ML22132A149
| ML22132A149 | |
| Person / Time | |
|---|---|
| Issue date: | 04/06/2022 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Brown, C; Snodderly, M; ACRS | |
| References | |
| NRC-1908 | |
| Download: ML22132A149 (260) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Docket Number: (n/a)
Location: teleconference Date: Wednesday, April 6, 2022 Work Order No.: NRC-1908 Pages 1-195 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 694TH MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 WEDNESDAY 9 APRIL 6, 2022 10 + + + + +
11 The Advisory Committee met via 12 teleconference, at 8:30 a.m., Joy L. Rempe, Chairman, 13 presiding.
14 15 COMMITTEE MEMBERS:
16 JOY L. REMPE, Chairman 17 WALTER L. KIRCHNER, Vice Chairman 18 DAVID A. PETTI, Member-at-Large 19 RONALD G. BALLINGER, Member 20 VICKI M. BIER, Member 21 CHARLES H. BROWN, JR., Member 22 VESNA B. DIMITRIJEVIC, Member 23 GREGORY H. HALNON, Member 24 JOSE A. MARCH-LEUBA, Member 25 MATTHEW W. SUNSERI, Chairman NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 1 DESIGNATED FEDERAL OFFICIAL:
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
3 1 CONTENTS 2 Call to Order and Opening Remarks . . . . . . . . 5 3 Radiation Embrittlement of Reactor Vessel 4 Materials . . . . . . . . . . . . . . . . . . . . 8 5 BWRX-300 Topical Report on Containment 6 Evaluation Methodology . . . . . . . . . . . . . 59 7 Point Beach Subsequent License Renewal 8 Application . . . . . . . . . . . . . . . . . 124 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 1 P R O C E E D I N G S 2 8:30 a.m.
3 CHAIRMAN REMPE: Good morning. This 4 meeting will come to order. This is the first day of 5 the 694th meeting of the Advisory Committee on Reactor 6 Safeguards. I'm Joy Rempe, the Chair of the ACRS.
7 Members in attendance are Ron Ballinger, 8 Vicki Bier. Charles Brown will be here soon; he's 9 been a little bit delayed by traffic. Vesna 10 Dimitrijevic, Greg Halnon, Walt Kirchner, Jose March-11 Leuba, Dave Petti, and Matthew Sunseri. I note we do 12 have a quorum today.
13 Today, the Committee is primarily meeting 14 in person with some of the ACRS staff, NRC staff, and 15 participants attending virtually.
16 The ACRS was established by the Atomic 17 Energy Act and is governed by the Federal Advisory 18 Committee Act. The ACRS section of the U.S. NRC 19 public website provides information about the history 20 of this Committee and documents, such as our Charter, 21 Bylaws, Federal Register notices for meetings, Letter 22 Reports, and transcripts of all full and subcommittee 23 meetings, including all slides presented at meetings.
24 The Committee provides advice on safety 25 matters to the Commission through its publicly-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 1 available Letter Reports.
2 The Federal Register notice announcing 3 this meeting was published on March 15th, 2022. This 4 announcement provided a meeting agenda, as well as 5 instructions for interested parties to submit written 6 documents or to request opportunities to address the 7 Committee. The Designated Federal Officer for this 8 meeting is Mr. Christopher Brown.
9 A communications channel has been opened 10 to allow members of the public to monitor the open 11 portions of the meeting. The ACRS is now inviting 12 members of the public to use the MS Teams link to view 13 slides and other discussion materials during these 14 open sessions.
15 The MS Teams link information was placed 16 in The Federal Register notice and agenda on the ACRS 17 public website. If you are a member of the public who 18 does not yet have the link, please email Lawrence 19 Burkhart at lawrence.burkhart@nrc.gov. Again, that's 20 lawrence.burkhart@nrc.gov.
21 It's my understanding we have received no 22 written comments of requests to make oral statements 23 from members of the public regarding today's session.
24 Periodically, the meeting will be open to 25 accept comments from participants listening to our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
6 1 meetings. The written comments may be forwarded to 2 Mr. Christopher Brown, the Designated Federal Officer.
3 During today's meeting, the Committee will 4 consider the following topics: Radiation 5 embrittlement of reactor vessel materials; BWRX-300 6 Topical Report on Containment Evaluation Methodology, 7 and the Point Beach subsequent license renewal 8 application. As stated in the agenda, portions of the 9 BWRX-300 Topical Report may be closed.
10 A transcript of the open portions of the 11 meeting is being kept, and it's requested that 12 speakers identify themselves and speak with sufficient 13 clarity and volume, so they can be readily heard.
14 Additionally, participants should mute themselves when 15 not speaking.
16 At this time, I'd like to ask other 17 members if they have any opening remarks.
18 Hearing none, I'd like to ask Ron 19 Ballinger to lead us through our first topic of 20 today's meeting. Ron?
21 MEMBER BALLINGER: Thank you, Chairman 22 Rempe. Today, we will have a discussion related to, 23 I guess I would call it the evolution of pressure 24 vessel embrittlement monitoring and prediction, and 25 the staff's work in this area and their proposals for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 1 a rulemaking to update, if you will, the process.
2 And we'll have presentations from Dave 3 Rudland, who's sitting in the front of us. And if you 4 would like, we will have him put his mask on, so we'll 5 be able to identify him.
6 (Laughter.)
7 MEMBER BALLINGER: And Elliot Long from 8 EPRI. So, with that, unless there are comments from 9 the members, Dave, it's all yours.
10 MR. RUDLAND: Thank you. Thank you.
11 First of all, I just want to say thanks to 12 the ACRS for allowing me to come here to talk about 13 these issues and the rulemaking plan that we put 14 together to address those.
15 And before we get started, I just wanted 16 to acknowledge the team that put this work together.
17 This has been an ongoing issue that we've been working 18 on, and it's been a cross-office issue that we've been 19 working on with NRR and Research, as well as NMSS, to 20 solve this problem. So, I wanted to just acknowledge 21 them and thank them for that.
22 And again, I want to thank you for having 23 this in person. I think is the first time in two 24 years that I've worn a tie. So, I'm glad to be able 25 to do that and to see all your faces in person.
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8 1 So, let me get started. I'm going to talk 2 about some issues today with reactor pressure vessel 3 embrittlement. We're going to focus a little bit on 4 the embrittlement trend curve that is in Regulatory 5 Guide 1.99, Rev 2, as well as the same trend curve 6 that's in 10 CFR 50.61. I'm also going to talk about 7 some issues with Appendix H surveillance testing. And 8 then I'll move into the staff's thoughts on how to 9 address those issues through rulemaking, as part of a 10 rulemaking plan that was delivered to the Commission 11 in March.
12 As a bit of a background, embrittlement 13 trend curves provide an estimate of the change in 14 fracture toughness. It's a regulatory requirement 15 that the licensees monitor the reactor pressure vessel 16 toughness. And so, we have embrittlement trend 17 curves that allow the right estimates for change in 18 fracture of toughness as a function of fluence.
19 In this illustration that's on the bottom 20 left, this red curve represents a schematic of one of 21 those, a prediction using that embrittlement trend 22 curve. And as you can see, the embrittlement 23 increases as a function of time and order of fluence.
24 And an embrittlement trend curve is just a 25 formulation, a curve fit to data that, in this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 1 particular case, this is a function of material 2 chemistry and the fluence level.
3 In addition, we have requirements for 4 surveillance capsule testing that provides monitoring 5 to ensure that this generic embrittlement trend curve 6 predicts the plant-specific behavior properly, so that 7 the data points on this schematic represent the data 8 capsules, materials that are pulled from these 9 capsules that are inside the plants and tested to make 10 a prediction of fluence. And again, embrittlement is 11 a function of these.
12 We then add a margin to that term, once we 13 have that. An embrittlement trend curve can be 14 adjusted to try to better represent the plant-specific 15 behavior. And then we add a margin to that, which 16 then gives us what's called the adjusted reference 17 temperature, which is then used in 10 CFR 50, Appendix 18 G, and 50.61, to give us both the pressure temperature 19 limits for a normal heatup and cooldown, as well as 20 screening criteria for PTS.
21 So, the ideal scenario is to have the 22 embrittlement trend curve, a margin that provides a 23 conservative prediction of embrittlement, and 24 surveillance data that covers all operating periods, 25 that allows us to verify that the plant-specific 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 1 behaviors are matching the trends that are provided in 2 the embrittlement trend curve.
3 However, if we have a trend curve behavior 4 that underpredicts the data, we can have some 5 uncertainty in what the real embrittlement is, and/or 6 if we don't have high fluence data, you can also have 7 an increased amount of uncertainty. And those 8 increased amounts of uncertainty can challenge the 9 confidence the staff has in ensuring that the plants 10 are safe from embrittlement.
11 CHAIRMAN REMPE: This is Chairman Rempe.
12 And I don't see any other mics in the room on except 13 David's and the ceiling, and I'm not hearing any other 14 noise.
15 MR. RUDLAND: I will move the microphone 16 a little closer. I don't know if that will help or 17 not. Okay, sure.
18 So, the staff's current perspective on 19 these issues -- before I go to the issues, I'm going 20 to tell you what our perspective is. It's not as if 21 I'm going to tell you the end before I tell you the 22 story.
23 The staff has high confidence that that 24 the current operating plant remain safe; that the 25 decisions that we've made in terms of embrittlement NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 1 and reactor pressure vessel safety still remain valid.
2 However, there is insufficient embrittlement 3 monitoring and underprediction of that embrittlement 4 that is in the current trend curves and in the current 5 regulations that will eventually impact the staff's 6 confidence in the integrity of the reactor pressure 7 vessel in long-term operation.
8 And by long term, you know, the first time 9 this may become an issue is if it was within about 10 10 years for pressurized thermal shock and much later, 11 about 20 years, for pressure temperature limits, due 12 to the expected fluence levels of the current 13 operating fleet. And as I'll show in a minute, both 14 the safety margins to brittle fracture and performance 15 monitoring will be impacted.
16 At this point again, like I've mentioned, 17 we've written a rule plan to the Commission, but 18 further work is really needed to determine which 19 plants are impacted by this potential issue, and I'll 20 talk about that.
21 MEMBER HALNON: Dave, this is Greg Halnon.
22 I'll know you'll probably get into this when you say 23 "a story," but the high confidence on the first bullet 24 is based on actual data trending with actual models 25 and other --
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12 1 MR. RUDLAND: That's correct. And where 2 the plants are right now in terms of (audio 3 interference).
4 MEMBER HALNON: Okay. So, when you say 5 it's going to be insufficient, that embrittlement 6 monitoring is down the road --
7 MR. RUDLAND: It's in the future.
8 MEMBER HALNON: -- it's past a certain 9 fluence level?
10 MR. RUDLAND: That's correct.
11 MEMBER HALNON: Okay. Thanks.
12 MR. RUDLAND: And I'll show some examples 13 of that. Let me start with the embrittlement trend 14 curves. The embrittlement trend curve that's 15 currently in Reg Guide 1.99, Rev 2, was published in 16 1988. It's a fit to the data that was available at 17 the time, which is about 177 surveillance data points.
18 Again, it's a function of the chemistry of the 19 material and the fluence level. That particular trend 20 curve, then, in 1991, was also included in 21 10 CFR 50.61 also. Recently, that embrittlement trend 22 curve was reevaluated for continued adequacy in 2014 23 and more detail in 2019, leading to today's effort.
24 So, this is an illustration of the issue.
25 In this plot, the vertical axis represents the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 1 difference between the embrittlement predicted by Reg 2 Guide 1.99 or 50.61 and the measured values from 3 surveillance data. So, a zero on the vertical axis 4 represents a perfect prediction. There is scatter in 5 the data. The solid-dashed lines represent the 6 standard deviation that's in the Regulatory Guide.
7 The horizontal axis is the fluence.
8 As you can see, the predictions are very 9 good until you get to a fluence of about 3 times 10 to 10 the 19th neutrons per centimeter squared. And then it 11 begins to deviate. And by the time you get to about 12 6 times 10 to the 19th neutrons per centimeter 13 squared, it become statistically significant, that 14 there is an underprediction. You end up with a 15 negative value on this chart, meaning that the 16 measured values are higher than the predicted values.
