIR REMPE: Thank you, Jose.
23 At this time, I would note to the court 24 reporter we're going to go off the record. We'd like 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
32 you to come back at 1:00 p.m. East Coast time.
1 (Whereupon, the above-entitled matter went 2
off the record at 9:06 a.m. and resumed at 1:00 p.m.)
3 CHAIR REMPE: Okay. It's 1:00 p.m. on the 4
East Coast. I'd like to ask Member Ballinger to take 5
us through our second topic for this meeting.
6 Ron?
7 MEMBER BALLINGER: Thank you, Madam 8
Chairman. We had a meeting on this topic for the 9
Vogtle LTA at our subcommittee meeting in June where 10 Southern Nuclear and Westinghouse presented an 11 exhaustive, I would say, very thorough presentation of 12 what they claim to do and an analysis that was 13 required.
14 The staff presented their analysis. Their 15 presentations today are a
subset of those 16 presentations. And I think the staff would like to 17 say something initially.
18 MR. MARKLEY: Yes. This is Mike Markley.
19 I'm Chief of Licensing for the Division of Operating 20 Reactor Licensing for the Vogtle site, Units 1 and 2.
21 I'd like to thank you all for helping and 22 reviewing this first-of-a-kind -- it is very much a 23 first-of-a-kind review for us in the fact that it's 24 rated in five percent enriched uranium-235 of 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
33 accident-tolerant fuel assemblies.
And it's 1
applicable only to the current burn-up limits the 2
staff has approved for Vogtle Units 1 and 2.
3 You're going to hear the highlights today 4
regarding our safety evaluation. The staff expects 5
future submittals from the industry requesting higher 6
enrichments and higher burn-ups for longer operating 7
cycles.
8 We have been having pre-licensing of the 9
other licensees and expecting other licensees to 10 submit a license amendment request to transition to 11 24-month cycles in the next few months. We appreciate 12 the subcommittee's questions and comments during its 13 June 21st meeting.
14 For today's briefing, the subcommittee 15 requests that the presentation discusses why NRC has 16 confidence in technical issues to resolve properly, 17 balancing engineering judgement and risk-informed, and 18 look at the long game of batch loading. We are 19 prepared to do that today.
20 The staff welcomes an ACRS letter. We 21 thank you for the opportunity to present and talk with 22 you today. We'll do our best to answer questions.
23 Thank you. I'll turn it over to Southern 24 Company staff.
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34 MEMBER BALLINGER: I think Southern 1
Company's on.
2 MR. MARKLEY: Yes.
3 MR. JOYCE: Thank you. This is Ryan 4
Joyce. I'm the Licensing Manager at Southern Nuclear.
5 First of all, I'd like to thank the ACRS 6
for their consideration of this important initiative 7
that benefits not just SNC but the entire industry as 8
we move to higher enriched fuels. It ultimately will 9
help ensure the safety, reliability, and economics of 10 nuclear power plants, ensuring nuclear power is a 11 viable energy source for many years into the future.
12 The agenda items I'll be discussing will 13 be the LTA program review, request exemptions, summary 14 of testing for adopting AXIOM cladding, and the 15 various analyses that were performed that ultimately 16 will demonstrate the due diligence performed to ensure 17 the LTAs will operate safely and within the analyzed 18 limits.
19 The goal of the program is to irradiate 20 higher enriched fuels in a commercial reactor and 21 generate data in support of future license 22 applications. Although it's a limited number of fuel 23 rods, this will allow us to exercise the regulatory 24 process and work through various issues associated 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
35 with ultimately loading higher enriched fuel 1
assemblies beyond five weight percent, and work 2
through logistical, regulatory, and legal challenges 3
that will come up as we go to this higher enriched 4
fuel.
5 A second objective is to obtain additional 6
data for accident-tolerant and advanced fuel 7
materials, fuel pellet and cladding materials.
8 A future goal of this, which is outside 9
the scope of this amendment, is to ultimately increase 10 the license burn-up limit and go to higher burn-up 11 fuels. Again, that's outside of this specific 12 amendment but that is the end in mind, to ultimately 13 allow the application of higher enriched, higher burn-14 up fuel.
15 For the LTA program, we have four LTAs.
16 Each LTA will contain four fuel rods with up to six 17 weight percent U-235. All LTA rods will have AXIOM 18 cladding. All but one LTA will have chromium coating.
19 About half the rods, 136 for LTA including higher 20 enriched rods, will have doped ADOPT pellets whereas 21 the other half, roughly 128, will be IFBA rods.
22 For the reactor core itself, it will be 23 about 16 higher enriched fuel rods out of 50,952 total 24 rods. So a very small percentage of the reactor core 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
36 will have this higher enriched fuel.
1 The license amendment request requested to 2
change Tech Specs 4.2.1, 3.7.18, and 4.3 regarding 3
these LTAs. Tech Spec 4.2.1 already allows test 4
assemblies that have not completed representative 5
testing to be in non-limiting core regions. In a few 6
slides from now, I'll discuss why we do not believe 7
this non-limiting requirement was met and why we felt 8
the need to explicitly revise Tech Spec 4.2.1.
9 Tech Spec 3.7.18 and 4.3 require an 10 assembly enrichment of less than five percent. Due to 11 the small number of fuel rods above five percent, so 12 only four out of 264 per assembly, this requirement is 13 still met. In other words, the average assembly 14 enrichment is less than five percent.
15 However, we felt it prudent to remove any 16 kind of regulatory uncertainty associated with whether 17 or not we were meeting this requirement. And as I 18 previously mentioned, we wanted to make sure we 19 exercised the regulatory process to lay a road map and 20 ultimately load fuel assemblies greater than five 21 percent enrichment.
22 As will be discussed in a few slides, as 23 part of this amendment we revised our facility 24 operating license to remove an exemption we had to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 70.24 as part of our change in licensing basis, which 1
will be discussed in a few slides.
2 As part of this, we requested an exemption 3
to 50.46 and Appendix K for AXIOM cladding. The 4
exemption request will apply to the application of 10 5
CFR 50.46 and 10 CFR 50 Appendix K of regulations to 6
the LTA design, utilizing AXIOM cladding in Vogtle 7
Unit 2.
8 In conjunction with the 17 percent maximum 9
local oxidation acceptance criteria prescribed by 10 50.46, a more restrictive criteria was assessed 11 consistent with the data presented in the AXIOM 12 topical reports.
13 For regulatory clarity, SNC decided to 14 adopt a newer 50.68 regulation to replace the older 15 70.24 regulation that was described in our facility 16 operating license. We felt that adopting 50.68 17 provided a clean regulatory foundation for moving 18 forward with high-enriched fuel assemblies.
19 Moving to 50.68 necessitated a requested 20 exemption, 50.68(b)(7), to allow the LTAs to have five 21 percent enriched fuel rods. Similar to our tech spec 22 requirements, 50.68 refers to an assembly enrichment 23 of less than five percent.
24 As previously mentioned, technically we do 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
38 meet this on an assembly average, a weight average 1
level, but we felt it prudent to once again request an 2
exemption in lieu of any regulatory uncertainty based 3
on the limited number of higher enriched rods.
4 The technical justification, the intent of 5
50.68 is to include inadvertent criticality. We 6
demonstrate through our very thorough analyses and 7
very conservative analyses that inadvertent 8
criticality well be precluded based on the 9
restrictions we have in place.
10 The NRC approved for the LTAs to include 11 their own criticality analysis. Adherence to these 12 analyses fulfill the 50.68 requirements. The 13 placement is the new storage racks, which is 14 administratively controlled.
15 AXIOM
- cladding, ADOPT
- pellets, and 16 chromium-coated Optimized ZIRLO cladding have all been 17 used in US PWR reactors, similar to Vogtle. In 18 addition, the AXIOM cladding in the topical reports 19 has been interviewed by the NRC and the ARCs.
20 The only novel feature without US 21 operating experience associated with these four LTAs 22 is a very limited number of fuel rods with greater 23 than five-percent enrichment. And again, this 24 represents a very small percentage of the overall 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
39 reactor core.
1 There's about 51,000 rods in the reactor 2
core, so 16 rods out of the 51,000. So a very small 3
percentage of the overall reactor core.
4 MEMBER MARCH-LEUBA: And remind me again 5
from the subcommittee, some of these rods are in 6
limiting positions. Or they're in positions that 7
we're not limiting until -- can you explain that?
8 MR. JOYCE: If I understand the question 9
about limiting, some of these will have a highest, and 10 as mentioned in a couple of slides, will be in --
11 MEMBER MARCH-LEUBA: I'll wait. Yes.
12 MR. JOYCE: Any other questions?
13 VICE CHAIR KIRCHNER: For the record, 14 could you quickly review how many rods with those 15 claddings and composition were tested?
16 MR. JOYCE: For Millstone, unit three.
17 The licensee stated there will be up to eight re-test 18 assemblies containing fuel rods fabricated with AXIOM 19 cladding inserted into the core for Millstone's team.
20 That was from May 2017.
21 Byron Unit 2 from an April 2019 amendment 22 requested to insert two LTAs designed by Westinghouse.
23 The LTAs were based on a vintage optimized fuel 24 assembly design. The licensee proposed to insert up 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
40 to 20 test rods of three different types between the 1
two LTAs. The rods contained a mixture of three 2
materials: uranium, fuel pellets ADOPT, and coated 3
optimizer of a low cladding.
4 VICE CHAIR KIRCHNER: So there's a good 5
database there from that LTA set of experiments?
6 MR. JOYCE: I would say we believe there 7
is a good database that supports those ADOPT, AXIOM, 8
and Permian cladding. I don't know if Westinghouse 9
has anything to add to that.
10 MR. SMITH: Ryan, this is Jim. We agree.
11 The basis for what we were doing is the other 12 applications.
13 MR. JOYCE: So to your question earlier, 14 the LTAs will have the highest linear heat generation 15 rate or local peaking for portions of the cycle, both 16 at steady state and transient conditions.
17 Nonetheless, the technical specification and limits 18 will continue to be met.
19 There are no additions to the reference 20 Tech Spec 5.6.5 needed for these LTAs as the current 21 methods are used to evaluate them.
22 MEMBER MARCH-LEUBA: Will the five percent 23 enrichment rods lean? What's the power inside the 24 bundle for those five-percent enrichment positions?
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41 MR. JOYCE: Radu, can you understand that 1
question?
2 MR. POMIRLEANU: Could you repeat it?
3 Could you repeat the question, please?
4 MEMBER MARCH-LEUBA: There are four bins 5
in the fuel element, an enrichment at about five 6
percent. What is the relative bin power for those 7
bins? Is it much higher than average, about average, 8
or lower than average?
9 MR. POMIRLEANU: It is not significantly 10 different from the leading bins in any given cycle.
11 MEMBER MARCH-LEUBA: Okay. That's good 12 enough. It's about the same. Thank you.
13 MR. POMIRLEANU: Yes. Thank you.
14 MR. JOYCE: You'll see in the slides.
15 It'll be within the FDA bin factor limits that are 16 already in the core report.
17 MR. POMIRLEANU: It's in the core report.
18 This Radu Pomirleanu from Westinghouse.
19 MR. JOYCE: For LOCA, the existing large-20 brick LOCA and small-brick LOCA analysis of record for 21 Vogtle are representative of the LTAs. The fuel is 22 negligibly impacted by the presence of the LTAs. The 23 50.46 acceptance criteria continues to be met.
24 For non-LOCA, the non-LOCA transient 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
42 analysis that depends only on parameters are not 1
impacted by the LTAs as these don't impact core 2
average heat transfer characteristics, decay heat, or 3
initial core energy. The events that depend on local 4
effects were evaluated for potential effects to the 5
LTAs.
6 It was determined that there were no 7
impacts to the codes or methods. Any impact due to 8
LTA features is offset by system margins. While the 9
LTAs may lead the core in factors, they are placed in 10 non-limiting locations with respect to analysis.
11 Ultimately, fuel-specific criteria applicable to each 12 accident continues to be met. There is no impact to 13 the source or consequences. The announcement remains 14 bounding.
15 For fuel rod performance, ADOPT and AXIOM 16 were explicitly modeled to PAD5. Premium coated 17 benefits for corrosion were conservatively not 18 included. For fuel rod design, there is no impact to 19 existing DNBR margin.
20 For core physics, the chromium coating and 21 ADOPT fuel pellets were explicitly modeled. There is 22 no impact to neutronics modeling for the fuel rods 23 about five weight percent. Core monitoring with 24 Beacon is ineffective.
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43 VICE CHAIR KIRCHNER: Can you put that in 1
context of looking ahead to what you want to do next, 2
where you go to higher enrichment? Higher average 3
fuel assembly will have a lot higher enrichment if you 4
put more rods in, obviously, that are six percent.
5 What you're saying here is that for these 6
analyses of record, fuel assembly isn't really -- the 7
margins that you have in a conservative analysis of 8
record are greater than the impact that you can 9
calculate as a result of putting just four rods in?
10 MR. JOYCE: You're speaking regarding the 11 core physics?
12 VICE CHAIR KIRCHNER: Physics and thermal 13 hydraulics.
14 MR. JOYCE: Radu, do you want to take the 15 lead on addressing that question?
16 MR. POMIRLEANU: Sure. This is Radu 17 Pomirleanu from Westinghouse. First of all, I'd like 18 to point out that there will be other licensing 19 submittals that will address publications with an 20 increasing number of higher enrichment fuel and higher 21 burn-ups. But yes, we expect a wider range of impacts 22 that will have to be addressed separately beyond the 23 scope of this submittal.
24 VICE CHAIR KIRCHNER: What I was trying to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 infer simply on your behalf is that was only four rods 1
in a bundle. The impact that it would calculate on 2
your core physics parameters and your thermal-3 hydraulics is really minimal. When you take the next 4
step, there will obviously be an impact on all these 5
evaluations.
6 MR. POMIRLEANU: Yes. That is correct.
7 MR. JOYCE: Any additional comments or 8
questions on slide 10?
9 For the criticality analysis, these 10 assemblies are only requested to be used in the unit 11 to prevent fuel pool. Unit 1 spent fuel pool storage 12 is prohibited.
13 The law addresses increased enrichment, 14 use of ADOPT pellets, and use of chromium-coated AXIOM 15 cladding with regard to storage criticality. Current 16 NRC-approved codes were applied to address LTA 17 storage.
18 For LTA storage not requiring the new fuel 19 storage ranks in Unit 2 2-out-of-4 checkerboard fuel 20 pool storage, a direct analysis was performed. The 21 new fuel storage rack analysis demonstrates a margin 22 of limits including dry, fully flooded, maximum, 23 moderation considerations. The Unit 2 storage racks 24 in Unit 2 2-out-of-4 checkerboard are confirmed for 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
45 the reload process.
1 For spent fuel pool storage, the current 2
burn-up limit for maximum enriched fuel assemblies is 3
about four gigawatt-days per MTU. That's the current 4
limit for about a five percent enriched fuel assembly.
5 To provide significant conservativeness 6
into this limit, a burn-up limit of 64 gigawatt-days 7
per MTU was selected for the LTA storage. So a margin 8
increase of 24 gigawatt-days per MTU. This is greater 9
than the eight percent margin.
10 The 64 gigawatt-days per MTU was selected 11 by additional storage options. Should the LTAs be 12 approved for operation, the burn-up rate is 64 13 gigawatt-days per MTU. In other words, 64 is beyond 14 our burn-up limits. We cannot go to 64 without a LAR.
15 But if we do ultimately go to a higher burn-up with 16 these four LTAs in the third cycle, we can store with 17 the other storage option.
18 MEMBER BALLINGER: This is Ron Ballinger.
19 You used the word if, but it's more like when.
20 MR. JOYCE. Yes, when/if.
21 MEMBER BALLINGER: Okay.
22 MEMBER HALNON: This is Greg. The Unit 1 23 spent fuel, is that prohibited because of lack of 24 analysis or a failed analysis?
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46 MR. JOYCE: I'll let Mike Werner answer 1
this.
2 MR.
WERNER:
Mike Werner from 3
Westinghouse. Essentially, the two pools have 4
different racks so we only analyze those in the Unit 5
2.
6 MEMBER HALNON: So it's lack of analysis?
7 MR. WERNER: Correct.
8 MEMBER HALNON:
Talk to me about 9
physically. Can you physically get a Unit 2 assembly 10 into the Unit 1 spent fuel pool?
11 PARTICIPANT: This is Matt. We're doing 12 this for Unit 2 because that analysis is specific to 13 Unit 2.
14 MEMBER HALNON: There's a gate between 15 them?
16 PARTICIPANT: That's correct.
17 CHAIR REMPE: Excuse me. Someone on the 18 line has their microphone open and it's hard for us in 19 the room to hear what the conversation is. So please 20 check your mics on your computers.
21 Please re-answer again, Matt. Sorry to 22 interrupt.
23 PARTICIPANT: No problem.
24 Yes, that's correct. It's a shared pool.
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47 They do have a physical gate between the two pools.
1 However, they can transfer fuels between them if 2
needed. As I stated before, this was more for an 3
analysis convenience sake to keep it specific to the 4
Unit 2 pool and not do the two.
5 MEMBER HALNON: Is it a routine or has 6
Unit 2 fuel gone to Unit 1?
7 PARTICIPANT: They have. There are cases 8
in a redesign scenario where we can pull from Unit 2 9
fuel to load Unit 1, but that's on a case-by-case 10 basis.
11 MEMBER HALNON: How do you control that?
12 Is that a typical type of --
13 PARTICIPANT: It is. It's controlled via 14 the procedures and processes.
15 MEMBER HALNON: The fuel accounting 16 process?
17 PARTICIPANT: That's correct.
18 MEMBER HALNON: And that's done by the 19 fuel engineering?
20 PARTICIPANT: That's correct.
21 MEMBER HALNON: Okay.
22 MR. JOYCE: When we analyze the old Yankee 23 fuel storage racks, the analyses are very different 24 with regard to the filters in the two fuel racks.
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48 MEMBER HALNON: Okay. You get rid of all 1
the borax and all that stuff, there's nothing that 2
you've got to worry about degradation in the Unit 2 3
fuel pool racks?
4 MR. WERNER: Mike Werner again. We're not 5
crediting any of the older filters.
6 MEMBER HALNON: Okay. Crediting any boron 7
in the pool as well?
8 MR. WERNER: Well, the boron is still the 9
boron credit.
10 MEMBER HALNON: But you aren't crediting 11 that?
12 MR. WERNER: Right.
13 MEMBER HALNON: Okay. Thanks.
14 MR. WERNER: The analysis that's currently 15 in place credits that. We've just built upon it.
16 We're not changing it.
17 MEMBER HALNON: Okay. They're credited 18 the same amount so one can't dilute the other one when 19 you open the gate?
20 MR. WERNER: I'm not sure what the --
21 MEMBER HALNON: I'd be interested in 22 understanding what physically can go wrong that can 23 cause the boron concentration to change. Typically, 24 if it was my plan, I would credit the same amount of 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
49 boron concentration in both pools so one can't dilute 1
the other.
2 MR. WERNER: I believe that the spec limit 3
is the same for both. It's 2,300, I believe.
4 MEMBER HALNON: Okay. That would be my 5
guess but I didn't want to guess.
6 MR. JOYCE: We can confirm that, but 7
what's listed is Unit 1 and Unit 2 -- we have combined 8
tech specs with some different requirements for the 9
Unit 1 and Unit 2 spent fuel pools. Where that 10 requirement is listed is common to both units.
11 MEMBER HALNON: Since I'm elaborating, 12 I'll ask the question, will the operator see anything?
13 Will they see anything during operation or is it 14 thermocouples, SPMDs, or anything in the core?
15 PARTICIPANT: I don't think we -- Radu, 16 I'll defer to you, but my personal answer to this is 17 they shouldn't see anything leave with four bins per 18 assembly. Even if they're in an instrumented 19 location, there could be some minor differences in 20 mapping when we take those in rack engineering.
21 But in terms of operation, I think what 22 we're expecting should be negligible and probably 23 would be within the amount which we typically see in 24 operation.
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50 MEMBER HALNON: Okay. Thanks.
1 PARTICIPANT: Radu, do you have anything 2
additional to add to that?
3 MR. POMIRLEANU: No.
4 PARTICIPANT: Thank you.
5 MR.
JOYCE:
That concludes my 6
presentation. Are there any additional questions or 7
discussions?
8 MEMBER BALLINGER: Okay. I don't know if 9
we have our consultants on the line. I don't know 10 who's online. If not, then thank you very much for 11 your presentation.
12 Now we should transfer to the staff.
13 CHAIR REMPE: I don't see any consultants 14 on the line.
15 MEMBER BALLINGER: Yes. I thought Steve 16 would be online. He was this morning.
17 MR. SCHULTZ: No questions or comments, 18 Ron.
19 MR. BLEY: Dennis is here too. No 20 questions either.
21 CHAIR REMPE: I didn't see you on the 22 list. Sorry.
23 MEMBER BALLINGER: Okay. Thanks again.
24 We just need to change out. The seat is still warm.
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51 There we go.
1 MR. LAMB: My name is John Lamb. I work 2
NRC's Division of Operating Reactor Licensing at the 3
Office of Nuclear Reactor Regulation, NRR. I'll 4
describe the licensing actions that SNC has requested.