17 And what this 180 degrees means in terms of impact, 18 we'll talk about that in a few minutes.
19 The other issue is with surveillance 20 capsules. Appendix H to Part 50 requires periodic 21 monitoring. And it incorporates by reference an ASTM 22 standard that's a 1982 standard that was originally 23 designed for a 40-year life, but is not very specific 24 in what the lifetime needs to be. Typically, these 25 programs were set up with four or five capsules within NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 1 the reactor, such that, by the time they pulled the 2 second-to-the-last capsule, they had enough data to 3 give them the embrittlement data for a 40-year life.
4 So, the last capsule was always allowed to 5 be tested at two times the design fluence. This also 6 allowed the last capsule to be held without testing --
7 just a byproduct of the 1982 standard.
8 In 1997, the Commission, through some work 9 that was happening at the Perry Plant, came with a 10 finding that, decided that, if there was going to be 11 a change to the surveillance capsule withdrawal 12 schedule, that if the plant wanted to make a change to 13 the capsule withdrawal schedule, that only a 14 verification check needed to be done verifying that it 15 was in conformance with the ASTM spec, but not based 16 on technical or safety considerations.
17 And because of those two things, that 18 allowed some plants to repeatedly delay their last 19 capsule to acquire higher and higher fluences, as they 20 applied for license renewal and/or subsequent license 21 renewal. And here's an example of that.
22 This particular plant, again, on the 23 vertical axis is neutron fluence; on the horizontal 24 axis is the date on which a surveillance capsule was 25 pulled. So, in this case, there was four out of five NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 1 capsules were pulled. And you can see the last 2 capsule was pulled around in the early 2000s. The 3 fifth capsule was supposed to be pulled later that 4 decade. However, it was delayed multiple times to try 5 to reach an 80-year fluence, as this particular plant 6 applied for subsequent license renewal. And so, this 7 plant has no data at this particular point in this 8 high fluence area.
9 Many licensees have done this, and here's 10 some examples here on this plot that show plants that 11 have delayed this last capsule pull, but not all have.
12 Some have been following and periodically pulling the 13 capsules.
14 So, what's the impact of that? This shows 15 the same data that I just talked about, embrittlement 16 on the vertical axis; fluence on the horizontal axis.
17 The four data points that I talked about here in this 18 plot here; you see four data. If a draw a trend curve 19 from Reg Guide 1.99 using just the chemistry of the 20 material, I would get this particular curve.
21 Like I mentioned earlier, you can adjust 22 the trend curve to match, to best fit your data. So, 23 the blue represents the best fit through those four 24 data points.
25 Again, there is no data of high fluence NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 1 for this plant. If I assume I have high fluence data 2 that follows the trend on the prior slide, I would 3 have data that's way up here. I could have a 150-4 degree underprediction of embrittlement of that 5 particular data.
6 Again, I can fit that data. If I fit all 7 of those data points, I get the yellow curve, but I 8 still have a 75-degree underprediction of 9 embrittlement. And that's because of the functional 10 form of the embrittlement trend curve. If you look at 11 the functional form, again, it was fit to only about 12 170 data points, to a limited number of fluence, 13 limited level of fluence. When you get beyond that, 14 and you try to extrapolate that trend, it actually 15 reaches a peak, and then decreases, which is causing 16 this behavior that you see.
17 MEMBER BALLINGER: This is Ron. What 18 you're saying is you can't extrapolate?
19 MR. RUDLAND: You can't extrapolate.
20 MEMBER BALLINGER: It's just an empirical 21 fit --
22 MR. RUDLAND: That's right. That's 23 correct.
24 MEMBER BALLINGER: If you could put enough 25 dimensions on the fit, you can make it do anything you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 1 want?
2 MR. RUDLAND: That's right. That's right.
3 So, again, this particular plot was just 4 for the data that I showed prior where they had no 5 data. This is a recent plant data that we just got 6 surveillance data for. The same plot. The same 7 things. And you can see they pulled their fifth data 8 point and it came way up here. Relative to where they 9 are in terms of life, their 80-year fluence should be 10 here. So, this data point is actually greater than 11 their 80-year fluence. But, again, the point is you 12 can see that, even with trying to fit that data, I 13 still could have up to 60-degree underprediction in 14 embrittlement.
15 So, what's the safety case on this? The 16 staff did a very robust, risk-informed analysis where 17 we did a series of analyses using probabilistic 18 fracture mechanics, looking at a variety of different 19 scenarios, with the amount of data that we had.
20 Again, we had to try to use whatever we had, and we 21 didn't have enough to do individual plant-specific 22 analyses. So, we tried to do something that we 23 thought was maybe bounding.
24 We also looked at the other tenets of 25 risk-informed decisionmaking. We looked at safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
18 1 margins. We looked at performance monitoring. We 2 looked at difference in depth. So, I wanted to talk 3 about a few of those today.
4 So, again, the conditional probability of 5 failure that we calculated may increase by several 6 orders of magnitude, and I'll show that, an example of 7 that in a second. But, again, because that's a 8 conditional probability, the expected transient of the 9 -- the transient frequencies that we assumed in the 10 analyses are probably low. It's unknown, but they're 11 probably low.
12 In some cases, the plant may exceed their 13 pressurized thermal shock screening limit if you 14 correct this underprediction. But, again, the 15 analysis suggests that the risk is low, but a lot of 16 uncertainty exists and a lot of plant-specific details 17 are unknown.
18 So, the staff also focused on looking at 19 safety margins and performance monitoring. And these 20 predictions, when we looked at safety margins, the 21 underprediction, as well as the increase in 22 uncertainty, really impacts the safety margins. And 23 I'll show an illustration of that in a second also.
24 And as I mentioned, if you're not testing 25 the material, and then you have performance monitoring NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 1 issues delaying these capsules and not testing them at 2 the higher fluences, it does not provide adequate 3 performance monitoring to ensure that the trends are 4 reasonable.
5 So, some examples of the risk. This 6 particular plot, again, is conditional probability of 7 failure for a variety of different flaw sizes. The 8 horizontal axis is, basically, this underprediction, 9 how much underprediction that there is. If I have a 10 50-degree underprediction, I could have about two-11 three orders of magnitude change in the conditional 12 probability. If I have 150 degrees, I can up to six, 13 five-to-six orders of magnitude change in conditional 14 probability.
15 So, that seems really large, but, again, 16 the transients that are used in this particular case 17 are very low. That drives the overall probabilities, 18 or through all crack frequencies, to be relatively 19 low. But we don't know exactly what those frequencies 20 are. We don't know what the actual plant fluence 21 variations are, and we don't know whether these 22 analyses are really bounding. So, we didn't want to 23 base our case just on the risk of failure or pressure 24 temperature limits.
25 For pressurized normal shock, the analyses NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 1 were very similar. Again, if we correct for the 2 underprediction, it could be that some of these plants 3 may go over their screening criteria, which is 270 for 4 plates, forgings, and welds, axial welds, and 300 for 5 circumferential welds. But we took that into account 6 on our analyses and, again, calculated the through-7 wall crack frequencies for that. Most of those were 8 relatively low, suggesting that, even if they passed 9 the screening criteria, it's not a huge risk problem.
10 Now we looked at safety margins. The 11 uncertainties in the risk calculations are high, like 12 I mentioned, and increasing with time. There's a lot 13 of unknowns also. So, we wanted to try to take a look 14 at what the fundamental safety principles were in 15 terms of the basis for design and operation.
16 And again, the margins that I'll be 17 talking about really provided the reasonable assurance 18 against brittle fracturing, and I think those are most 19 impacted.
20 I can show that through this simple 21 illustration. This is a pressure temperature plot.
22 Pressure on the vertical axis; temperature on the 23 horizontal axis. The operating window is a typical 24 operating window for a plant that may be cooling down 25 from a higher temperature or higher pressure. They NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 1 stay within this window until they get to a low-2 pressure condition.
3 There is a structural limit out there that 4 can be calculated or estimated. If the plant cools 5 down faster than it decreases the pressure, it could 6 cross that limit and increase, significantly increase 7 the risk of embrittlement fracture.
8 So, what we do in the regulations is we 9 have an offset that has a particular margin on it, an 10 adequate margin, so that we regulate, then, this PT 11 limit. So that, when they're cooling down, they need 12 to stay to the right of these PT limits. And those 13 margins are dictated by the amount of uncertainty 14 that's in the analysis to calculate the structural 15 limit. So, more margin is needed for greater 16 uncertainty.
17 MEMBER BALLINGER: This is Ron. This 18 typical operating -- is that correct? I mean I've 19 operated a plant, and in that gray area down at low 20 temperature, that's a pump curve. And so, you go into 21 that; you can cavitate the pump. So, you're not in 22 the typical operating window here.
23 MR. RUDLAND: Well, this is the typical --
24 yes, the typical, this is the typical cooldown. I 25 guess maybe that's a better terminology.
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22 1 MEMBER BALLINGER: Because you're 2 actually --
3 MR. RUDLAND: A typical heating cooldown.
4 MEMBER BALLINGER: You operate actually 5 between the gray area and what now would be the orange 6 line. Because there's a box in the upper right which 7 is the actual operating regime for the plant, right?
8 MR. RUDLAND: You know, the operating 9 regime is up in this region. The plants have 10 administrative controls --
11 MEMBER BALLINGER: Yes.
12 MR. RUDLAND: -- to keep them away from --
13 MEMBER BALLINGER: Yes.
14 MR. RUDLAND: -- pressure temperature 15 limits. That's correct.
16 But there's no regulatory control over 17 those operations. So, what I'm talking about here is 18 just the regulatory control that we have over that.
19 With this underprediction --
20 VICE CHAIRMAN KIRCHNER: Isn't it, though, 21 Dave, true that the tech specs, you operate within 22 that, quote-unquote, "operating window," and that 23 provides regulatory control because you approve those 24 tech specs --
25 MR. RUDLAND: That's right. One of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 1 tech specs are these pressure temperature limits.
2 VICE CHAIRMAN KIRCHNER: Yes, exactly.
3 So, there a regulatory basis for stating within it, 4 even if it's not prescribed --
5 MR. RUDLAND: Right, but --
6 VICE CHAIRMAN KIRCHNER: -- in 10 CFR --
7 MR. RUDLAND: Yes, but the pressure-8 temperature curve is a tech spec limit.
9 VICE CHAIRMAN KIRCHNER: Yes.
10 MR. RUDLAND: And the fact is that it's 11 not in the right place because of this 12 underprediction. So, then, that's kind of the point 13 I'm trying to make here, is that, you know, for the 14 proper embrittlement prediction, this orange curve is 15 where PT curve should be, but because of the 16 underprediction, it's actually in this red area. And 17 so, we actually have a reduced margin to structural 18 failure because the plants believe that they're 19 following -- the plants are then instructed to follow 20 this PT limit that's kind of the reddish color here 21 because of that underprediction.
22 Then they can actually, you know, again, 23 if it wasn't for the administrative controls, they 24 could actually cool down in a region that's between 25 these orange and red curves. And that margin is, of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
24 1 course, reducing, and continues to reduce as the 2 uncertainty gets larger. So, the margin is going 3 down, but the uncertainty is going up. And so that's 4 why we have these --
5 MEMBER BALLINGER: This is Ron. I'll say 6 it again. That design, the typical operating window 7 should be moved up and to the left, and should be 8 between the gray or bluish-gray area and the orange --
9 MR. RUDLAND: Yes.
10 MEMBER BALLINGER: Wait a minute.
11 MR. RUDLAND: Really, I guess really what 12 the shaded area is, it's the typical cooldown --
13 MEMBER BALLINGER: Right, right.
14 MR. RUDLAND: Yes, you're right. You're 15 right. Are there any other questions?
16 VICE CHAIRMAN KIRCHNER: What is the 17 largest contributor to uncertainty in your opinion, 18 Dave?
19 MR. RUDLAND: I think the largest 20 contributor is not having enough data at high fluence 21 at this point, yes. I mean, again, as you look at the 22 amount of data points -- I'll send you back just real 23 quick to this plot here.