5 Charley Peabody, who is virtual and works 6
for the Nuclear Systems Performance Branch, SFSB, in 7
the Division of Safety Systems, DSS and NRR, will 8
discuss the updated final safety analysis report, the 9
FSAR, Chapter 15, accident analyses, the loss of LOCA 10 accident analyses and the non-LOCA accident analyses.
11 Brandon Wise from SFNB, DSS, and NRR will 12 discuss the code analysis and fuel rod design. Kent 13 Wood from SFNB, DSS, and NRR will discuss the fuel 14 handling and storage. Mike Markley, who you heard 15 through the introductions earlier, is the Branch Chief 16 from Doral and NRR and will provide the conclusion.
17 SNC has requested four licensing actions.
18 One is a license amendment request and three are 19 exemptions. The license amendment request is to 20 revise the license condition 2D and four technical 21 specifications.
22 The proposed change in license condition 23 2D is to delete a 1986 exemption to Title 10 of the 24 Code of Federal Regulations, 10 CFR Section 72.4, 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
52 which is criticality analysis requirement and 1
criticality accident requirements, as SNC is 2
voluntarily adopting 10 CFR 50.68.
3 ADOPT, AXIOM, Prime ZIRLO, and Optimized 4
ZIRLO are trademarks or registered trademarks of 5
Westinghouse Electric Company.
6 The three tech specs that SNC is proposing 7
is to revise, one, Tech Spec 3.7.18, which is fuel 8
assembly storage in the fuel storage pool. The second 9
one is Tech Spec 4.2.1, fuel assemblies. And the 10 third one is Tech Spec 4.3, fuel storage.
11 Technical Specification 3.7.18 refers to 12 fuel storage in a fuel storage pool. Therefore, the 13 tech spec note is added to the fuel storage for the 14 accident-tolerant fuel, ATF, lead test assemblies, 15 LTAs, to meet the Tech Spec 4.3.
16 Tech Spec 4.2.1 allows ZIRLO, Zircaloy, 17 and Optimized ZIRLO only. Therefore, the tech spec 18 change is needed for the insulation of the ATF LTAs.
19 In addition, Tech Spec 4.2.1 states a 20 limited number of lead test assemblies that have not 21 completed representative testing may be placed in non-22 limiting core regions. Therefore, the tech spec 23 change is needed to allow --
24 MR. PEABODY: Excuse me, John. Can you 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
53 advance the slides so they can see what you're talking 1
about? Thank you.
2 MR. LAMB: There are a limited number of 3
lead test assemblies that have not been completed.
4 Representative testing may be placed in non-limiting 5
regions. Therefore, a tech spec change is needed to 6
allow SNC to place the four ATF LTAs in limiting core 7
locations in Vogtle Unit 2.
8 Technical Specification 4.3 allows up to 9
five weight percent uranium-235. Therefore, the tech 10 spec is needed for storage of the ATF LTAs.
11 The first exemption is 10 CFR 50.46. The 12 second one is 10 CFR 50 Appendix K. Those proposed 13 exemptions are needed to allow the use of AXIOM 14 cladding. The third exemption is 10 CFR 50.68(b)(7),
15 which allows a greater-than-five weight percent 16 uranium-235.
17 I'm going to turn it over to Charley 18 Peabody to discuss the FSAR Chapter 15, accident 19 analyses.
20 Charley, are you there?
21 MR. PEABODY: Yes. Thanks, John. Can you 22 advance to the next slide? All right.
23 So the accident analyses, most of the 24 accidents were addressed with one of the points that 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
54 was mentioned in the previous discussion. The overall 1
enrichment only increases negligibly by adding the 2
four LTAs when you are addressing core-wide effects.
3 That didn't really change any of the dose 4
consequences for accidents that are evaluated from a 5
core-wide standpoint. There were a few accidents that 6
were evaluated that do have local effects, locked 7
rotor and RCPs. Actually, there were a couple of 8
others.
9 The only one that ended up still being 10 limiting was the rod injection analysis. So the tech 11 specs are going to reflect that the LTA core locations 12 will remain appropriately limiting for rod injection 13 accidents for LTA utilization.
14 That's all on this. Are there any 15 questions on the accident analysis?
16 VICE CHAIR KIRCHNER: Charley, this is 17 Walt Kirchner. When you say limiting, the LTAs cannot 18 go into a control rod position?
19 MR. PEABODY: That's correct.
20 VICE CHAIR KIRCHNER: Is that what you're 21 saying in plain English?
22 MR. PEABODY: Yes.
23 VICE CHAIR KIRCHNER: Okay, just for the 24 record.
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55 MR. WISE: This is Brandon Wise for the 1
staff. I just want to correct that they can go in rod 2
locations but not limiting rod locations.
3 VICE CHAIR KIRCHNER: Okay. That is 4
determined by previous analysis of record? How do you 5
determine a priori which are limiting and which 6
aren't?
7 MR. WISE: I believe SNC and Westinghouse 8
have determined which locations are limiting with 9
respect to control rod location. I believe they tend 10 to be the same locations for most reloads, so they 11 won't be in those locations. And I'm sure there will 12 be some confirmatory analysis to confirm that.
13 MR. PEABODY: Yes. It also would depend 14 on the individual rod worth. I know rods towards the 15 center of the core have more rod worth than control 16 rod locations on the periphery.
17 MR. WISE: All right. I am Brandon Wise 18 with the NRC's Nuclear Methods and Fuel Analysis 19 Branch. I did the review for the code analysis and 20 fuel rod design.
21 For the most part, the codes used by SNC 22 and Westinghouse for the analysis of the LTAs are 23 mostly applicable to the LTAs with the exception of 24 some enrichment limits. Given the limited number of 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
56 higher enriched rods and the amount of enrichment that 1
they have, the overall enrichment of the assembly 2
still remains below five percent, and in a way still 3
remains in the range of applicability for codes such 4
as PARAGON and NEXUS.
5 Overall, there's very little change in the 6
neutronic performance resulting from the higher 7
enriched rods. I'd expect more of a change from the 8
ADOPT fuel pellets, which are being justified by not 9
being in limiting positions with respect to control 10 rod injection due to the potential for more severe 11 control rod injection accidents due to the increased 12 density of the ADOPT fuel pellets.
13 As far as thermal-hydraulic codes go, 14 there's basically no impact as a result of enrichment 15 or ADOPT fuel pellets. There are chromium-coated 16 rods. Although one rod is chromium-coated, we would 17 expect there to be safety enhancement as a result of 18 the chromium coating, but they're not crediting that 19 enhancement.
20 We did examine some potential detrimental 21 effects of the chromium coating and determined that 22 there's no significant loss of margin associated with 23 any potential detrimental effects that were outlined 24 in the ATF-ISG.
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57 So overall, we expect the AXIOM cladding 1
to perform as well as the Optimized ZIRLO cladding, 2
which is what's in the co-resident fuel. And there's 3
no credit for the chromium-coated cladding.
4 Next slide, please.
5 I'll go into a little bit more detail on 6
each of the four LTA characteristics, the first being 7
the AXIOM cladding. It's a zirconium alloy cladding 8
that is expected to demonstrate better in reactor 9
performance compared to Optimized ZIRLO.
10 For the sake of this application, we 11 assumed it performs as well as Optimized ZIRLO.
12 Therefore, it's on par with the rest of the co-13 resident fuel.
14 The chromium coating is a thin chromium 15 coating. It has corrosion resistance, enhanced 16 corrosion resistance compared to Zircaloy cladding.
17 There's no impact to the thermal-hydraulic analysis.
18 The coating is extremely thin and thermal-19 hydraulic analysis doesn't have enough resolution to 20 even capture the reduction in flow area that would 21 result from the chromium coating. And of course, 22 there's no benefit taken for the chromium coating.
23 For each assembly there's four enriched 24 rods enriched to six percent. There's minimal impact 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
58 to neutronic performance as a result because the 1
overall assembly enrichment is going up by a 2
negligible amount.
3 There's actually a much higher increase in 4
fuel content associated with the ADOPT fuel pellets, 5
which are a slightly higher density. And they are 6
coated with chromium and alumina.
Several 7
characteristics are different from standard UO2 8
pellets. That enhances the performance of those ADOPT 9
pellets.
10 One of the concerns the staff had with the 11 ADOPT fuel pellets was the increased fuel content and 12 the potential for more severe reactivity in accidents 13 such as control rod injection. This was dispositioned 14 by limiting the locations in which the LTAs can be 15 stored in the core to non-limiting locations with 16 respect to control rod injection.
17 Next slide, please.
18 Any questions for fuel rod design or 19 coating analysis? Okay. I'll hand the presentation 20 over to Kent Wood, who will discuss fuel handling and 21 storage.
22 MR. WOOD: Good afternoon. My name is 23 Kent Wood. I'm here to do fuel handling and storage.
24 We've done this several times. You're getting four 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
59 rods in higher enrichment to six percent, which adds 1
a very small amount of material. The theoretical 2
increase in density actually has much more visible 3
material to the fuel assemblies.
4 For the analysis, they compared the 5
analysis. They used applied engineering judgement to 6
compare this analysis to the analysis of record and 7
then added copious amounts of margin.
8 In particular, for the new fuel storage 9
and for the spent fuel storage for the 2-out-4, which 10 is the fresh fuel assembly storage in the spent fuel 11 pool, they credited IFBA, which is not in the analysis 12 of record for the new fuel storage or spent fuel 13 storage. So that provides a lot of margin.
14 As Southern said, they really can't get to 15
-- they also did an analysis where they can credit 16 burn-up, but they can't get that burn-up they're 17 crediting of 64. It will add another LAR, but I guess 18 they won't have to talk to me that time.
19 So we looked at that. That provides 20 margin. We looked at their analysis of record. They 21 have a copious amount of margins for that 64 gigawatt-22 days for the burned fuel.
23 For the LTR they did for their accident, 24 which was a multiple misloading where they modeled all 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
60 of the LTAs together, they did an analysis. They got 1
acceptable results, but it didn't have the copious 2
margins that the others did. But they also didn't 3
explicitly credit the IFBA, so I took that into 4
account. I said they're going to have a margin there 5
as well so we don't need to go any further.
6 Based on
- that, we have reasonable 7
assurance that this license amendment and the LTAs 8
will meet 10 CFR 50.68(b)(2) and (b)(3), which are the 9
new fuel storage, fresh fuel high and dry, fuel rod 10 moderation, and flooded conditions. And then also 11 (b)(4), which is the spent fuel pool, which is 12 considered a fully flooded application.
13 So we think that that's going to be that.
14 It's reasonable to have an exception to 10 CFR 15 50.68(b)(7), which is the enrichment limit. That 16 concludes my presentation.
17 MR. MARKLEY: Mike Markley again. I'm 18 Branch Chief of Licensing for Vogtle. We appreciate 19 the feedback that we've received from the subcommittee 20 and each of the members. This is a first of a kind 21 for us.