24 As you look to this plot, what I didn't 25 mention is that the red data points are U.S.
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25 1 surveillance data, and the gray data points are 2 international surveillance data.
3 So, you can see that there is not a great 4 amount of data in this region where the 5 underprediction occurs to really get a feel for what 6 the embrittlement is. So, I know I have an issue with 7 the embrittlement trend curve. That has some 8 uncertainty with it. I don't have any -- I have very 9 limited data at higher fluence, which also increases 10 that uncertainty.
11 MEMBER BALLINGER: We have limited data 12 for U.S. plants at high fluence.
13 MR. RUDLAND: That's right.
14 MEMBER BALLINGER: But that doesn't 15 invalidate the gray data.
16 MR. RUDLAND: Exactly right. That's 17 exactly right. That's right.
18 And again, in terms of any individual 19 plant, the largest uncertainty is the lack of -- you 20 know, if they have data or not at the high fluence 21 time. Because that really, then, is a measure of what 22 their actual embrittlement state is. So, if we don't 23 have the actual embrittlement state of any individual 24 reactor, then that uncertainty on whether that reactor 25 is nearing a limit or causing an issue is unknown; the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
26 1 uncertainty is large.
2 Okay. So, in quick summary, the current 3 state of knowledge, the generalized analyses that we 4 did, the probabilistic fracture analyses suggests that 5 the overall risk per fracture is low. But those 6 uncertainties that are associated with that analysis 7 are high and increasing with time, especially due to 8 the lack of surveillance data.
9 And so, safety margins can be impacted and 10 are decreasing as that uncertainty increases. And 11 delaying these capsules and not testing at high 12 fluence represents a lack of sufficient performance 13 monitoring.
14 I know Elliot will be talking about a 15 little bit more some of the ongoing industry programs 16 to try to increase the amount of data that we have at 17 high fluence. So, he'll be talking about some of 18 that. But, again, those data aren't expected for 19 years.
20 And these issues that I'm talking about 21 now really are mainly impacted by plants that are, or 22 will be, at higher fluence, which is, in this 23 particular case, about 6 times 10 to the 19th neutrons 24 per centimeters squared.
25 MEMBER HALNON: Dave, this is Greg. Do we 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
27 1 understand physically what's happening; why it's not 2 following what 1.99 is --
3 MR. RUDLAND: Yes, I think so. I think 4 Ron pointed it out. It's an empirical fit.
5 MEMBER HALNON: But, I mean, 6 materialistic, though, materials-wise, what is 7 occurring that's different than the predicted 8 microscale? I mean, what --
9 MR. RUDLAND: I don't think anything 10 different is happening. I think it's the same 11 behavior that was happening at lower fluence. I just 12 think our predictions don't extrapolate well because 13 we didn't have the data. So, we empirically fitted to 14 the data that we had. And then, when you try to 15 extrapolate that, it didn't work out very --
16 MEMBER BALLINGER: It's a simple case of 17 overfitting. In other words, you've got X data 18 points, and you do an n-parameter fit for X data 19 points. You can get a pretty darn good correlation 20 within that database. But when you get outside that 21 database, there's no physical -- there's no physics 22 anywhere.
23 MR. RUDLAND: Yes, yes.
24 MEMBER BALLINGER: And so, it just goes 25 nuts. That's exactly what it shows if you plot that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 1 curve.
2 MR. RUDLAND: Yes.
3 MEMBER BALLINGER: It just goes weird.
4 MR. RUDLAND: Yes.
5 MEMBER HALNON: I get the plot and I get 6 the math piece of it, but what's happening at an 7 atomic level that's causing the embrittlement, and 8 does that make sense in the higher fluences? Now that 9 we know that, should we have seen that back when we 10 first started seeing embrittlement, and the 11 materials --
12 MEMBER BALLINGER: But the trend is 13 actually consistent at higher --
14 MR. RUDLAND: Yes, yes.
15 MEMBER BALLINGER: So, it's not going 16 weird.
17 MR. RUDLAND: Yes.
18 MEMBER BALLINGER: The correlation is 19 going weird.
20 MR. RUDLAND: ASTM has taken all of the 21 worldwide embrittlement data and done another fit, 22 basically, using the same types of inputs, a couple of 23 other inputs, but, basically, the same kind of inputs, 24 and are able to predict all that data at the high 25 fluence.
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29 1 VICE CHAIRMAN KIRCHNER: But another way 2 of asking Greg's question, though, is from a materials 3 standpoint, first of all, it suggests to me that the 4 current curve, as you say, it's just an empirical fit 5 and there's no physics behind it. In other words, 6 it's fit empirically to data, but it doesn't 7 necessarily have -- well, it's a function of fluence, 8 I suppose, but only in the sense that it fit the data.
9 But the empirical -- the correlation doesn't have any 10 physics. That is my point.
11 MR. RUDLAND: Yes, there's a relationship 12 with copper and nickel, right, because they know those 13 are the things that impact embrittlement.
14 VICE CHAIRMAN KIRCHNER: Right. Right.
15 But now, we know enough about materials that, isn't 16 there a way, Ron, to build a model --
17 MR. RUDLAND: A mechanistic model?
18 VICE CHAIRMAN KIRCHNER: -- a mechanistic 19 model?
20 MR. RUDLAND: You know, research has 21 been -- I don't know how many years -- 30 years of 22 research; 40 years research has been going on, and 23 trying to be able to mechanistically model 24 embrittlement and the correlation between temperature 25 and fluence, and the material behavior has been a bit 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
30 1 elusive for the researchers.
2 MEMBER BALLINGER: Yes, and it's just your 3 PhD against my PhD.
4 (Laughter.)
5 MEMBER BALLINGER: Well, it's metallurgy.
6 So, that ought to tell you a lot.
7 VICE CHAIRMAN KIRCHNER: But, like in CHF 8 correlations, there are physical properties in the 9 correlation that make sense.
10 MEMBER BALLINGER: Somebody gives you 150 11 data points. They don't tell you where they come from 12 or anything. You go into Excel; you hit "Fit," and, 13 lo and behold, you'll get a fit.
14 (Laughter.)
15 MR. RUDLAND: I think it was a little bit 16 more sophisticated than that, but --
17 (Laughter.)
18 MEMBER BALLINGER: But, well, it was 19 pretty close to that. But then be careful going 20 outside that range because it just goes crazy.
21 MR. RUDLAND: And again, I think the ASTM 22 group that I just talked about spent some time trying 23 to look at really what the drivers are as best they 24 could, based on the physics, as we know it. So, there 25 are other parameters in the newer fits that aren't in 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 1 the Reg Guide 1.99.
2 MEMBER BALLINGER: And by the way, NRC 3 staff participated in that analysis. Am I correct?
4 MR. RUDLAND: Yes. Oh, yes.
5 MEMBER BALLINGER: So, the NRC staff has 6 been up to their ears --
7 MR. RUDLAND: And I just saw a comment 8 come in in the chat, which I can't see right now, from 9 Dan Widrevitz, who I know is on the ASTM committee 10 that is working on that stuff. So, we do have staff 11 on it.
12 MEMBER MARCH-LEUBA: Please don't share 13 technical information in the chat because it is not 14 recorded in the transcript.
15 MR. RUDLAND: So, Dan, if you wanted to 16 say something, I guess you have the opportunity to do 17 that.
18 MR. WIDREVITZ: Sure. I just wanted to 19 mention that there were some mechanistic insights 20 incorporated into the varieties of fits. And so, this 21 isn't a totally pure "let's throw an algorithm at our 22 data methodology," even back for Reg Guide 1.99. It's 23 just the issue is the particular methodology and the 24 data that they had at the time are not extrapolating 25 well now. But there's always been a certain level of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 1 mechanistic insight in all of these trend curves.
2 MEMBER BALLINGER: But, in the end, as 3 soon as you put an adjusting parameter on it, it 4 becomes an empirical fit.
5 MR. RUDLAND: Agreed.
6 MEMBER BIER: I could have probably done 7 it, despite knowing nothing about materials.
8 (Laughter.)
9 MEMBER BALLINGER: You would have probably 10 done a better job.
11 MR. RUDLAND: Okay. So, I want to talk a 12 little bit about who's impacted by this. And again, 13 like I mentioned earlier, yes, the embrittlement 14 underprediction is a function of fluence. And it's 15 for the thing that, the term put this in that it's in 16 the regulations and in our guidance. It becomes an 17 issue when you get to this fluence level about 6 times 18 10 to the 19th neutrons per centimeter squared.
19 And so, again, it's mainly a PWR issue, 20 because the BWRs will never get to that, at least in 21 the expected 80 years of life for those plants. So, 22 about 34 percent of PWRs may be impacted.
23 But, again, it's important to point out 24 that a lot of plant-specific details are really needed 25 to determine which, if any, one individual plant would NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 1 be affected by this or not. So, the staff needs to do 2 a little bit more work to determine how to focus these 3 on plants that are impacted.
4 And, of course, anybody who delays their 5 last capsule would be impacted by this lack of 6 surveillance data. And that's true for license 7 renewal and subsequent license renewal.
8 VICE CHAIRMAN KIRCHNER: Now, Dave, you 9 picked 8 times 10 to the 19th to display. Could you 10 explain --
11 MR. RUDLAND: Yes, I just --
12 VICE CHAIRMAN KIRCHNER: -- how you came 13 up with 8 times 10 to the 19th? Because I would think 14 you would go -- I mean, at 6 times 10 to the 19th, you 15 see a divergence. What's so magical about 8 rather 16 than --
17 MR. RUDLAND: No, there isn't any --
18 VICE CHAIRMAN KIRCHNER: -- 10 to the 19 20th?
20 MR. RUDLAND: Yes, I showed it just for 21 illustration on how fast it's dropping off in terms of 22 the number of plants that are impacted. I'm not sure 23 off the top of my head what the expected highest 24 fluence is of the plants that are coming in, but it is 25 around, you know, 20 to the 20th, right? So, this is 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
34 1 halfway in between those in 6 and in 8 --
2 MEMBER BALLINGER: Are you going to say 3 something about where this fluence is measured in 4 relation to where the calculations require?
5 MR. RUDLAND: Well, and I did briefly at 6 the beginning. So, again, this particular number 7 represents an ID fluence. So, this is the fluence 8 that's at the ID of the vessel. So, when you're doing 9 a PTS calculation, you use the fluence value that's at 10 the ID of the vessel. When you're doing a PT 11 calculation, you use the fluence that's at the 1/4T 12 and 3/4T locations.
13 MEMBER BALLINGER: So, at those, it's much 14 less than that?
15 MR. RUDLAND: That's right. Since it 16 attenuates through the wall, it's much less at a 1/4T, 17 which is why at the beginning I said it becomes to be 18 a problem about 10 years for PTS and 20-something 19 years for PT.
20 MEMBER HALNON: Is there a general feeling 21 what fluence level would become so operationally 22 restrictive that they won't be able to operate these 23 plants?
24 MR. RUDLAND: I think that, as long as we 25 fix things and they change their PT curves and follow NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 1 that, you know, that's a plant-specific issue on when 2 they think it's not enough operating space for them to 3 operate in, where they will need to do something 4 (audio interference), or whatever other mitigative 5 techniques they would have.
6 MEMBER HALNON: Okay, but you don't have 7 a feel --
8 MR. RUDLAND: I don't. No, I don't have 9 a feel. Again, that's a plant-specific problem.
10 All right. So, to solve this, the staff 11 came up with some alternatives that we wrote up in our 12 rulemaking plan to change this. And the first 13 alternative is, basically, to do nothing, to make no 14 changes, but to handle this issue on a plant-specific 15 basis, either by orders or by something within their 16 license renewal applications, or by some generic 17 communications. So, that's the first alternative.
18 I'm going to talk about the third one 19 first, next. The third alternative is a comprehensive 20 solution where we replace the embrittlement trend 21 curves in the regulations, as well as in the guidance, 22 to a new trend curve that is better suited for the 23 expected periods of operation, as well as to include 24 some additional requirements in Appendix H for 25 additional surveillance testing and long-term 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 1 operation.
2 And then the second alternative is a much 3 more focused alternative, where we're only going to 4 make the changes and requirements effective for those 5 plants that may be impacted. So, it's a much more 6 focused solution.
7 And how we do that is something that staff 8 still needs to figure out, but we figure that -- yes.
9 So, I'll talk about what we decided to do here in a 10 second.
11 No matter what we do, it's going to be a 12 backfit. So, if we do it on a plant-specific basis, 13 there will be individual backfits. If we do it in 14 terms of rulemaking, it will be one backfit. But, 15 either way, it's going to have to be a fact that we'll 16 have to go through the proper analysis to justify the 17 backfit.
18 And again, modification of H will require 19 some plants to test a capsule during SLR period, that 20 they maybe weren't planning to, or sometime other, 21 maybe in license renewal at their (audio 22 interference). And then the modification to the trend 23 curve may cause some plants to modify their PT limits 24 and/or PTS calculations.
25 But, again, for at least alternative two, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 1 we want to make sure that we only focus on those 2 plants that are highly affected.
3 MEMBER BALLINGER: Is there a likelihood 4 that using a more sophisticated probabilistic 5 analysis, some of this "hit," if you will, could be 6 avoided?
7 MR. RUDLAND: It's possible. Again, it's 8 not the analysis sophistication that is needed; it's 9 the inputs, right? So, we need plant-specific inputs.
10 That's really what it comes down to.
11 So, the staff recommended alternative two, 12 which, again, is a focused solution. It's the one 13 that would be the least resource-intense and affect 14 the plants that are highly impacted without impacting 15 those that at this point don't need it. Like I 16 mentioned earlier, the BWRs don't have an 17 underprediction problem, and they won't for 80 years.
18 And so, it's possible that we can work the guidance 19 and regulations to exclude those from many changes 20 until some point where they get to those higher 21 fluence levels. But how we're going to do that needs 22 to be worked out during the regulatory basis, because 23 it's going to take some investigation into plant-24 specific details and other things to be able to do it.
25 I'll quickly talk about the schedule that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
38 1 we had in the rule plan. Like I mentioned, we have 2 some work to do in the regulatory basis development.
3 So, we actually have about 14 months to do that, once 4 we receive an SRM from the Commission, and after that, 5 developing the proposed rule and final rule are both 6 about the same time, about 15 months each. And that 7 puts us at about 2027. If we get an SRM sometime a 8 little bit later this year, that puts at 2027 before 9 we issue the final rule.
10 MEMBER HALNON: How much time do we have 11 before the first plant would reach the fluence of 12 concern?
13 MR. RUDLAND: Like we talked about, for 14 the ID, it's about 10 years.
15 MEMBER HALNON: So, this has got a lot of 16 margin to it?
17 MR. RUDLAND: It's got some margin to it.
18 MEMBER HALNON: Okay.
19 MR. RUDLAND: That's why we decided to do 20 this now instead of waiting until it was an issue.
21 MEMBER BALLINGER: But, to be clear, it's 22 really an issue of the effect of uncertainty on the 23 predictions?
24 MR. RUDLAND: That's correct.
25 MEMBER BALLINGER: It has nothing to do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 1 with the actual performance of the vessel?
2 MR. RUDLAND: What it has to do is it has 3 to make sure that they're operating to the right PT 4 curves.
5 MEMBER BALLINGER: Right.
6 MR. RUDLAND: Yes.
7 MEMBER BALLINGER: Yes, there's a mean 8 that's going on there, but then there's this big 9 uncertainty band which you showed.
10 MR. RUDLAND: Yes.
11 MEMBER BALLINGER: And that's the thing 12 that's getting us --
13 MEMBER HALNON: But, I mean, that's coming 14 from the science piece; I'm coming from the operations 15 piece. When does my PT curve change to the point --
16 MR. RUDLAND: That it becomes a problem.
17 MEMBER HALNON: -- that it becomes a 18 problem, and I have to change my operating procedures 19 and my simulator exams, and everything else that I do?
20 So, I'm looking at the operations piece.
21 MEMBER BALLINGER: They have to do a PT 22 curve every startup, right?
23 MR. RUDLAND: They have to do a PT curve 24 anytime something changes. So, if they change their 25 fluence levels. A lot of times, their PT curves are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
40 1 out farther in EFPY than they're operating, basically.
2 MEMBER HALNON: Right, and that's what I'm 3 talking about -- when the PT curve gets so close to 4 the operating curve that I have an issue with my 5 training, and everything else that goes on.
6 MR. RUDLAND: And to your point, Ron, the 7 scatter that's in the first part of that is handled by 8 margins. We put a margin term on there to handle 9 that. But it's when it becomes non-conservative that 10 it is just more than our margin.
11 Okay. So, in summary, we have high 12 confidence that right now the plant remains safe and 13 the licensing actions remain valid, because this issue 14 is really a longer-term, time-dependent issue. And 15 like I just mentioned, it's about 10 years before we 16 begin to have an issue for PTS and about 23 years 17 before PT limits.
18 And then that really just impacts our 19 confidence in the integrity of the vessel for long-20 term operation. But we need to do further work, and 21 we want to be proactive to ensure that reasonable 22 assurance is still maintained. And we're going to do 23 that through a risk-informed, performance-based 24 solution.
25 Again, we delivered the rulemaking plan to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 1 the Commission in March. And we want to develop a 2 change to the rule that is focused on only those 3 plants that are impacted.
4 I think that's all I have.
5 MEMBER BALLINGER: I'm having trouble 6 finding the clock. Oh, there it is.
7 Great.
8 Are there any questions from the members 9 before we switch? I don't think it's late enough to 10 have a break. So, should we just push forward?
11 VICE CHAIRMAN KIRCHNER: Ron, I have a 12 question, if I may?
13 MEMBER BALLINGER: Uh-hum.
14 VICE CHAIRMAN KIRCHNER: Dave, is there 15 any -- you know, this is an area where there's been a 16 lot of work done by the industry and the staff.
17 There's a lot of regulatory history invested in this, 18 or effort I should say, not history.
19 Is there active research over in RES 20 looking at this matter and the falloff when you get to 21 beyond 6 times 10 to the 19th fluence?
22 MR. RUDLAND: I think the only thing 23 that's really active that's going on is our 24 involvement into Code standards, and allowing those 25 types of things to be developed by a wide variety of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
42 1 technical experts. So, it's the ASTM standard, as 2 well as in ASME to developing standards related to the 3 prediction of fluence.
4 And so, we're relying -- you know, we 5 weren't being involved in those, but we're relying on 6 those experts as well as our experts.
7 VICE CHAIRMAN KIRCHNER: I didn't phrase 8 my question well. Yes, the standards work is 9 important, and that's the codification, an accepted 10 standard showing this phenomenon informed with more 11 data. Okay, I get that. You update the standard.
12 But I meant more from a real research 13 standpoint.
14 MR. RUDLAND: Right, phenomenological, 15 working on a phenomenological problem? No, there's 16 not right now, no.
17 VICE CHAIRMAN KIRCHNER: Maybe that's 18 something for research engineering, when we have our 19 presentations from RES, we could inquire about.
20 MEMBER BALLINGER: The problem is it's 21 just -- now this is a personal opinion -- it just 22 isn't needed. I mean, if there was true uncertainty, 23 then that would be one thing. But the general trend 24 plus margin is fine.
25 And the research, you know, extracting 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
43 1 samples from vessels that were retired or, you know, 2 all of this taking samples is extremely expensive.
3 And those vessels, when they were put in service, 4 there was not a whole lot of knowledge about the 5 vessels themselves. And so, you have to sort of 6 divine what those conditions were in the vessel to 7 start with in order to make sense of the data that you 8 get from doing Charpy tests or fractured toughness 9 tests. So, it becomes very expensive with minimal 10 payoff.
11 VICE CHAIRMAN KIRCHNER: What about the 12 delayed capsule testing?
13 MEMBER BALLINGER: Well, they're going to 14 get those anyway.
15 MR. RUDLAND: Yes. And I think EPRI is 16 going to talk about some of that here.
17 VICE CHAIRMAN KIRCHNER: Okay. So, there 18 is an effort, though, then, to --
19 MR. RUDLAND: There's an EPRI effort.
20 VICE CHAIRMAN KIRCHNER: -- inform, better 21 inform, the space that we're looking at?
22 MR. RUDLAND: Yes, I think Elliot is going 23 to talk about that.
24 VICE CHAIRMAN KIRCHNER: Thank you.
25 MEMBER BALLINGER: If some of these NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 1 capsules were to come out and, all of a sudden, 2 there's, you know, something really weird, then they 3 would have to figure out; that may be something 4 different.
5 MR. RUDLAND: Researchers have been 6 chasing mechanisms that cause an increase in 7 embrittlement for years. I mean, there's been a lot 8 of talk of late-blooming phases that occur that cause 9 the embrittlement. And they've been chasing that for 10 years without much success.
11 MEMBER BALLINGER: But it put a lot of 12 graduate students through school, especially at Santa 13 Barbara.
14 (Laughter.)
15 MEMBER BALLINGER: Okay, anybody else?
16 (No response.)
17 MEMBER BALLINGER: Then again, I ask the 18 same question: it's too early for a break?
19 CHAIRMAN REMPE: I think you have another 20 presentation that's coming up.
21 MEMBER BALLINGER: Yes.
22 CHAIRMAN REMPE: And so, I think it's a 23 little early for a break.
24 MEMBER BALLINGER: Yes.
25 CHAIRMAN REMPE: Let's go ahead and --
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45 1 MEMBER BALLINGER: Okay. So, Elliot, are 2 you up and ready? Whoops, here we go. Shoot.
3 Elliot, are you up and ready?
4 MR. LONG: Elliot Long speaking. Can 5 everyone hear me?
6 MEMBER BALLINGER: Good. The floor is 7 yours.
8 MR. LONG: Okay. Chris or Dave, can you 9 share my slides? Or do you want me to share myself?
10 CHAIRMAN REMPE: Chris, are you doing it?
11 MR. BROWN: Elliot, if you want to, you 12 can do it. If not, I can do it.
13 MR. LONG: All right. Let me see if I can 14 pull this off.
15 MEMBER BALLINGER: We see it now.
16 CHAIRMAN REMPE: Elliot, yes, if you could 17 switch to presentation, that would be great. Thank 18 you.
19 MEMBER BALLINGER: Okay, let's go.
20 (Pause.)
21 MEMBER BALLINGER: I think somebody's on 22 mute or --
23 CHAIRMAN REMPE: I don't know if Elliot's 24 on mute. Elliot, we see your hand up and we've lost 25 your slides, and we can't hear you. Oh, it muted you.
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46 1 MR. BROWN: I can also share it.
2 (Pause.)
3 CHAIRMAN REMPE: You can't unmute?
4 Oh, can you share, Chris? I can lower 5 your hand, but I can't unmute you, from what I can 6 tell.
7 (Audio interference.)
8 CHAIRMAN REMPE: Oh, you're coming through 9 a phone, your sound is. I was looking at your 10 presentation.
11 MR. BROWN: You might have to use star-6 12 to unmute.
13 CHAIRMAN REMPE: Yes.
14 MR. LONG: When I shared my slides, it 15 wouldn't let me unmute. I muted and I couldn't 16 unmute. I apologize.
17 CHAIRMAN REMPE: Okay. Are you using 18 star-6 on your phone line, if you're coming through a 19 phone line?
20 MR. LONG: Yes, and that's when I heard 21 the disembodied voice that I wasn't allowed to unmute.
22 CHAIRMAN REMPE: Oh, okay.
23 MR. LONG: And that's after I shared my 24 slides. It was good before.
25 CHAIRMAN REMPE: Chris Brown is going to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 1 share his slides. So, just give us a minute here, and 2 we'll hook you up.