22 We know that we're going to be getting 23 batch loads of higher burn-up and higher enrichment, 24 and they're probably coming faster. So again, we want 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
61 to come to the ACRS early and often, get the feedback 1
from you, and take that into consideration as we go.
2 The NRC staff determined for this request 3
that there's reasonable assurance that the health and 4
safety of the public will not be endangered by 5
allowing SNC to use the four ATF LTAs in limiting 6
locations without completion of representative testing 7
for up to two cycles, except for the locations where 8
the LTAs may not be placed in core regions that have 9
been shown to be limiting with respect to control rod 10 analysis.
11 Again, we would appreciate a letter. We 12 don't need one to proceed, but we really do value the 13 ACRS' feedback. Thank you.
14 MEMBER BALLINGER: Thank you.
15 Questions from the members or consultants?
16 Hearing no others, thank you once again, both Southern 17 Nuclear and the staff, for a good presentation.
18 I think we're back to you, ma'am.
19 CHAIR REMPE: We need to give public 20 comment.
21 MEMBER BALLINGER: Sorry about that. Are 22 there any people out in the public that would like to 23 make a comment? If that's true, please state your 24 name and make your comment.
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62 Hearing none, back to you, Madam Chair.
1 CHAIR REMPE: Okay. This time I will 2
accept. Thank you. I believe you do have a letter.
3 Do you want to read it in?
4 MEMBER BALLINGER: Yes.
5 CHAIR REMPE: We're going to go off the 6
record.
7 (Whereupon, the above-entitled matter went 8
off the record at 1:43 p.m. and resumed at 3:29 p.m.)
9 CHAIR REMPE: Okay. It's about 3:30 here 10 on the East Coast. I'm going to turn the meeting back 11 over to Member March-Leuba to lead us through our 12 third topic for today.
13 MEMBER MARCH-LEUBA: So the topic right 14 now is ARITA from Framatome, which is their transient 15 analysis methodology for essentially everything but 16 LOCA and logic.
17 We covered this topic in our subcommittee 18 meeting on June 22nd. And without further ado, Greg 19 is going to give us some remarks from the staff.
20 MR. SUBER: Thank you. Good afternoon.
21 My name is Gregory Suber. I am the Deputy Director 22 for the Division of Operating Reactor Licensing in the 23 Office of Nuclear Reactor Regulations.
24 I'd like to thank the full committee for 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
63 the opportunity to discuss the staff's review of the 1
Framatome topical report. As previously stated, the 2
staff did present their findings in the ACRS 3
subcommittee meeting in June. We will be providing a 4
high-level overview of the findings and conclusions 5
today.
6 As mentioned in the opening remarks to the 7
subcommittee, this is an effort that culminated over 8
a four-year period. We appreciate Framatome's efforts 9
in working with the staff and resolving some very 10 difficult and complex issues.
11 The staff also appreciates the ACRS review 12 of this topical report and the safety evaluation. The 13 staff plans to issue the final safety evaluation 14 either later this month or hopefully early in August.
15 I look forward to your comments. Thank you.
16 MEMBER MARCH-LEUBA: Thanks, Greg.
17 From Framatome, Allan?
18 MR. McGINNIS: Hi. I'm Allan McGinnis, 19 Licensing Manager for Framatome. I just want to take 20 a second to thank the ACRS members for their time 21 today and listening to the information that's going to 22 be provided on our ARITA topical report and the NRC's 23 draft safety evaluation.
24 ARITA is a methodology that is a key part 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
64 of Framatome's efforts to upgrade our current PWR 1
analysis methodologies with more sophisticated 2
methods. As such, ARITA is going to be part of the 3
basis and platform for further development in advanced 4
fuel management and accident-tolerant fuel.
5 Framatome has a great deal of resources 6
invested in ARITA. We are very anxious to get the 7
final safety evaluation issued so that we can 8
implement the methodology and its benefits for the 9
industry. Thank you.
10 MEMBER MARCH-LEUBA: I wanted to remind 11 everybody that this is an open session. So if we want 12 to discuss questions, don't mention numbers and we'll 13 stay safe.
14 MR. McGINNIS: I also want to thank the 15 NRC for their efforts. They've put a lot of effort 16 in, especially in these last several months to get the 17 safety evaluation out and work with us on some last-18 minute issues. We appreciate that.
19 I'm going to go ahead and pass it on to 20 Buck Barner, who has been the lead on this project for 21 the majority of the development and review.
22 MR. BARNER: Thank you. This is Buck 23 Barner. I'll be presenting today on ARITA.
24 I want to first off say thank you and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 welcome. I'm excited to be here to talk about ARITA 1
in this full ACRS meeting. I appreciate your 2
willingness to be here and listen to us.
3 We're excited about ARITA. We're excited 4
about what it brings to the industry. I'm looking 5
forward to getting it implemented for the first time 6
and excited to hear that we're moving forward with the 7
SE. With that being said, we'll go ahead and move on 8
now into the presentation.
9 So to start off, we'll start with an 10 overview and go into some background and history of 11 the topical. We'll look at the approval request and 12 the range for applicability, describe the evaluation 13 model, and finalize with a summary.
14 Next slide.
15 So what is ARITA? ARITA stands for 16 ARTEMIS/RELAP Integrated Transient Analysis 17 Methodology. As has been said, it is a non-LOCA 18 method. It covers the Chapter 15 non-LOCA events 19 except for the control rod injection.
20 It is a non-parametric approach, which is 21 a novel approach for non-LOCA methodologies, and does 22 employ a Monte Carlo sampling approach. It was 23 developed using the guidance in SRP 15.0.2. It does 24 have some elements of the in-depth process, but was 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
66 generally followed using the 15.0.2 guidance.
1 Along with the standard non-LOCA 2
methodology that's in there, we did address several 3
additional topics. We'll talk about the motivation 4
for that in future slides. That will also include 5
methodologies for mixed-core power distribution 6
control set points and fuel assembly reconstitution.
7 Next slide.
8 To give you some background and history, 9
there's a little time line across the top here. As we 10 move through the slides, we'll progress from left to 11 right, starting up there with some advanced codes of 12 method development.
13 In the 2006 time frame is when we began 14 development of a new set of advanced PWR codes and 15 methods, really with a focus on our neutronics solver 16 ARCADIA, the core TH solver COBRA-FLX, and fuel 17 performance code GALILEO. Around the same time in the 18 industry, there was also a push to replace legacy 19 methods, which was fortuitous and in good timing with 20 what we were doing internally.
21 So with those two things in mind, we moved 22 forward with a goal with several aspects in mind.
23 That was to ensure that we're using our state-of-the-24 art modeling. We wanted to use our best codes and our 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
67 best capabilities that we could. We wanted to 1
implement the best practices we have from decades of 2
our experience in all three of our regions -- the US, 3
French, and German -- and really take advantage of 4
that experience that we have.
5 We wanted to simplify the topical reports 6
and interdependencies to reduce complexity. So we 7
wanted to take what had traditionally been in several 8
smaller topical reports and combine them into a 9
single, consistent topical report. And we also wanted 10 to do all this to facilitate development for future 11 method development and continuing innovations.
12 After these first three codes were 13 approved, our first methodology topical area was 14 approved in 2020. And now today we're here talking 15 about ARITA, which represents that final realization 16 of our objective to bring innovation and improvement 17 to the industry through our advanced codes and 18 methods.
19 Next slide.
20 So a brief time line of the developments.
21 We began working on ARITA and having pre-submittal 22 meetings February 2015. We submitted the topical 23 report in August 2018. After that time, we moved into 24 the licensing and review phase.
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
68 We had the first set of RAIs on December 1
'19. I won't go into the details. We had a lot of 2
interaction with the NRC through RAIs and resolutions 3
to those, audits back and forth, and ultimately over 4
the last five years ended up with the draft SE, which 5
was provided in April of 2023.
6 So we've been committed to this topical, 7
committed to the industry to make sure that we get 8
this thing through. We want to see it through, and 9
we're excited to hear that the SE is at the final 10 steps.
11 Next slide.
12 Moving forward, what does this mean and 13 what does it bring to the industry? Now that we have 14 advanced modeling of actual plant behavior, it helps 15 us understand plant response and safety margins 16 better. With this understanding, it actually helps us 17 focus in areas where safety is really most important 18 and ensuring compliance with the regulations.
19 So events where safety is really being 20 challenged is where we're able to focus and really 21 understand things better, have a better understanding 22 of wherever margins are, and ensure that the plants 23 remain safe.
24 Along with safety, it also helps us bring 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
69 new value to the industry, things like addressing 1
regulatory changes. Reg Guide 1.236, it actually 2
addresses through area but any additional regulatory 3
changes, we believe ARITA sets us up well for 4
addressing any sort of new regulatory changes that may 5
be coming.
6 Also core design and optimization, and 7
then it also helps us move forward with advanced 8
initiatives like advanced fuel management with 9
increased enrichment and high burn-up. So all this 10 allows us with our higher fidelity simulation to 11 provide increased understanding and confidence in our 12 plant safety.
13 Next slide.
14 So range of applicability, we're looking 15 for a Chapter 15 methodology excluding control rod 16 injection. As I mentioned before, we're also looking 17 for applicability to our mixed-core methods, methods 18 to analyze the power density limiting condition of 19 operations and the core safety limit lines, power 20 distribution
- control, and fuel assembly 21 reconstitution.
22 It's generally applicable to Westinghouse 23 pressurized water reactor designs, as well as CE 24 pressurized water reactor designs. It does use 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 approved critical heat force correlations that are not 1
applicable to the correlations in the topical itself, 2
but it will use approved critical heat force 3
correlations.
4 It's mostly within the range of applicable 5
constituent codes that are used within the 6
methodology, which include ARTEMIS/RELAP, S-RELAP5, 7
COBRA-FLX, and GALILEO.
8 Furthermore about the evaluation model, we 9
talked about the codes. ARTEMIS is our 3D nodal 10 simulator code, previously approved in ANP-10297.
11 That's the ARCADIA topical report, which also includes 12 the APOLLO2-A code. That was originally submitted and 13 approved in 2013 and a supplement in 2018.
14 COBRA-FLX is the subchannel thermal-15 hydraulics code, which was approved in 2010. GALILEO 16 is the fuel performance code. That was approved in 17 2020. And S-RELAP5 is our system thermal-hydraulics 18 code that was previously applied in EMF-2310, 19 originally in 2004 and again in 2011.
20 So using these codes, we have developed 21 what we call a three-evaluation-model variance, which 22 we'll refer to as the coupled system thermal-hydraulic 23 and neutronics model, the 0D thermal-hydraulics model, 24 and the static core evaluation model. We'll get into 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
71 more detail on that in the next slides.