3 MR. LONG: Okay. Yes, sorry about that 4 technical difficulty.
5 (Pause.)
6 MR. BROWN: Elliot, I have your slides 7 here.
8 MR. LONG: This is good. Everyone ready 9 again?
10 CHAIRMAN REMPE: Yes, we are.
11 MR. LONG: Okay. I will start over.
12 Okay. I am Elliot Long with the Electric Power 13 Research Institute. I am the current Technical Leader 14 for Reactor Pressure Vessel Integrity and Low Alloy 15 Steel Research.
16 I'm going to speak today about our RPV 17 embrittlement monitoring and prediction in long-term 18 operation and some of the current programs that EPRI 19 has today.
20 Next slide. We have two current, ongoing, 21 long-term programs to generate high fluence capsule 22 data for the domestic fleet. The first is the PWR 23 Coordinated Reactor Vessel Surveillance Program, 24 CRVSP, and the second is the PWR Supplemental 25 Surveillance Program, PSSP. And I'm going to talk 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
48 1 about both of those today.
2 And then, afterwards, I want to review 3 some of the prior EPRI conclusions from the November 4 2019 ACRS meeting, and then take a look at the 5 potential impact to PT limit curves.
6 Next slide, please. So, we'll start off 7 with CRVSP discussion.
8 Next slide. As I mentioned, the goal of 9 this project is to optimize the current U.S. PWR 10 surveillance capsule withdrawal schedule to increase 11 the amount of high fluence data available to the 12 industry. This data can then be used to inform future 13 embrittlement trend curve correlations with applicable 14 fluence changes for 60 years and beyond.
15 The original revision of this report was 16 completed in 2011, and it reviewed every reactor 17 vessel surveillance program from the U.S. fleet, and 18 then recommended changes for certain plans to select 19 withdrawal schedules for the capsules, to increase the 20 amount of high fluence data by the selected date of 21 2025.
22 Last year in 2021, EPRI MRP updated this 23 report to review how we did; what has happened; what's 24 left to do, and when will we have that data available 25 across the fleet.
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49 1 This update included a review of all 2 capsules withdrawn since 2011. It looked at future 3 capsule pull schedules. It updates any capsule 4 fluence values, and we also analyzed and addressed any 5 plants that closed or are planned to close in the near 6 term.
7 Next slide. The results of this program 8 to date are that 16 of the 30 CRVSP capsules are 9 tested or plan to be tested. The remaining 14 10 capsules -- there are 14 remaining capsules -- half of 11 these will not be tested or are not planned to be 12 tested, mainly due to plant shutdowns, or they will be 13 delayed beyond 2025. So, we'll still get the data; it 14 will just take a little bit longer.
15 In summary, for high fluence data that is 16 currently available, there are 48 capsules with a test 17 of fluence greater than 3 times 10 to the 19th. Four 18 of these are greater than 8 times 10 to the 19th. By 19 2025, when the remaining seven planned capsules will 20 be tested, all of them will be over 3 and two of those 21 will be over 8 times 10 to the 19th.
22 Any questions on the CRVSP program before 23 we move on to the PSSP?
24 (No response.)
25 MR. LONG: Okay. Next slide. PWR 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
50 1 Supplemental Surveillance Program, PSSP. This is a 2 similar program to the CRVSP. The goal is to generate 3 additional high fluence surveillance data, again, to 4 inform development of embrittlement trend curves 5 applicable to high fluence or long-term operation.
6 The overall objections are to fill in gaps 7 in the tested surveillance capsule database and to 8 utilize commercial PWR data to inform future 9 embrittlement trend curve and not to need to rely on 10 test reactor data.
11 The overall project goal is to irradiate 12 two supplemental surveillance capsules for 13 approximately 10 years before withdrawal testing, 14 evaluation, and publication of the reports.
15 These two supplemental capsules have an 16 additional 288 Charpy specimens from 27 unique 17 materials across the three main material types --
18 plates, forgings, and welds.
19 The data generated from these capsules 20 will yield 24 new transition temperature shift values 21 and, also, three additional upper-shelf energy 22 results. The fluence range on the specimens is 4.5E19 23 up to 1.2E20. So, well into the future for fluence, 24 very high fluence.
25 So, this slide documents some of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 1 history of this program. The program designed and 2 fabricated two supplemental capsules containing 3 previously-irradiated and reconstituted PWR materials.
4 These two surveillance capsules were inserted into 5 Farley Unit 1 in October of 2016, and then Shearon 6 Harris in April of 2018. The fabrication report was 7 published in 2016, at the right.
8 Let me describe, first, the images. So, 9 the image at the right shows how you make a 10 reconstituted Charpy specimen. The top figure, broken 11 Charpy, is after testing of a Charpy, you end up with 12 two halves, of course. One of the halves is machined 13 into the specimen insert with end tabs, which are then 14 welded -- in the middle image there. The welded and 15 reconstituted specimen is then machined square, and 16 then cut to length at the bottom. And that bottom, 17 the fifth picture there, is then inserted into the new 18 reconstituted capsule.
19 The current plan is to withdraw Farley 20 Unit 1 in the spring of 2027 and the Shearon Harris 21 capsule in the fall of 2028. Those schedules were 22 revised in the draft as part of the work done for 23 MRP-326, Revision 1.
24 The testing of the capsules will then 25 commence through 2028 to 2030. The capsule reports NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 1 will be delivered approximately 18 months after the 2 withdrawal date in two separate reports. And then, 3 after the data is completed and compiled, we will 4 analyze the impact, 2030 to 2032.
5 That bottom right picture is the image of 6 one of the capsules. That's the Shearon Harris 7 capsule seated in the capsule holder inside the 8 reactor vessel, and you're looking downward at the top 9 end plug of the surveillance capsule.
10 Any questions on the PSSP program before 11 we move forward?
12 (No response.)
13 MR. LONG: Okay. I now want to review 14 some of the prior EPRI conclusions on the potential 15 revision of Reg Guide 1.99, R2. In November of 2019, 16 EPRI made an original presentation to the ACRS, 17 discussing the potential revision of the Reg Guide.
18 The conclusions from that meeting -- that are on the 19 right -- have not changed.
20 If, in the future, we were to revise this 21 Reg Guide, the AFPM E900-15 model remains the 22 preferred embrittlement trend curve today.
23 As Dave mentioned a few times in his 24 presentation, below a fluence target of about 16-19, 25 the current Reg Guide remains adequate. It's above NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
53 1 that fluence threshold where it begins to break down.
2 Consistent with that idea, no BWRs will 3 reach that threshold. And when we're looking at PT 4 limit terms, we're really focused on the 1/4T limit.
5 So, what fluence will reach the 1/4T fluence of 16E19?
6 So, on the next slide, I detail how much 7 surface fluence you would have to have before your 8 1/4T or 3/4T influence is equal to 16E19 threshold.
9 So, to do this, I utilize the attenuation formula from 10 the current Reg Guide.
11 Again, I want to note that BWR units will 12 not reach a fluence of 16E19. In fact, the highest 13 1/4T SLR fluence is actually less than 1E19 at 80 14 years for the U.S. BWR fleet.
15 But the story is a little different for 16 the PWRs. The table at the top right shows what 17 surface fluence you need to reach 16E19 at 1/4T or 18 3/4T.
19 So, let's do an example. For the 20 Westinghouse Four-Loop plant with a B&W vessel, the 21 thicknesses are approximately 8.5 inches and you need 22 a surface fluence of 9.99 on 10 to the 19th. You have 23 a corresponding 1/4T fluence of 6 before the Reg Guide 24 begins to break down.
25 So, looking at this chart, it is observed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 1 and understood that the Three-Loop Westinghouse design 2 has the highest surface fluence at end of life. We 3 can also understand that the 3/4T fluence values are 4 well beyond anything we would expect at the surface.
5 For example, the larger Westinghouse Four-Loop with 6 the CE vessel needs a surface fluence of 28.3 times 10 7 to the 20 to have a 3/4T fluence of 16E19.
8 So, I look next at the surface fluence 9 values for the six plants that have submitted SLR 10 applications. That chart is shown on the bottom 11 right. For five out of six, the surface fluence 12 values is below the value needed to have 1/4T fluence 13 of 16E19. For example, Plant C, the surface fluence 14 is 5.56, and that value is below the value necessary 15 of 9.62 to have an issue for PT limits through 80 16 years of operation.
17 Only one plant will ever hit a 1/4T 18 fluence of 16E19 for 80 years. That's predicted to 19 occur well into the SLR operating period, and I 20 believe, consistent with Dave, it will be about 23 21 years into the future. So, that 23 years is the first 22 time where the PT limit for the first plant, the lead 23 plant, would be impacted by a potential change to the 24 Regulatory Guide.
25 Any questions on that?
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55 1 (No response.)
2 MR. LONG: Thank you. That is the end of 3 my presentation.
4 (Pause.)
5 CHAIRMAN REMPE: Okay, try again.
6 MEMBER BROWN: Yes, Charlie Brown. I 7 think I asked this question last time; I've forgotten 8 the answer. One of them was you emphasized, for the 9 PWRs, that you wanted to get away from test reactor 10 data and get actual plant data. Is that a cost issue 11 or is it a materials issue that you want to be more 12 conforming with the actual materials that are in the 13 plants? That's question one.
14 MR. LONG: Okay. We would like to be as 15 close to conforming with the actual material in the 16 plants as we can. All of the materials in the PSSP 17 program were originally irradiated in a commercial, 18 domestic U.S. reactor. The Charpys were tested, and 19 then reconstituted into new Charpys, and they'll get 20 a higher fluence data point when they are removed and 21 tested from their host plant in the PSSP program.
22 MR. RUDLAND: And the data suggest that 23 there's no -- sorry -- the data suggests that there's 24 a flux event.
25 MEMBER BROWN: So, there is a difference NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 1 between test reactor data and in plant --
2 MR. RUDLAND: Yes.
3 MEMBER BROWN: Okay.
4 MR. RUDLAND: It may be able to be 5 corrected, right? But I think in order to be true to 6 that, what Elliot said, it's better to use the actual 7 surveillance data.
8 MEMBER BROWN: Okay. My second question 9 was on the Charpy specimens. When you break it, and 10 then you cut it, I mean, it's broken, so you have to 11 machine it to reconstitute it. And I think I asked 12 this question. Are they big enough that the machining 13 doesn't detract from your ability to get the data you 14 want?
15 MR. LONG: Yes. There are ASTM standards 16 that we complied with to meet the reconstitution 17 dimension, and these specimen passed those criteria.
18 So, yes.
19 MEMBER BROWN: Okay. All right. Thank 20 you. Now that you said that, I remember that answer.
21 MR. RUDLAND: They put the notch in areas 22 that are far enough away from the cut edges --
23 MEMBER BROWN: Yes, I knew that was an 24 issue, just from past experience.
25 VICE CHAIRMAN KIRCHNER: And then do they 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
57 1 reanneal them and everything else to --
2 MEMBER BROWN: They shouldn't do that.
3 MR. RUDLAND: No. No, because they want 4 more fluence --
5 VICE CHAIRMAN KIRCHNER: Okay. So, you're 6 just accumulating fluence on an existing sample?
7 Thank you.
8 MEMBER BROWN: Thank you very much.
9 MR. LONG: You're welcome.
10 MEMBER BALLINGER: Other questions?
11 (No response.)
12 MEMBER BALLINGER: Okay. I think on the 13 schedule we have an opportunity for public comments.
14 So, in keeping with that, if there are any members of 15 the public that would like to make a comment, I think, 16 since you have a Teams invitation, it should be easy 17 to do. Please identify yourself and make your 18 comment.
19 (No response.)
20 MEMBER BALLINGER: If you're on a phone 21 call, I think you have to do the star-6 thing still, 22 right? Yes.
23 (No response.)
24 MEMBER BALLINGER: No? The five-second 25 rule. Hearing no public comments, thank you very NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 1 much, both of you, for the presentations and, by the 2 way, for the next years of work that's been going into 3 this.
4 And I'll turn it over to the Chair.
5 CHAIRMAN REMPE: Thank you. So, at this 6 point, we're going to go off the record and we are 7 going to take a break. Is 10 minutes enough, folks, 8 or 15 minutes? Okay, so let's do a 15-minute break, 9 and we're going to restart at 9:55.
10 And Ron has a draft letter.
11 Oh, you want to do a 20-minute break? I'm 12 looking at the computer clock, okay, Jose?
13 MEMBER MARCH-LEUBA: Yes.
14 CHAIRMAN REMPE: Okay, yes. So, at 9:55, 15 we're going to restart, based on the computer clock, 16 not the clock in the room.
17 And we'll listen to Ron reading his draft 18 letter.
19 And at that point, we're going to recess, 20 folks.
21 We'll be coming back on the record at 22 12:30 p.m. for the BWRX-300 discussion. Thank you.
23 (Whereupon, the above-entitled matter went 24 off the record at 9:40 a.m. and resumed at 12:30 25 p.m.)
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59 1 CHAIRMAN REMPE: Okay, the clock on the 2 computer says it's 12:30 p.m. on the East Coast, and 3 we're back in session.
4 And at this point, I would like to ask 5 Member March-Leuba to lead us through our second topic 6 today on BWRX-300's Containment Evaluation Topical 7 Report.
8 MEMBER MARCH-LEUBA: Thank you, Member 9 Rempe.
10 I assume we're going to be listening to 11 our General Electric Hitachi Topical Report on 12 BWRX-300 Evaluation Method for the Containment.
13 So, without much more ado, we're going to 14 let the staff make introductory remarks.
15 Mike Dudek?
16 MR. DUDEK: Thank you.
17 Technical Session Chair Mark Dudek, and 18 I'd like to once again thank Chairman Rempe and the 19 rest of the Committee for this opportunity to present 20 the staff's findings associated with the BWRX-300 21 Topical Report entitled, "Containment Evaluation 22 Methodology."
23 We presented to the Subcommittee on this 24 Topical Report on March 18th; just outstanding 25 discussion associated with those efforts, a lot of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
60 1 good questions, a lot of good answers. And the 2 staff's back today to address some of the concerns 3 that were identified by the Committee and to give 4 another high-level overview of the Topical Report.
5 And really, at the end of the day, you know, the staff 6 just wants to present its findings associated with the 7 Topical Report and the methodology on this methodology 8 for containment thermal hydraulic performance.
9 This was a year-long effort by the staff, 10 and really, it involved the staff taking a look at the 11 containment evaluation method through the TRACG, the 12 Transient Reactor Analysis Code, and then the reactor 13 pressure vessel and the containment through the GOTHIC 14 code. There were several RAIs. There was an audit.
15 And as I said, this was a year-long effort to verify 16 the acceptance criteria anywhere from AOOs, loss of AC 17 power, ATWS, small break LOCAs, and large break LOCAs.
18 Tangentially to this SER being approved by 19 the staff in early January, we've been working in 20 parallel on just an outstanding effort with the 21 Canadians to do a joint review on this Topical Report.
22 And really, the purpose of that joint review, we had 23 an MOU with Canada to conduct it. We embedded staff 24 with each one of the organizations for continuous 25 learning. Both regulators evaluated how the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
61 1 performance would be used in each other's regulatory 2 infrastructures, our pre-application engagements, and 3 during our application reviews. We had mutual 4 understanding. We had several workshops to kick off 5 this initiative in which each regulator presented its 6 regulatory infrastructure and how we do business.
7 So, to use an analogy that I've been 8 using, we learned how to dance. We learned how each 9 other operates. We learned how each other reviews.
10 We learned how to conduct a joint review. It was just 11 an outstanding effort with Canada.
12 And at the end of the day, we had four 13 mutual learnings. We had four joint conclusions and 14 an observation that we had. The learnings for this 15 report can be found in ML22091A201.
16 Now what's important, also, about this 17 joint review is that nothing in the report fetters the 18 powers of either CNSC or NRC, and nothing in the 19 report can be construed or interpreted as affecting 20 the jurisdictions or discretion of either regulator.
21 There are no new learnings identified in the joint 22 report, but there is a fulsome description of exactly 23 how the staffs interacted, exactly what they reviewed, 24 and how they did it. Just an incredible effort.
25 So, I would be open to any discussions or 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
62 1 questions associated with that joint report. And 2 without any further ado, I'll turn it back to Jose 3 March-Leuba and GEH to actually present on the topic 4 at hand, which is the SER for containment evaluation 5 performance.
6 Thank you.
7 MEMBER MARCH-LEUBA: Yes, Mike, this is 8 Jose. Just for clarification, this joint report was 9 an excellent activity that the U.S. performed. I'm 10 sure it produced a better review of the Topical Report 11 than without it.
12 MR. DUDEK: Right.
13 MEMBER MARCH-LEUBA: But it does not 14 affect the SER from the NRC. When GEH submits a 15 license to build a BWRX-300, they will have to refer 16 to the SER that you wrote, not the joint report, 17 correct?
18 MR. DUDEK: That is correct. That is 19 correct. The SER is the United States' publication of 20 our safety findings. I think CNSC took their own 21 learnings back and will incorporate their learnings 22 into their own report. But, yes, the SER is the 23 U.S.'s findings on the Topical Report submitted to us 24 by our applicant.
25 MEMBER MARCH-LEUBA: Yes. Now on this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
63 1 joint effort, in my personal opinion, I find it very 2 valuable. And we, as a Committee, I think, would like 3 to hear more about the effort itself, something at the 4 30,000-feet level to oversee what the whole effort is.
5 We don't want to know the details or limitations and 6 conditions, and everything, but at the high level.
7 And our people will be in contact with your people to 8 see if we can establish a time mutually acceptable.
9 CHAIRMAN REMPE: Right. I'm not sure if 10 you're the one to be asking this, Mike, but I know 11 from the popular press of this MOU, and interactions 12 with other folks here at the agency who have mentioned 13 it. And I know it encompasses more than just thermal 14 hydraulics. It includes fuel reviews and other 15 aspects of licensing applications.
16 And so, yes, I would like to share what 17 Jose is mentioning about having a general overview of 18 the types of activities and documents that have been 19 produced. And so, our staff will be reaching out to 20 the appropriate person, and we will be following this 21 up.
22 MR. DUDEK: Absolutely. I've taken that 23 for action. This is the third joint review that we 24 have on the books, and I think we have a few more 25 planned. So, we would be more than open to coming and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
64 1 talking to the Committee about the initiative.
2 CHAIRMAN REMPE: Wonderful. Thank you.
3 MEMBER MARCH-LEUBA: And with that, we'll 4 give GE the floor, George Watkins, and I won't try to 5 butcher your title. You introduce yourself.
6 MR. WATKINS: Okay, I will.
7 Good afternoon. My name is George 8 Watkins. I am the Vice President of New Nuclear 9 Plants and Products Licensing at GE Hitachi.
10 Today, we are going to discuss the LTR 11 NEDC-33922P, the BWRX-300 Containment Evaluation 12 Method Licensing Topical Report. The BWRX-300 design 13 leverages our economically-simplified boiling water 14 reactor design with some additional innovations to 15 improve simplicity of the plant while also enhancing 16 safety, reducing cost in the end for the plant to 17 ensure its commercial viability.
18 So, one area in the evolution of the 19 design from ESBWR to BWRX-300 is in the containment 20 design. The ESBWR had a dry well/wet well containment 21 with a suppression pool. It used a different 22 methodology for post-accident cooling, a different 23 methodology for reactor pressure control and level 24 control after an accident.
25 So, we have a much simpler, dry, inerted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 1 containment which has no major subcompartments. So, 2 it's a very simplistic containment design to analyze.
3 So, as a part of that simplification, we looked at 4 using a more simplified approach than TRACG for 5 evaluating the thermal hydraulic response of the 6 containment atmosphere after an accident.
7 So, therefore, we decided on GOTHIC, the 8 universally-used code in a lot of applications, and we 9 went about developing the evaluation method with 10 GOTHIC, based on the BWRX-300 conceptual design. And 11 so, this was the whole reason for the LTR, was to 12 present this as the evaluation method going forward 13 for future licensing applications.
14 So, I want to thank the NRC management and 15 staff for their review; also, CNSC for their joint 16 review of this LTR. I agree that I think it greatly 17 improved the review process. We did have a very 18 significant audit process as part of this, which also 19 aided in developing a very good product, in my 20 opinion.
21 So, with that, I'm going to turn it over 22 to Lisa Schichlein, who is our U.S. Licensing Manager.
23 MS. SCHICHLEIN: Good afternoon. My name 24 is Lisa Schichlein and I'm the U.S. Licensing Manager 25 for New Power Plants and Products at GE-Hitachi. I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
66 1 want to thank the ACRS Full Committee for the 2 opportunity to present BWRX-300 Containment Evaluation 3 Method Licensing Topical Report.
4 In addition to George Watkins we have in 5 attendance from GEH, the U.S. licensing manager for 6 the BWRX-300; and Necdet Kurul, principal engineer in 7 Containment. Supporting us on the call we have the 8 lead licensing engineers on the topical report:
9 Frostie White and Lou Lanese, as well as Charlie Heck, 10 who's a consulting engineer for nuclear applications; 11 Dan Pappone, the chief consulting engineer for plant 12 performance; David Hines, the principal engineer for 13 plant integration; and Rosanne Harrington, who's the 14 manager of LOCA and containment analysis for GE-15 Hitachi. In addition we have Dr. Tom George on the 16 call, who's a consultant from Zachry Engineering.
17 As questions arise I may ask one or more of those 18 folks to address the question.
19 GEH is seeking NRC approval to apply an 20 analysis method for evaluating the dry containment 21 thermal hydraulic performance for the BWRX-300 small 22 modular reactor using the TRACG and GOTHIC computer 23 codes. The driver for this new method is the inerted 24 dry containment of the BWRX-300, therefore we're using 25 a different code than we have used in the past for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
67 1 reactor and containment thermal hydraulic performance.
2 Specifically, GOTHIC is now being used with inputs 3 from TRACG to evaluate thermal hydraulic performance.
4 The scope of the topical report included 5 method description and qualification, sensitivity 6 studies, and demonstration cases. The analysis method 7 used for the BWRX-300 containment thermal hydraulic 8 performance demonstrates that the containment design 9 complies with 10 CFR 50, Appendix A General Design 10 Criteria 2, 4, 6 -- sorry, 2, 4, 16, 38, 41, 50 and 11 51.
12 MEMBER MARCH-LEUBA: Let me -- more 13 clarification. The methodology that we are approving 14 today does not demonstrate anything. Can be used to 15 demonstrate --
16 MS. SCHICHLEIN: Yes.
17 MEMBER MARCH-LEUBA: -- once we have a 18 design, correct?
19 MS. SCHICHLEIN: That is correct.
20 MEMBER MARCH-LEUBA: Okay. Thank you.
21 MS. SCHICHLEIN: The specific acceptance 22 criteria for the BWRX-300 containment performance are 23 discussed in Section 4 of the NRC-approved licensing 24 topical report, NEDC-33911P, which is entitled BWRX-25 300 Containment Performance. And I'd like to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
68 1 emphasize that GEH is not seeking NRC approval for any 2 exemptions from regulatory requirements.
3 There are several design features of the 4 BWRX-300 that are relevant to the containment 5 evaluation methodology in the topical report. First, 6 the containment is a dry enclosure inerted with 7 nitrogen at near atmospheric pressure during normal 8 operation. The containment design pressure and 9 temperature are within the experience base of 10 conventional BWRs. And there are no sub-compartments 11 containing large bore high energy lines and these sub-12 compartments have sufficiently large openings so the 13 boundaries of the sub-compartments do not experience 14 large pressure differentials from pipe breaks outside 15 of the sub-compartments.