1 I'll review the coupling and the 2
evaluation model variance. First we'll talk about the 3
coupled EM. I have an illustration here on the left-4 hand side describing that, with ARTEMIS on the left 5
and S-RELAP5 on the right.
6 Here we have ARTEMIS performing the 3D 7
core neutronics modeling and simulation, with S-RELAP5 8
providing the system thermal-hydraulics simulation, 9
with RELAP passing the mass flow and temperature and 10 core outlet pressure to ARTEMIS, and ARTEMIS passing 11 the core power back to S-RELAP5 through the sector 12 heating rates.
13 That also can work with boron 14 concentrations if they're important to the specific 15 event. We regard it as an optional transfer, only if 16 it's needed.
17 This is what's used in the coupled EM 18 evaluation. It does this in a transient mode for the 19 Chapter 15 transient analyses. There's a time-20 dependent multi-physics solution there. This model is 21 used for cases where SAFDLs are being evaluated and 22 non-SAFDLs.
23 MEMBER MARCH-LEUBA: If I understand 24 correctly, ARTEMIS is the core simulation that has the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 3D core inside the core. RELAP is used for the 1
environmental conditions, simulators, all that.
2 MR. BARNER: Correct.
3 MEMBER MARCH-LEUBA: I don't want to call 4
it the balance of plant, but it's kind of like the 5
balance of plant. Let's call it the balance of the 6
core.
7 MR. BARNER: Balance of the core.
8 MEMBER MARCH-LEUBA: Yes. Okay.
9 MR. BARNER: The second one is what we 10 call the 0DEM. This is similar to what we've seen in 11 legacy methods where data is generated in the 12 neutronics simulator, or ARTEMIS in this case, which 13 is provided to the simulator in S-RELAP5.
14 And S-RELAP5 is there to run stand-alone.
15 Note that this is only used for non-SAFDL events, 16 things like overpressure, secondary overpressure.
17 MEMBER MARCH-LEUBA: So this is when 18 you're trying to provide figures that are mostly the 19 primary boundary? Overpressure, temperatures, things 20 like that, not core performance?
21 MR. BARNER: Correct. And then finally, 22 we have the stack EM, which is Artemis used in stand-23 alone mode. That's used for events where the system 24 thermal-hydraulics is needed.
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73 So for static events such as a statically 1
misaligned rod or a mis-load analysis, then the 2
ARTEMIS core model is used. And that's also used for 3
SAFDLs. You'll see that whenever a SAFDL is being 4
evaluated, the ARTEMIS 3D model is what's being used 5
to evaluate the standards.
6 A little bit of information on the 7
statistical approach. Again, it's a Monte Carlo 8
sampling approach. If you feel it better manages the 9
complexity than the couple of non-LOCA transient 10 analyses, there's a lot of moving parts to this. It 11 allows us to sample that appropriately.
12 It uses a non-parametric approach based on 13 Wilks method. It uses ordered statistics to make a 14 95/95 statement on the figures and merits.
15 For cases where there's events where 16 multiple figures and merits are required for a single 17 event, it does account for that and has an ability to 18 handle multiple figures and merits from a statistical 19 standpoint.
20 I'll also note that some parameters are 21 still biased. This is a statistical method. There 22 are places where we have bias and moved things to be 23 conservative and be more consistent with the safety 24 analysis and not so much a best estimate. So we do 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
74 try to be realistic, and also provide additional 1
margins and conservatisms through biasing some 2
parameters.
3 Because we are using a non-parametric 4
approach, we don't necessarily need to develop full 5
distribution results. We can use a smaller number of 6
realizations, say 59 cases, to make that actual 7
statistical statement. And this is similar to what's 8
been done in previous approaches I've presented to the 9
ACRS and have been approved by the NRC before.
10 If you look over here on the left, this is 11 just an illustration of a typical transient where you 12 may see 59 realizations of a transient across time.
13 And then you're able to use this to help with your 14 statistical statement. This is the type of output 15 that you might get from this.
16 MEMBER MARCH-LEUBA: Can you give me an 17 example of when you use bias in parameters as opposed 18 to sample uncertainty?
19 The way I understand it is when actually 20 obtaining the certainty is too complicated -- one 21 example you gave us during the subcommittee meeting 22 was the position of the control rod when you have an 23 asymmetric -- give us an example of when you bought 24 it.
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75 MR. BARNER: Peaking would be a place 1
where we would use a bias. So if you're radial 2
peaking, you'll use a bias in that case.
3 MEMBER MARCH-LEUBA: So you're using a 4
limiting radial distribution?
5 MR. BARNER: We'll pick it up to the tech 6
spec.
7 MEMBER MARCH-LEUBA: Instead of searching 8
through your core for the words, conditions, you just 9
place it on tech spec?
10 MR. BARNER: The details, I would have to 11 have a full session. But yes, that's the general.
12 MEMBER MARCH-LEUBA: Okay. Thank you.
13 MEMBER ROBERTS: What do your outlet 14 distributions look like? You talked about that in the 15 opening session.
16 Would you say bifurcations or something 17 kind of fishy happening past the 95 percent point in 18 the distributions or do they look pretty much normal?
19 MR. BARNER: We have not seen anything 20 like that, not something we would look at we saw 21 something in distribution that looked strange, we 22 would investigate it. We have not seen that in 23 anything that we've done yet.
24 MEMBER ROBERTS: Is there anything in the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76 methodology that looks for that?
1 MR. BARNER: There are some things we 2
built in for certain parameters where we needed to 3
look at certain areas, but that will probably be 4
talked about more in closed soon. We do that as the 5
methodology --
6 MEMBER MARCH-LEUBA: If you really want 7
the answer to the question, we can sit up a clothes 8
station in five minutes. So don't strain yourself.
9 Just do not give any proprietary questions here. We 10 will select the proper venue.
11 MR. BARNER: So I just want to summarize 12 ARITA. It is a 3D coupled statistical, non-LOCA 13 methodology applicable to see plants.
14 It represents a metered milestone in our 15 commitment to state-of-the-art modeling and it 16 incorporates decades of art industry experience, sets 17 us up, sets the foundation for future development 18 moving forward. Ultimately, the insurance confidence 19 and plant safety compliance with all regulations and 20 requirements is higher fidelity modeling.
21 I think that's all we have.
22 MEMBER HALNON: Just real quick, earlier 23 in the presentation you talked about enhanced 24 marbling, innovation, all these great things that you 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
77 did. Being a layman at this kind of stuff, could you 1
just give me one or two things that are innovative and 2
enhanced in this?
3 You talked about Monte Carlo. I've heard 4
that around forever. You talked about code coupling 5
and I've heard that forever. So what is it that 6
you've spent your last decade on?
7 MR. BARNER: Good question. So yes and 8
yes. It's both a statistical approach and the core 9
coupling, as well as the 3D transient.
10 The ARTEMIS simulator gives us the real 11 feedback. The fact that we have that coupling 12 feedback is what really gives us the ability to see 13 things.
14 The Monte Carlo sampling is around that to 15 take advantage of that coupling and be sure we're not 16 breaking the coupling by doing things that the codes 17 themselves wouldn't allow by taking unrealistic 18 conditions and combining them with the other.
19 So when we were able to do that and couple 20 them all together, that's what really took us to the 21 next level. You said coupling is nothing new and the 22 statistical piece is nothing new, but really it was 23 the combining of those two pieces together --
24 MEMBER HALNON: So more power in smarter 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 brains putting it together and tying the knowledge 1
together so that it all works together?
2 MR. BARNER: Yes.
3 MEMBER HALNON: All right. Thanks.
4 MEMBER MARCH-LEUBA: Any more questions 5
from the members? Then let's switch with the staff.
6 I want you to introduce yourselves and 7
speak clearly into the microphone.
8 MR. GEELHOOD: I'll go ahead and get 9
started. I'm Ken Geelhood. I'm the Project Manager 10 for Reports. I work in NRR and I'm Project Manager.
11 MR. HELLER: Hi. I'm Kevin Heller. I 12 work in the Nuclear Methods and Fuel Analysis Branch.
13 I'm obviously one of the reviewers on ARITA.
14 MR. LEHNING: And John Lehning. I'm also 15 a technical reviewer in the Nuclear Methods and Fuel 16 Analysis Branch.
17 MR. HELLER: All right. Let's launch into 18 it then. Again, I'm Kevin Heller. With me here is 19 John Lehning. It's our pleasure to be here in front 20 of the full committee to present our review of the 21 ARITA topical report. We certainly want to thank you 22 for your time.
23 As I noted on the slide here, I just 24 wanted to quickly point out Pacific National Northwest 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 Laboratory served as our consultant on this review and 1
provided input that the staff used as a basis for the 2
draft safety evaluation.
3 Next slide, please.
4 Okay. I think Framatome did a good job 5
talking about the evaluation model and what that is.
6 So really, the only thing I want to point out on this 7
slide is that the staff's review of ARITA really did 8
focus more so on the unique aspects of the 9
methodology, so primarily the calculational procedure 10 and the uncertainty treatment. I've got a couple of 11 slides on that.
12 MEMBER MARCH-LEUBA: Is it the case that 13 most methodologies have already been approved? The 14 individual components, codes were already approved?
15 MR. HELLER: Yes, that would be correct.
16 A large portion of these codes, yes.
17 MEMBER MARCH-LEUBA: And you leveraged 18 that for approval? You didn't have to redo it?
19 MR. HELLER: Exactly. We'll see that.
20 Next slide, please.
21 Okay. To talk a little bit about the 22 regulatory requirements and guidance, the full list of 23 the regulations and guidance that the staff used can 24 be found within the safety evaluation. The list here 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
80 includes the key requirements and guidance.
1 In particular I wanted to point out 2
Chapter 15 of the Standard Review Plan, which has 3
guidance on how the staff ought to review evaluation 4
models. And of course, the evaluation model 5
development and assessment process or MDAP, which is 6
in Regulatory Guide 1.203, that was used to structure 7
parts of the safety evaluation in order to ensure it 8
was a comprehensive assessment of the ARITA evaluation 9
model.
10 With that, I think we can go to the next 11 slide.
12 Okay. So a little bit of quick overview 13 here on the review history. ARITA, the review can 14 actually be broken up into four phases.
15 So if you actually proceed forward one 16 more time? There we go. We can see the four review 17 phases.
18 I really want to point out that ARITA was 19 one of the most complex, challenging, and intensive 20 reviews that I think either of us has really been a 21 part of in our time at the NRC, not only because there 22 were a number of first-of-a-kind issues but just given 23 the complexity of the subject material, which in the 24 past might have spanned several different topical 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 reports.
1 We're seeing a lot of it really all in one 2
here. A wide range of in-scope event responses, 3
physical phenomena to be understood, limitations, and 4
available data for doing non-LOCA statistical 5
analyses, and the intricacies of the calculational 6
procedure that Framatome proposed.
7 One of the main challenges which probably 8
contributed to this is that to some measure, the sheer 9
amount of information that was needed to be reviewed 10 for this particular topical report. It's clear there 11 was a lot of work and a lot of information that went 12 into this.
13 Starting initially with beginning the 14 review, we really didn't have enough information to 15 begin drafting the safety evaluation or really know 16 which way some of the identified issues would actually 17 be resolved. So it took a bit of time to resolve a 18 number of the RAIs. Many major issues would actually 19 not end up getting resolved until a couple of years or 20 maybe a little bit later within the review effort.
21 So to kind of put that in a little bit of 22 perspective, the total volume of RAI responses was 23 basically equivalent to another topical report.
24 There's an additional 800 pages of information that 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 was provided on top of the original submittal for 1
about 1,600 pages in total.
2 What I want to point out at the last phase 3
there, the draft SE phase, during that final phase 4
there was not a lot of interaction that the staff had 5
with Framatome in that particular portion because the 6
staff was focused on completing the draft safety 7
evaluation.
8 There were a handful of new limitations 9
and conditions that were identified to close out open 10 issues that Framatome's final RAI responses didn't 11 resolve, but we left off the review at Framatome's 12 request given the needed safety evaluation date.
13 So all parties understood that there were 14 some issues that would probably have to get resolved 15 through additional limitations and conditions. And 16 that's why you can see a number of the limitations and 17 conditions afford alternative approaches or 18 justifications.
19 All right. Next slide, please.
20 Okay. So talking about the calculational 21 procedures, what I'm going to try to do is high-level 22 talk about what it is that we looked at and why we 23 found it to be acceptable. When it comes to the 24 calculational procedure, the key review issue that the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
83 staff identified was associated with assessing the 1
ARITA methodology against regulatory guidance 2
concerning suitably conservative initial conditions.
3 When applying this guidance, the staff 4
considered what is the objective of the ARITA 5
methodology with respect to its intended statistical 6
statement and how is that going to be achieved.
7 Ultimately, the tech specs and the safety analyses 8
need to be in alignment.
9 In consideration of the staff's concern 10 was that the as-proposed method would not provide 11 conservative results for plant operation throughout 12 the allowed operating domain. So during the course of 13 the review, through interaction with the NRC staff and 14 through RAI responses, Framatome modified the 15 calculational procedure to establish suitably 16 conservative event definitions.
17 The revised approach is really 18 characterized -- I've got a couple of bullets here.
19 Achieving a 95/95 or a 95 percent probability with 95 20 percent confidence tolerance limits over the allowed 21 operating domain. Utilizing conservative sampling for 22 highly influential parameters. Really, it's kind of 23 a compromise between a traditional approach and a 24 best-estimate sampling approach.
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84 MEMBER MARCH-LEUBA: We talked in the 1
subcommittee about this best-estimate plus 2
uncertainty. We can eventually agree that this 3
approach is best-estimate plus uncertainty plus 4
biases, biases being conservative initial conditions, 5
conservative power distributions. Do you agree on 6
that?
7 MR. HELLER: Yes, I would agree with that.
8 There were a number of biases that were utilized for 9
conservatisms. I'll actually speak to that a little 10 bit on a future slide.
11 MEMBER MARCH-LEUBA: If you state it in 12 writing, we'll talk about whether calling it biases or 13 penalty factors. What do you prefer?
14 To a mathematician, bias means a lot to 15 you. But as a member of the public, I don't know what 16
-- a bias is something psychological. We'll have a 17 spirited discussion later.
18 MR. HELLER: It sounds like an interesting 19 philosophical discussion.
20 MEMBER MARCH-LEUBA: It is the same thing, 21 right?
22 MR. HELLER: I think effectively, as far 23 as mechanistically at the end of the day, yes.
24 So when it comes to the calculational 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 procedure, it was through the establishment of this 1
conservative event definition that the staff 2
ultimately found the calculational procedure to be 3
acceptable.
4 Next slide, please.
5 All right. So talking about the treatment 6
of uncertainty. NRC staff utilized the guidance 7
provided in Chapter 15.0.2 of the Standard Review Plan 8
to review ARITA's methodology treatment of 9
uncertainty. And per this guidance analysis, it 10 should address all important sources of uncertainty.
11 When we're dealing with a statistic-based 12 uncertainty method, one of the major challenges is 13 developing sufficient data to support the uncertainty 14 distributions for a wide array of phenomenon and 15 processes. With that in mind, the NRC staff's review 16 of the ARITA methodology, its uncertainty treatment 17 focused on three areas.
18 The first area was the uncertainty and the 19 input parameters, specifically the distributions and 20 their technical basis. Staff assessed the uncertainty 21 treatment for all the input parameters. As modified 22 by RAIs and limitations and conditions, the staff 23 found the treatment of those uncertainties to be 24 acceptable.
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86 The second area was the uncertainty in the 1
calculated figures of
- merit, specifically the 2
statistical basis for the method to derive the 3
uncertainties in those figures of merit. Framatome 4
proposed approaches to determine the uncertainties in 5
those figures of merit on univariate and multi-varied 6
conditions.
7 Staff assessed those approaches and found 8
them acceptable because they appropriately determined 9
a 95/95 tolerance limit. We actually discussed the 10 details of this within the subcommittee meeting.
11 And then
- lastly, ARITA's overall 12 statistical calculational procedure, specifically 13 whether the procedure assures fidelity of the results.
14 The staff reviewed the process proposed by Framatome 15 and found it acceptable because it's a statistically 16 rigorous approach that assures 95/95 tolerance limits 17 are appropriately calculated.
18 And it appropriately avoids biasing of the 19 procedure itself based on knowledge of the 20 realizations, which could degrade tolerance limits.
21 So it avoids that.
22 Next slide.
23 MEMBER ROBERTS: I guess I could ask you 24 the same question I asked Framatome. With non-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
87 parametric statistics, it's possible to have the 95 1
percent, which is kind of good because you're actually 2
getting more information about the potential worst-3 case scenarios that could happen. Sometimes depending 4
on what you have, they could be very bad.
5 In this case, I'm guessing because you've 6
got so many biased parameters, there aren't that many 7
parameters left that could vary and conspire to come 8
up with some sort of localized minimum or maximum. I 9
was wondering if that's what you'd look for.
10 MR. LEHNING: This John Lehning from the 11 staff. That was a concern that we brought up in the 12 early part of the review.
13 Where I think we ended up, as you alluded 14 to in the question, by focusing in on, as Kevin 15 mentioned, some of the highly influential parameters 16 and some of the biases that Framatome had input into 17 its method, we didn't see evidence of that. And we 18 don't believe there's a reason to think that there 19 would be some special amount of bifurcation or 20 spreading of this distribution beyond what's typical.
21 And we think, in fact, it might be less of 22 an issue for this method because of the biases as 23 compared to some other methods out there that may 24 sample things even more broadly without some the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 biases and special focus on some of these highly 1
influential parameters.
2 MEMBER ROBERTS: Okay. Thank you.
3 MR. HELLER: All right. Next slide.
4 Okay, so validation assessment. In 5
practical terms, assessment and validation of an 6
evaluation model involves checking the results of its 7
computations against test facility measurements for 8
reference data to provide insight into the credibility 9
of the model's predictive capability for similar 10 postulated conditions.
11 Framatome formulated the bulk of that 12 assessment for the ARITA methodology in terms of the 13 constituent codes, which we already mentioned. There 14 was some interval validation comparison for those 15 constituent codes, but there was also a lot of code or 16 model-based validation provided.
17 It was often cited by past review efforts.
18 That frequently made use of measured data and 19 experimental data. So staff examined those past 20 reviews in consideration of the intended application 21 of the ARITA methodology and concluded they largely 22 support the ARITA methodology.
23 In cases where there was limited 24 validation for codes or models, staff concluded that 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
89 some of the conservatisms that were present in the 1
methodology were limitations and conditions acceptably 2
compensated for this. Therefore, the staff concluded 3
that the constituent codes have sufficient 4
capabilities for modeling the phenomena that are 5
relevant to the set of events within the scope of 6
ARITA.
7 Talking about then the evaluation model 8
variance themselves, for the 0D and static evaluation 9
model variance, staff found those to be very similar 10 to stand-alone treatments. And therefore, ultimately 11 found those to be acceptable.
12 For the coupled evaluation model variant, 13 the staff ended up assessing the information exchange 14 between the coupled codes. And also Framatome 15 provided integral assessments. Those results were 16 found to be acceptable.
17 So the NRC staff ultimately then has 18 reasonable assurance that the ARITA methodology has 19 adequate capability of modeling Westinghouse and 20 combustion engineering PWRs for the applicable events.
21 For supplementary evaluation model 22 features, there's a series of what we refer to as 23 supplemental evaluation model features for ARITA.
24 These can serve as additional functionalities that 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
90 Framatome has incorporated into the methodology.
1 For the boron dilution, the mixed core 2
treatment, and for fuel assembly reconstitution, staff 3
found these were acceptable when consideration was 4
given to code conservatisms in some limitations and 5
conditions.
6 For the set points and the power 7
distribution control, the staff ultimately found that 8
these were evolutionary updates to existing methods, 9
retaining the overall existing approaches. And 10 therefore, the staff also found them to be acceptable.
11 Next slide, please.
12 At the conclusion of drafting the safety 13 evaluation, staff imposed a total of 28 limitations 14 and conditions. These can be broadly broken down as 15 seen in the table here.
16 So there's about 16 of them that, really, 17 they're there to assure appropriate treatment of 18 uncertainties or the application of a statistical 19 process. There's approximately ten that are there to 20 assure appropriate application of the ARITA 21 methodology.
22 And then there are two to assure that 23 conservatisms which were cited as justification for 24 the methodology are representative of future fuel 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 designs in existing plants.
1 The staff found that these limitations and 2
conditions were necessary in order to ensure the 3
acceptability of the methodology based on a full 4
review of the material that was contained in the 5
docketed materials that Framatome submitted.
6 The last point then. After receiving the 7
draft safety evaluation for proprietary review, 8
Framatome did inform the NRC staff it intends to 9
pursue resolution of some of the concerns underlying 10 two of the limitations and conditions, 18 and 19, in 11 a future regulatory review process.
12 We've had some dialogue as to possible 13 avenues that could be pursued through that as far as 14 the vehicles. For example, an update or addition to 15 a topical report, inclusion within a plant license 16 amendment request, or another avenue. We anticipate 17 hearing from Framatome proposing a pathway in the near 18 future for that.
19 MEMBER MARCH-LEUBA: Can you give us a 20 high-level description of limitation conditions 18 and 21 19?
22 MR. HELLER: I was trying to do it off the 23 top of my head, but we do actually have a back-up 24 slide. Slide 10, I believe. It's actually after the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 conclusion.