16 The limiting large pipe breaks evaluated 17 in the topical report are the main steam pipe and 18 feedwater pipe with all large pipe breaks assumed to 19 rapidly isolate at the reactor pressure vessel nozzle.
20 For small breaks the limiting small break is an un-21 isolated instrument line break.
22 MEMBER MARCH-LEUBA: And I can probably 23 say this in open session: The small breaks are 24 therefore the limiting ones?
25 MS. SCHICHLEIN: Correct. The containment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
69 1 evaluation method involves the use of two codes, as we 2 mentioned before: TRACG and GOTHIC. TRACG is used to 3 evaluate the mass and energy release in a manner 4 consistent with its use in evaluating the mass and 5 energy release for the ESBWR. The ESBWR TRACG-LOCA 6 method is applied to the BWRX-300 mass and energy 7 release calculations to evaluate the large and small 8 steam and feedwater pipe break case and conservative 9 cases -- base and conservative cases.
10 GOTHIC is used to evaluate the containment 11 thermal hydraulic response and uses a new containment 12 model developed for BWRX-300. The GOTHIC code has 13 been benchmarked to separate effect and integral tests 14 and benchmarking to the test data of a similar size 15 containment is included in the topical report.
16 The containment response evaluation method 17 for BWRX-300 includes both base and conservative cases 18 and for the conservative cases key inputs, 19 assumptions, and modeling parameters are 20 simultaneously conservatively biased, which is the 21 same approach used and approved for the ESBWR.
22 In conclusion I would like to wrap up this 23 presentation by restating that GEH is seeking NRC 24 approval for application of an analysis method to be 25 used for evaluating the BWRX-300 dry containment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
70 1 thermal hydraulic performance. GEH is not requesting 2 NRC approval for exemptions from any regulatory 3 requirements. And just to restate, the method is 4 using two computer codes: TRACG and GOTHIC. TRACG 5 uses the applicable parts of the approved topical 6 report EBC33083E-A, Revision 1, for the application of 7 TRACG to ESBWR, which is incorporated in the approved 8 ESBWR design certification. And GOTHIC is the 9 standard code used for thermal hydraulic containment 10 evaluations in the nuclear industry.
11 With the new methodology individual key 12 inputs, assumptions, and modeling parameters are 13 conservatively biased simultaneously, which is the 14 same approach taken for the ESBWR containment method 15 in the NRC-approved topical report entitled TRACG 16 Application for ESBWR.
17 That concludes my brief presentation. Are 18 there any questions or comments for GE-Hitachi?
19 MEMBER MARCH-LEUBA: Yes, this is a 20 methodology report and we still don't have an X-300 21 design, but in your report you have provided a large 22 number of demonstration analysis on conceptual 23 concepts for your reactor. Do we have a lot of margin 24 on your conceptual X-300 or -- I mean, yes, can we 25 talk about margins in an open session?
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71 1 MR. KURUL: Currently we do have margin in 2 our calculations. The results first. It is the large 3 break LOCA calculations that are the limiting cases, 4 which is back through containment design limits.
5 Small break cases are non-limiting. They are just 6 there to demonstrate that they are not --
7 MEMBER MARCH-LEUBA: Can you state your 8 name for the record?
9 MR. KURUL: It's Necdet Kurul, GE-Hitachi.
10 For the large break cases we do have sufficient 11 margin. And as we develop containment design further, 12 we are watching that margin so that we don't end up 13 coming right up against the design limits as we 14 develop the containment further. And if it so happens 15 that we end up getting into the margins much, then we 16 would have to increase the design limit and the 17 containment design would have to accommodate that 18 increase in the --
19 MEMBER MARCH-LEUBA: Basically we have to 20 put more concrete on the containment?
21 MR. KURUL: More concrete actually we find 22 that we have more concrete then. But we put more 23 concrete or put more steel, whatever we need to do to 24 increase design pressure. So any changes in the 25 containment design development goes along with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 1 maintaining margin.
2 MEMBER MARCH-LEUBA: Yes, and I realize 3 that you are more focused than I am on containment.
4 The small break LOCA is more limiting on conductivity, 5 correct? And that will be a transient analysis, not 6 containment?
7 MR. KURUL: That's correct.
8 MEMBER MARCH-LEUBA: Yes.
9 MEMBER HALNON: Yes, let me step back to 10 the licensing aspect. It doesn't sound like either 11 code is new to the NRC and it's not new to the 12 application of establishing containment performance.
13 Are you asking for this approval as a feel good to 14 make sure that you're confident going forward or is it 15 something that really has not previously been approved 16 for use by the NRC even though it's on this 17 application?
18 MR. WATKINS: Yes, it's George Watkins, 19 GEH. Yes, the simple answer is we do want to mitigate 20 licensing risk going forward with our applications 21 with the various utilities. The genesis of our 22 licensing topical report strategy has been to mitigate 23 new features that are in the BWRX-300 beyond the ESBWR 24 base design so that when we get to the point of an 25 actual applicant for a construction permit or a COLA, 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
73 1 whichever path we end up going down there, these 2 issues will already have some certain regulatory basis 3 going forward. So that's the real reason that we're 4 asking for these approvals at this time.
5 The BWRX-300 containment is -- kind of 6 goes back to the root. Some of our early BWRs had dry 7 containments, but everything since then has been the 8 suppression-type containment. So we wanted to make 9 sure that we could demonstrate to the NRC staff that 10 we had the analytical capabilities and understood the 11 method sufficiently to apply to this new design.
12 MEMBER HALNON: Okay. So its iterative 13 approach is like you said for regulatory certainty? I 14 mean this could easily have been done -- once you get 15 your design -- your containment designed you could 16 have established all this at --
17 MR. WATKINS: That is correct.
18 MEMBER HALNON: -- one time.
19 MR. WATKINS: So it is appropriate to say 20 we're asking approval of the methodology, not its 21 total application. That will come with an actual 22 license application. And the staff will of course 23 talk about their limitations to conditions which 24 address primarily those types of issues that --
25 MEMBER HALNON: Okay. I was just --
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74 1 (Simultaneous speaking.)
2 MR. WATKINS: -- apply to an actual 3 application.
4 MEMBER HALNON: All right. I was trying 5 to figure out the iterative approach that you took --
6 MR. WATKINS: Right.
7 MEMBER HALNON: -- and I understand that.
8 Thank you.
9 MR. WATKINS: Okay. Thank you.
10 MEMBER MARCH-LEUBA: Members, any more 11 questions for GE?
12 VICE CHAIRMAN KIRCHNER: I'd just like to 13 make a comment and I'm pleased to see that you're 14 using the TRACG code. Just for the record, TRAC had 15 its genesis in investments made by research at Los 16 Alamos starting in 1976. I wrote the transfer package 17 for the code. I bet if you showed me your code, I 18 wouldn't recognize it anymore.
19 (Laughter.)
20 VICE CHAIRMAN KIRCHNER: But GE picked up 21 the code. I'd estimate that was '78, 1978-1980 22 timeframe. I don't know if any of your colleagues are 23 still with you that had picked that code up in that 24 time, but I was one of them at that time.
25 So, but the point of my comment isn't to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
75 1 draw attention to me; it's to draw attention to the 2 RES investments that were made way back and still 3 paying off for the industry and for the staff because 4 TRACE had some similar genesis. Thank you.
5 MEMBER MARCH-LEUBA: Any more comments or 6 questions for GE?
7 Okay. So we're going to have a short 8 transition to the staff presentation. Will you have 9 to set up your computers here?
10 While we're in transition I want to say 11 that this is the first time we have had visitors for 12 outside in this room in -- since forever. So you're 13 very welcome. Thank you for being here.
14 CHAIRMAN REMPE: It's really nice to see 15 people and actually see their smiles or frowns at us, 16 as the case may be.
17 MEMBER MARCH-LEUBA: Yes, and I have asked 18 Ken for tomorrow's presentation, which you'll be 19 hearing, to have name tags so nobody calls George Greg 20 again.
21 (Pause.)
22 MEMBER MARCH-LEUBA: Okay. So we're back 23 in session. We have the staff properly set on the 24 table. And please go ahead and start making your 25 presentation.
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76 1 MR. SHEA: Good afternoon. This is Jim 2 Shea of the staff. Before we start I wanted to --
3 Becky Patton wanted to make some opening remarks.
4 MS. PATTON: Yes, this is Becky Patton.
5 I'm the branch chief of the Nuclear Methods Systems 6 and New Reactor Branch in NRR. My colleague Scott 7 Krepel from one of the other reactor systems branches 8 addressed the Subcommittee last time. As me mentioned 9 we had an integrated review team of multiple branches 10 working on this as well as our colleagues in the 11 Office of Research who provided essential support 12 throughout the review. And then there was also 13 coordination with the CNSC on this review.
14 So there were a lot of people involved.
15 There was very close communication including the year-16 long audit that took place during the review. And I 17 understand the Subcommittee meeting was very engaging.
18 There were a lot of questions. We really appreciate 19 the Subcommittee's look into these matters and we look 20 forward to the presentation today as well.
21 The staff did look at the issues they felt 22 were most important, most significant in terms of 23 safety, and that's always our number one thing for all 24 of this. And even though the review was conducted on 25 a more compressed timeframe than would be typical for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 1 this type of review, because we did it in such an 2 integrated manner and with resources put to bear, we 3 believe it was very effective and it was also a 4 benefit to be able to discuss with our colleagues at 5 the CNSC. So I appreciate the attention of the 6 Committee and go forward with the presentation.
7 MEMBER MARCH-LEUBA: Thank you. Let me 8 interrupt before you go in. Remind everybody, we're 9 in open session. If we get into a closed -- we have 10 a closed line if necessary, but I hope we don't have 11 to use it. Just stay non-proprietary.
12 MR. SHEA: Good afternoon. I am James 13 Shea of the NRC's Office of Nuclear Reactor Regulation 14 in the Division of New and Renewed Licenses. I am the 15 lead project manager for the pre-application 16 activities for the GEH BWRX-300 design.
17 I wanted to thank the ACRS Full Committee 18 for this time to allow the staff to present its review 19 of the Licensing Topical Report: Containment 20 Evaluation Method, as reflected in the staff's 21 advanced safety evaluation submitted publicly on March 22 9th, 2022.
23 The staff's review focused on the approval 24 of the BWRX-300 containment evaluation methods for 25 evaluation of the design-basis events. As stated in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
78 1 the safety evaluation report, the NRC will evaluate 2 the compliance of the final design of the BWRX-300 3 containment performance during future licensing 4 activities in accordance with 10 CFR Part 5, or 10 CFR 5 Part 52.
6 The staff presented its approval of this 7 topical report to the ACRS BWR Subcommittee on March 8 18th, 2022 and the staff today will provide a high-9 level summary of the Subcommittee presentation.
10 Just to recap, the LTR was submitted to 11 the NRC on September 25th, 2020. Revision 1 12 supplemented with requests for additional information, 13 RAI, responses was submitted on November 19th, 2021 14 and Revision 2 was submitted on December 17th, 2021.
15 I will now turn over the staff 16 presentation to our lead NRR technical staff reviewer 17 for the containment evaluation methods, Syed Haider.
18 MR. HAIDER: Thank you, Jim.
19 Good afternoon to everyone. My name is 20 Syed Haider. I am the lead reviewer of the GEH 21 licensing topical report, LTR NEDC-3392P, which is an 22 LTR related to evaluating the containment pressure and 23 temperature response that would result from the 24 limiting design-basis accidents involving mass energy 25 release from the reactor pressure vessel into the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 1 containment for the BWRX-300 SMR.
2 My Nuclear Systems Performance Branch at 3 NRR, the lead for the review under Branch Chief Scott 4 Krepel. My NRR colleagues Carl Thurston and Shanlai 5 were the reviewers for the LTR part related to the 6 TRACG modeling and mass energy release calculation 7 methodology. I myself mainly reviewed the GOTHIC 8 modeling and containment pressure and temperature 9 response calculation methodology. Chang Li from NRR 10 reviewed combustible gases inside the containment.