1 MEMBER MARCH-LEUBA: I don't have it on 2
mine.
3 MR. HELLER: You don't have it?
4 You should have it there. There we go.
5 Okay.
6 MEMBER MARCH-LEUBA: We will have it if 7
you put it on the record.
8 MR. HELLER: Yes.
9 MEMBER MARCH-LEUBA: If it's on the 10 record, you have given us a copy.
11 MR. HELLER: Yes. So limitation and 12 condition 18 deals with the axial peaking factor.
13 Basically, what it says is licensees shall justify the 14 approach that Framatome proposed for treating its 15 uncertainty. Or they could conservatively estimate it 16 via what's iterated within the limitation and 17 condition, or they could justify another approach.
18 Really where this is coming from is the 19 staff couldn't conclude from the material that was 20 provided that Framatome's proposed approach directly 21 followed.
22 MEMBER MARCH-LEUBA: Couldn't we have 23 followed with a bias? That's typically for LOCA, 24 right, or not?
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93 MR. LEHNING: I think that in essence is 1
similar to what we've found is being done here. So 2
yes, there is a delicate balance here in terms of 3
that.
4 We did try to take what we thought was 5
somewhat of a bounding approach, especially on 19 6
because of the limitations and some of the data that 7
underlies that item. So I think that's probably what 8
I can say here in this session. We could go into --
9 MEMBER ROBERTS: I have a simple question.
10 So you had 28 total limitations and conditions. I 11 imagine some of those were legacy LNCs from the 12 previous reviews because certainly you always have 13 limitations and conditions.
14 Were some resolved from the past? Did a 15 number of them just not apply? How many are new?
16 MR. LEHNING: I think they basically are 17 new in general. And now the existing -- I think what 18 you're referring to is there's previous safety 19 evaluations that the staff has written on these other 20 methods.
21 We have words in our safety evaluation 22 that basically, the previous safety evaluations also 23 have to be respected. So we didn't incorporate any of 24 those existing limitations and conditions into our 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
94 safety evaluation. The understanding is that those 1
methods would be used.
2 MEMBER ROBERTS: So these are kind of in 3
addition to the legacy applications. Okay. So 4
really, in a sense there's more than 28. There's 28 5
plus the old ones too. Okay. Those ones don't get 6
captured again in your SE?
7 MR. LEHNING: That's true. The way that 8
the methods work, there probably are some -- for 9
instance, if we were to look back at -- I think in 10 particular, the ones that are the clearest are the 11 ARTEMIS and GALILEO because those codes would be used 12 in the same manner basically that they're being used 13 in those code specifics or the topical reports that we 14 reviewed.
15 The S-RELAP5 might be a little bit 16 different in the sense of the realistic LOCA topical 17 report. There are some aspects of that that may or 18 may not necessarily -- some of the limitations might 19 trivially be satisfied because they're not really 20 applicable to non-LOCA analysis.
21 I'd say again, for the 28 items that are 22 in the safety evaluation for ARITA, a lot of those 23 deal with things that are specific to ARITA in terms 24 of the uncertainty distributions and the calculation 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
95 procedure for that evaluation model. So there is not 1
a repeat of existing safety evaluation limitations and 2
conditions into this safety evaluation we've written 3
for ARITA.
4 MEMBER ROBERTS: Okay.
5 MR. HELLER: So for limitation and 6
condition 19, this is somewhat a similar situation in 7
that there was a proposed approach for treating the 8
uncertainty with the axial peaking factor for a 9
specific event. The staff could not conclude from the 10 information that was provided before the review ended 11 that the approach was appropriate.
12 MEMBER MARCH-LEUBA: The solution was the 13 staff proposed what they considered to be a 14 conservative treatment of the uncertainties, which 15 Framatome considers to be too conservative; is that 16 correct?
17 MR. HELLER: I think that would be a fair 18 assessment.
19 MEMBER MARCH-LEUBA: Probably the solution 20 would be the first LAR to be followed by a supplement.
21 That seems like a path forward, but that's my opinion.
22 MR. HELLER: So really, the last thing I 23 wanted to point out with this is -- you can see it 24 there at the bottom of the slide -- the introduction 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
96 of these limitations and conditions, they reflect only 1
where we ended the review based on the docketed 2
material. There was a cut-off date and it was 3
understood that there might be some things that end up 4
resulting. And these were two of them.
5 MEMBER MARCH-LEUBA: Okay. Thank you. Go 6
back to your conclusions.
7 VICE CHAIR KIRCHNER: It begs a different 8
question, a more generic question on limitations and 9
conditions. How many of them are based on physics and 10 how many of them are based on events? In other words, 11 the methods of the first order don't know it's a 12 streamline break.
13 You put those boundary conditions on the 14 system in the analysis but is it -- do you see where 15 I'm going with this? It sounds like a lot of 16 limitations and conditions on top of those that 17 already exist. So are they mainly because of 18 inadequacies and treating physical events or just the 19 range that the code is being applied to?
20 MR. LEHNING: Maybe I'll take a first 21 crack at that. I don't think a lot of them have to do 22 with basically physics or perceived inadequacies in 23 the way that the physics of the code works.
24 I think we feel pretty comfortable with 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 those codes and the models. And any of the issues 1
with those would have been identified in these 2
previous reviews that we just discussed in response to 3
a previous question.
4 So I'd say the majority, if not all, of 5
these limitations and conditions really go to how the 6
calculation ought to be done and whether we agreed 7
with the specific way that Framatome said we're going 8
to -- maybe assumptions, inputs, or other things like 9
that much more so.
10 MR. HELLER: I would simply follow that up 11 with -- I guess I would ultimately end up reiterating 12 to a certain extent but using different words.
13 Alluding back to what Framatome said, the novel 14 aspects to this, these constituent codes are being 15 applied in a particular manner.
16 So you can kind of think of it as there's 17 a wrapper around those codes. It's the nature of that 18 wrapper that -- those limitations and conditions exist 19 to ensure that it's being applied appropriately.
20 VICE CHAIR KIRCHNER: Okay. I'm testing 21 you because I didn't want to hear that it was 22 something fundamental in the neutronics or whatever.
23 It's more how it's tied together and how it's applied.
24 Okay.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
98 MR. HELLER: Yes. That would be the case.
1 Okay, conclusions then. Staff found that 2
the ARITA methodology is acceptable for modeling in-3 scope Chapter 15 events from the Standard Review Plan.
4 This includes all three of the ARITA evaluation model 5
variants, the associated calculational process, and 6
the statistical uncertainty methodology. The staff 7
also found that the supplementary evaluation model 8
features are acceptable.
9 MEMBER MARCH-LEUBA: Can you explain what 10 those are?
11 MR. HELLER: Let me actually go back up in 12 my notes here. You can stay on this slide.
13 MEMBER MARCH-LEUBA: Is it boron?
14 MR. HELLER: Yes. Boron dilution, the 15 setpoints approach, the fuel assembly reconstitution, 16 the mixed core treatments.
17 MEMBER MARCH-LEUBA: That's good, a good 18 example. We have memorized the Section 3.8.
19 MR. HELLER: So the NRC staff found it was 20 acceptable. And the staff's conclusions then are 21 predicated upon two things.
22 First, the ARITA methodology being used 23 within its proposed range of applicability. And that 24 licensees acceptably address the limitations and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
99 conditions within the safety evaluation. That's it.
1 MEMBER MARCH-LEUBA: Thank you.
2 Any more questions from the members?
3 MEMBER HALNON: I just wanted to explore 4
back on your schedule slide. You had an acceptance 5
review. Did you take it back or just not accept it?
6 You said the first acceptance review was 7
completed and then you had to supplement it. I know 8
this doesn't affect the fact that everything is 9
acceptable and you guys have done a good and thorough 10 job of getting through this, but that's a long time 11 and a lot of RAIs.
12 MR. LEHNING: Basically, what happened 13 there was that we found information during the 14 acceptance review that staff felt was not fully 15 complete, and so there was a supplement. It wasn't --
16 MEMBER HALNON: So this was if you can 17 provide this, then we can -- it wasn't extended beyond 18 the 45 days?
19 MR. LEHNING: It wasn't rejected.
20 MEMBER HALNON: Okay.
21 MR. LEHNING: There's an option to reject 22 it, accept it, or to accept it with a supplement.
23 This went down the path of accepting it with a 24 supplement.
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100 (Simultaneous speaking.)
1 MR. LEHNING: Correct. That's true. It's 2
at this very coarse level of detail. Is all the 3
information we need there to do the full detailed 4
review? And we found that there was not. This yellow 5
bar represents basically the time to complete that.
6 MEMBER HALNON: As you ramped up your 7
understanding of it, you can see it getting more and 8
more detailed and more involved. Did you predict it 9
was going to take that long? Or did it kind of just 10 occur because of all of the technical complexities?
11 MR. LEHNING: We did. I think we had a 12 pretty clear idea that the review schedule was going 13 to be substantially longer than the standard two-year 14 review period that we assume for topical reports, but 15 there are a lot of uncertainties.
16 For example, there were some issues that 17 even in -- I think there were three pre-submittal 18 meetings for ARITA. There were some issues that we 19 identified there very early on that continued to 20 persist through the review and maybe didn't even get 21 resolved until June 2022.
22 That was because, I think, the vendor in 23 this case had a principal idea that they were in the 24 right, that their approach was justified, and it took 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
101 some time before we could get to that point where 1
there was this agreement on going forward with that.
2 So yes, the exact length of time was not 3
completely clear, but we had a clear idea from the 4
very beginning. Framatome was aware of that when they 5
got into the review.
6 MEMBER HALNON: I mean, the 19 meetings 7
and all that, we've approved entire reactor plants in 8
much less time. I can only imagine the complexity and 9
how deep you guys got. It's good. I'm glad you 10 presented that because it gives us the context of just 11 how complex it was.
12 MEMBER MARCH-LEUBA: Any more questions 13 from members, even the member on the phone line? I 14 assume if you don't raise your hand, you don't want to 15 talk.
16 MEMBER DIMITRIJEVIC: No questions, Jose.
17 MEMBER MARCH-LEUBA: Thanks, Vesna.
18 So at this point, I'd like to open the 19 floor for members of the public. If somebody wants to 20 place a comment on the open transcript, please do so 21 now. I hear none.
22 Madam Chair?
23 CHAIR REMPE: Thank you.
24 At this point, I'd like to tell the court 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
102 reporter we're going off the record for the rest of 1
this meeting. Thanks for your support.
2 (Whereupon, the above-entitled matter went 3
off the record at 4:20 p.m.)
4 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
Ryan Joyce - SNC July 12th, 2023 Lead Test Assembly (LTA) License Amendment Request (LAR) - 707th ACRS Meeting -
July 12-14, 2023
2 Agenda
- Vogtle LTA Program Overview
- Requested Exemptions
- Summary of In-Pile Testing (ADOPT, AXIOM, Cr-coated cladding)
- Analysis
3 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 Westinghouse ATF LTAs with higher enrichment capable of higher burnup
- 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.
4 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
- 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
6 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
7 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.
8 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)
9 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
10 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
11 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
Questions/Discussion
13
- AOR: Analysis of Record
- BWR: Boiling Water Reactor
- DNB: Departure from Nucleate Boiling
- GWd: Gigawatt Days
- IFBA: Integral Fuel Burnable Absorber
- LAR: License Amendment Request
- LBLOCA: Large LOCA
- LOCA: Loss of Coolant Accident
- LTA: Lead Test Assembly
- MLO: Maximum Local Oxidation
- MTU: Metric Ton Uranium
- NRC: Nuclear Regulatory Commission
- PCT: Peak Clad Temperature
- PIE: Post-Irradiation Examination
- PWR: Pressurized Water Reactor
- SBLOCA: Small LOCA
- SLB: Steam Line Break
- RWSC: Rod Withdrawal from Subcritical
- SFP: Spent Fuel Pool
- SNC: Southern Nuclear Company
- TS: Technical Specification
- wt.%: weight percent Acronyms and Terms
14 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
NRC Staff Presentation Accident Tolerant Fuel (ATF)
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
- 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 U-235 - Uranium 235 UO2 - Uranium Dioxide MWd - Megawatt-Day MTU - Metric Ton Uranium ZrB2 - Zirconium Diboride IFBA - Integral Fuel Burnable Absorber LOCA - Loss-of-Coolant Accident EM - Evaluation Model RCS - Reactor Coolant System 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 "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 ML21053A027 3/16/2021 NRC issued its completeness and withholding Determination for EPRI DVR for MUR Topical Report ML21110A049 5/2/2022 NRC issued request for additional information regarding regulatory review scoping issues (RAIs)
ML22118A052 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 the RAI 14 Response ML23103A150 5/6/2023 NRC issued the draft safety evaluation for EPRI proprietary review ML23088A184 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)
Determination of CTP relies on feedwater flow measurements 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 uses statistical analysis of multiple plant measurements, to provide accurate CTP readings 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
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee, July 12th, 2023 2
2 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 ACRS Full Committee, July 12th, 2023 3
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 4
Background and History
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.
Advanced C&M Development ARITA Development ARITA Licensing ARITA Future
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee, July 12th, 2023 5
Background and History
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 Advanced C&M Development ARITA Development ARITA Licensing ARITA Future
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee, July 12th, 2023 6
Background and History
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 Advanced C&M Development ARITA Development ARITA Licensing ARITA Future
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee, July 12th, 2023 7
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 8
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 9
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 10 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 11 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 12 Acronyms AFM - Advanced Fuel Management AREA - ARCADIA Rod Ejection Accident ARITA - ARTEMIS/RELAP Integrated Transient Analysis CE - Combustion Engineering CHF - Critical Heat Flux CRE - Control Rod Ejection CSLL - Core Safety Limit Lines EM - Evaluation Model FOM - Figure of Merit LOCA - Loss of Coolant Accident LPD LCO - Local Power Density Limiting Condition of Operation Non-LOCA - non-Loss of Coolant Accident NRC - U.S. Nuclear Regulatory Commission PDC - Power Distribution Control PWR - Pressurized Water Reactor SAFDL - Specified Acceptable Fuel Design Limits SRP - Standard Review Plan
ARITA - ARCADIA/RELAP Integrated Transient Analysis Methodology ACRS Full Committee, July 12th, 2023 13 Thank you
Any reproduction, alteration, transmission to any third party or publication in whole or in part of this document and/or its content is prohibited unless Framatome has provided its prior and written consent.
This document and any information it contains shall not be used for any other purpose than the one for which they were provided.
Legal and disciplinary actions may be taken against any infringer and/or any person breaching the aforementioned obligations.
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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NRCStaffsReviewof FramatomeTopicalReportANP10339P, ARITA-ARTEMIS/RELAPIntegrated TransientAnalysisMethodology OpenPresentationtothe AdvisoryCommitteeonReactorSafeguards July12,2023 K. Heller,U.S.NRC J. Lehning,U.S.NRC K. Geelhood,D.Richmond,T.Zipperer, B. Schmitt,D.Engel,PNNL
Introduction
- TheARITAmethodologyisastatisticalapproach forperformingmostStandardReviewPlan(SRP)
Chapter15reactorsafetyanalyses
- NotincludingLOCAandrodejection
- ApplicabletoconventionalWestinghouseand CombustionEngineeringPWRs
- ARITAinvolvesthreedistinctevaluationmodel variants
- ThecodesusedintheARITAmethodologyhave beenpreviouslyreviewedbytheNRCstaff
- NRCstaffsreviewfocusedmainlyonthe calculationalprocedureanduncertainty treatments 2
Key Regulatory Requirements and Guidance
- 10 CFR 50,AppendixA,GeneralDesign Criteria,e.g.,
- GDC10,ReactorDesign
- GDC15,ReactorCoolantSystemDesign
- 10 CFR 50.36,TechnicalSpecifications
- 10 CFR 50.67or10CFRPart100DoseLimits
- StandardReviewPlan,Chapter15
- RegulatoryGuide1.203,EvaluationModel DevelopmentandAssessmentProcess 3
Review History FRAMATOME NRC ANP10339P Submitted Aug2018 Acceptance Review Complete Nov2018 Acceptance Review Supplement Mar2019 Onsite Audit Sept2019 Batch1 RAIs(13)
Issued Dec2019 Prospective L&CsFirst Discussed Dec2020
Response
to13RAIs Mar2020
Response
to37RAIs Jul2020
Response
to16RAIs Dec2020
Response
to26RAIs May2021 UpdatedTopical Report/Final RAIResponse Jun2022 Batch2 RAIs(79)
Issued Apr2020 ACRSSC Jun2023 DraftSE Apr2023 2018 2019 2020 2021 2022 2023 19Audits&Meetings toResolveRAIs Nov2020- April2022 4
Accptnce Supplmnt Detailed Review RAIResolution DraftSE
Calculational Procedure Keyreviewissue:AssessingARITAmethodologyagainstregulatory guidanceconcerninginitialconditions,forexampleSRP15.0.I.6.C.ii:
- Thereviewerverifiesthatinitialconditionsusedintheanalysesare suitablyconservative WhatisobjectivewithrespecttoARITAsintendedstatistical statementandhowtoachieveit?
- TechnicalSpecificationsandsafetyanalysesneedtoalign Staffsconcern:assubmittedmethodmaynotprovideconservative resultsforplantoperationthroughouttheallowedoperatingdomain Establishaconservativeeventdefinition
- Achieve95/95forpostulatedeventovertheallowedoperatingdomain
- Conservativesamplingofhighlyinfluentialparameters
- Compromisebetweentraditionalandbestestimate samplingapproaches 5
Treatment of Uncertainty
- TheARITAmethodologyusesastatistical samplingapproachforuncertaintyquantification
- Wilksmethod:nonparametricestimateof95/95
- Challengeliesindevelopingsufficientdatatosupport uncertaintydistributions
- NRCstaffsreviewfocusedontreatmentof:
- Uncertaintyininputparameters
- Asmodified,treatmentsfoundacceptable
- Uncertaintyincalculatedfiguresofmerit
- Univariateandmultivariateapproachesacceptable
- Overallstatisticalcalculationprocedure
- Rigorousdeterminationoftolerancelimits
- Appropriatelyavoidsbiasingthatcould degradetolerancelimits 6
Validation Assessment
- ConstituentCodes
- SRELAP5,ARTEMIS,COBRAFLX,GALILEO
- Mostvalidationscitedpastreviews
- Conservatismsandlimitationsandconditionsaccountfor cases/areasoflimitedvalidation
- EvaluationModelVariants
- 0DandStatic
- Verysimilartopriorstandaloneapprovals
- Coupled
- Assessedinformationexchangeandtheprovidedintegralassessments
- SupplementaryEvaluationModelFeatures
- Acceptablewithappropriateconservatismsinplace
- Largelyevolutionaryupdates 7
Limitations and Conditions
- Atotalof28LimitationsandConditions
- Approximatebreakdown:
- FramatomeintendstoaddressL&Cs18 and19inafuturereview 8
16 StatisticalProcess and/or UncertaintyTreatment 10 MethodologyApplication 2
ConservatismAssurance
Conclusions
- TheNRCstafffoundtheARITAmethodology acceptableformodelinginscopeSRPChapter15 events,including
- allthreeARITAevaluationmodelvariants
- theassociatedcalculationalprocess
- thestatisticaluncertaintymethodology
- TheNRCstafffoundthesupplementaryevaluation modelfeaturesdescribedinSection3.8ofitssafety evaluationacceptable
- Thestaffsconclusionsarepredicatedupon
- theARITAmethodologybeingusedwithinitsproposed rangeofapplicabilityinSection13.0ofANP10339P
- licenseesacceptablyaddressinglimitationsand conditionsinSection5.2ofthestaffs safetyevaluation 9
Limitations and Conditions 18 and 19
- LimitationandCondition18
- LicenseesshalljustifyFramatomesproposedapproachfortreatingthe uncertaintyintheaxialpeakingfactor(FZ),conservativelyestimateit, orjustifyanotherapproach
- Impetus:StaffcouldnotconcludeFramatomesproposedapproach directlyfollowsfromtheinformationpresentedduringthereview
- LimitationandCondition19
- Forthestatedevent,licenseesshalltreattheuncertaintyfortheaxial peakingfactor(FZ)forthepostscrammainsteamlinebreakeventvia thespecifiedconservativeapproachorjustifyanalternative
- Impetus:NRCstaffconcludedFramatomedidnotadequatelyjustify theproposeduncertaintytreatmentwouldwellcharacterizetheevent conditions
- IntroductionofL&Cs18and19reflectsonlywherewe endedthereview,basedondocketedmaterial 10
Future Framatome Plans to Address Limitations and Conditions 18 and 19
- NRCstaffandFramatomehavediscussedpotentialpathways toaddressLimitationsandConditions18and19inafuture regulatoryreview
- Additionaltechnicaljustificationmayreduceoreliminateimpactof theselimitationandconditions
- Additionaltechnicaldialoguewouldfacilitatetimelyresolution
- NRCstaffwillcompletecurrentreviewofANP10339Pand issuefinalSEasscheduled
- Framatometoproposepathwaytoaddresstechnicalissuesin LimitationsandConditions18and19inaseparatereview process
- e.g.,expeditedreviewofupdate/revisiontothetopical report,inclusioninaplantlicenseamendment request,etc.
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