11 Throughout the review the NRC staff was 12 supported by the research staff that included Peter 13 Lien, Joe Staudenmeier, Andrew Ireland, and Shawn 14 Campbell. Research was responsible for the 15 development of the TRACE and MELCOR confirmatory 16 models that were used to analyze and validate the 17 TRACG and GOTHIC model results and trends presented in 18 the LTR.
19 This slide shows the outline of the NRC 20 staff's presentation today during which we will not 21 discuss any proprietary information about the BWRX 22 LTR. Should there be a need, we can have a closed 23 session to respond to any proprietary questions.
24 First, the staff will give an overview of 25 the purpose and scope of the BWRX-300 LTR. It's worth NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
80 1 mentioning here that even though the LTR title is 2 BWRX-300 Containment Evaluation Method, it covers the 3 methodology and modeling details for both the BWRX 4 containment as well as the reactor systems.
5 First, we will explain the acceptance 6 criteria for the BWRX-300 containment response and 7 underlying regulatory requirements as followed in the 8 evaluation methodology.
9 Then we will highlight the BWRX 10 containment design background that drives the core of 11 the evaluation methodology.
12 We will review the key components of the 13 BWRX-300 design as they related to the TRACG and 14 GOTHIC code-based reactor systems and containment 15 analysis methodologies with the eventual objective of 16 predicting the containment thermal hydraulic response 17 for design-basis accidents with sufficient 18 conservatism.
19 Then we will give a summary of the BWRX 20 containment evaluation method demonstration analyses 21 that are presented in the LTR.
22 After this my NRR colleague Carl Thurston 23 will present a few slides on our TRACG codes-based 24 mass energy release calculation methodology review.
25 After that I will present a few slides on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 1 our review of the GOTHIC code-based containment 2 response calculation methodology.
3 Then I will provide an overview of the NRC 4 staff's confirmatory analysis that were conducted to 5 develop additional insights and support our regulatory 6 review findings.
7 At the end Carl and I will describe the 8 four conditions and limitations on the LTR as they 9 emerged during the staff review of the LTR.
10 Finally, we will present the staff 11 conclusions for the LTR review.
12 This slide describes the purpose and scope 13 of the BWRX-300 containment evaluation methodology LTR 14 as reviewed by the staff. Basically GE-Hitachi has 15 submitted this LTR to obtain the NRC's approval of the 16 BWRX-300 containment pressure and temperature analysis 17 methodology. The approved methodology is intended to 18 be used to perform the BWRX-300 containment safety 19 analyses, to support either a licensing application 20 for a construction permit and operating license under 21 CFR 50 or a design certification and combined license 22 application under 10 CFR Part 52.
23 This slide also highlights the four 24 acceptance criteria used in the methodology for the 25 BWRX-300 containment response. These acceptance NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 1 criteria are identified in the LTR and are essentially 2 driven by several NRC regulations.
3 The first acceptance criterion is a rooted 4 in General Design Criteria for -- or GDCs 16 and 50 of 5 10 CFR Part 50, Appendix A, and ensures that the peak 6 containment pressure and maximum shell temperature 7 calculated for the limiting mass energy release 8 design-basis accident are bounded by the BWRX-300 9 design pressure and temperature with sufficient 10 margin.
11 The second acceptance criterion is driven 12 by GDC 38 for containment heat removal. That is 13 interpreted by the Standard Review Plan as to ensure 14 that the containment pressure is reduced to less than 15 50 percent of the peak accident pressure for the most 16 limiting LOCA within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
17 GDC 38 also requires that the containment 18 pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for all LOCAs are maintained 19 below 50 percent of the peak pressure for the most 20 limiting LOCA which constitutes the third acceptance 21 criterion for the LTR.
22 These three acceptance criteria 23 collectively ensure that the containment structure can 24 accommodate the pressure and temperature condition 25 resulting from any mass energy release from the RPV NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
83 1 without exceeding the design leakage area and with 2 sufficient margin.
3 The fourth acceptance criterion requires 4 that the containment atmosphere remain sufficiently 5 mixed such that deflagration or detonation does not 6 occur inside the containment.
7 MEMBER HALNON: Syed, on your most 8 limiting LOCA is it the same most limiting LOCA for 9 each one of those, or it is the most limiting LOCA for 10 that specific design criteria? In other words, it 11 would not be the same and the one that gets you to 12 below 50 percent in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may not be the same that 13 challenges your 50 percent average. Is it the same or 14 is it different?
15 MR. HAIDER: It's the same. I think that 16 your question has two part. I heard you're also 17 asking whether the peak containment pressure accident 18 is the same as the peak shell temperature?
19 MEMBER HALNON: No, no. The most limiting 20 LOCA --
21 MR. HAIDER: Yes.
22 MEMBER HALNON: -- where the second bullet 23 there talks about the one that gets you down to --
24 it's got to be less than 50 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
25 MR. HAIDER: That's right.
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84 1 MEMBER HALNON: Is that the same most 2 limiting LOCA scenario for the third bullet, which is 3 the one that may challenge the 50 percent heat 4 pressure?
5 MR. HAIDER: Okay. The third bullet 6 applies to all LOCAs.
7 MEMBER HALNON: Okay. In other words --
9 (Simultaneous speaking.)
10 MEMBER HALNON: -- the most limiting LOCA 11 is the one that challenges that the most?
12 MR. HAIDER: That's right.
13 MEMBER HALNON: That's not necessary the 14 same scenario all the way through?
15 MR. HAIDER: That's correct.
16 MEMBER HALNON: Okay. Thanks.
17 MEMBER MARCH-LEUBA: But this is related 18 to what I mentioned earlier that there is the most 19 limiting LOCA with respect to containment pressure, 20 core uncovery, long-term cooling and heating. So when 21 you say most limiting LOCA, I assume you refer to the 22 peak pressure --
23 MR. HAIDER: That's correct.
24 MEMBER MARCH-LEUBA: -- in this context?
25 MR. HAIDER: That's correct. In this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
85 1 context, that's right.
2 MEMBER MARCH-LEUBA: Yes, because even 3 though we're evaluating a methodology for containment 4 -- to identify what the containment atmosphere does 5 during LOCAs, I know you have also included in the 6 review what happens inside the vessel, especially with 7 no condensables and such. So you keep in mind the 8 whole -- the holistic safety of the reactor, not just 9 the containment when you were doing your review, 10 correct?
11 MR. HAIDER: That's correct.
12 So now we will summarize the state of the 13 art of the BWRX containment design and its relation 14 with the staff review. This slide presents the 15 salient design features of the BWRX-300 containment 16 that are relevant to the evaluation methodology for 17 the design-basis accidents.
18 These features drive several initial and 19 boundary conditions in the containment GOTHIC model.
20 Basically BWRX-300 has a dry containment that is 21 inerted with nitrogen for normal operation. It does 22 not have a suppression pool inside the containment 23 like ESBWR had. This information is relevant to the 24 initial containment pressurization and the post-25 accident mixing of steam and radiolytic gases inside 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
86 1 the containment.
2 A key design feature of the BWRX-300 3 design is early closure of the RPV isolation valve for 4 the line breaks that limits the mass energy release 5 into the containment during large break LOCA which 6 dictates the peak containment pressure. However, the 7 RPV isolation valve does not close for small breaks 8 that remain un-isolated resulting in a continuous 9 break flow throughout the small break LOCA analysis.
10 The BWRX-300 design employs a passive 11 containment coolant system, or PCCS, to remove heat 12 from the containment to a reactor cavity pool that is 13 located above the containment. PCCS plays an 14 important role in the long-term containment pressure 15 reduction and mitigation. In this regard the LTR 16 demonstration analyses were performed with the 17 specific PCCS design configuration and described in 18 the LTR. The reactor cavity pool that's used -- the 19 PCCS for containment heat removal is located above the 20 containment dome.
21 So at this stage it's worth emphasizing 22 that the applicant has not submitted an ESAR or DCD 23 yet and the containment design pressure and 24 temperature are not available. Therefore the staff 25 has performed this LTR review with an understanding NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
87 1 that the BWRX-300 design is not final and the analyses 2 presented in the LTR are demonstration cases.
3 So the objective of the present review is 4 to evaluate the applicability of the methodology of 5 the BWRX-300 design and assess the degree of 6 conservatism built into the methodology. And this has 7 required the staff to focus on the uncertainties in 8 the methodology and to predict trends without getting 9 unduly hooked to exact figures. The staff understands 10 that the inputs and outputs near ready for the final 11 design that will be submitted at the licensing stage.
12 MEMBER MARCH-LEUBA: I have a comment. I 13 hear from the applicant something I like very much and 14 which I have inferred by reading that the approach is 15 to develop and approve the containment pressure --
16 overpressure methodology calculation so that they can 17 design the containment properly as opposed to -- what 18 they could have done is guess at what the containment 19 needed to be and then hope their methodology proves 20 that they were right. So the fact that they chose to 21 define and get approved the methodology by which the 22 containment will be judged and then build the 23 containment according to that is smart. I think it's 24 a good approach. Do you agree?
25 MR. HAIDER: Yes. Yes, I agree because NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 1 this too -- the limit of margin is not coming to --
2 into the play let. We are dealing with uncertainties 3 and conservatisms --
4 MEMBER MARCH-LEUBA: Yes.
5 MR. HAIDER: -- in this review.
6 MEMBER MARCH-LEUBA: It is -- I mean I'm 7 guessing right here, I'm assuming that the margin 8 they're going to have for containment will be 9 sufficient to be comfortable, but won't be over design 10 because that costs money. But by going this approach 11 of having the methodology given approval ahead of 12 time, we can agree on what pressure will be in the 13 containment and what pressure you need to design your 14 containment. So that's -- we haven't seen that for 15 all the applicants that -- while I'll been here in 16 this table. Some of the other applicants review their 17 methodology after they got a design.
18 MEMBER HALNON: Is it fair to say the 19 demonstration cases are close to the ballpark where 20 they expect the design parameters to be in the final 21 design so that we know that we're at least in this 22 general area? Methodology works fine?
23 MR. THURSTON: Yes, this is Carl Thurston.
24 MEMBER HALNON: Thanks.
25 MR. HAIDER: Yes.
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89 1 MR. THURSTON: Yes. So yes, that is 2 correct. So we understand that the design of the 3 reactor vessel is pretty much final, near or 4 finalized. So we expect that the conservatism that we 5 deserve for this demonstration cases will be carried 6 over exactly.
7 MEMBER HALNON: Okay. So the exact values 8 you see coming in are going to be pre-representative 9 of what you've been feeling?
10 MR. HAIDER: That's true. That's true.
11 MEMBER HALNON: Okay.
12 MR. HAIDER: But at least by presenting 13 the demonstration cases they are able to scale the 14 level of conservatism that exist in the good sense of 15 -- or the level of conservatism is built into the 16 methodology. The margin will be on top of. And they 17 may choose to in some --
18 MEMBER HALNON: Is there an idea of what 19 the uncertainties are? Thanks.
20 MEMBER MARCH-LEUBA: Carl, for the benefit 21 of the court recorder, can you state your name because 22 he doesn't know your voice.
23 MR. THURSTON: Yes, Carl Thurston. Thank 24 you. NRR DSS SB.
25 MR. HAIDER: So for further methodology NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
90 1 evaluation purposes for the present BWRX-300 2 containment design the LTR presents demonstration 3 analyses and results for the thermal hydraulic 4 performance in order to show that the acceptance 5 criteria are satisfied for the limiting design-basis 6 accident. In this backdrop the LTR presents results 7 for large break loss of coolant accident, and small 8 break loss of coolant accident inside the containment.
9 The containment design-basis analyses 10 presented in the LTR include both liquid and steam 11 space breaks and the applicant showed that all 12 acceptance criteria were satisfied for the 13 demonstration cases, however the applicant would be 14 required to demonstrate satisfying the acceptance 15 criteria for the BWRX-300 design submitted at the 16 licensing stage while also meeting the four 17 limitations and condition the staff has imposed on the 18 LTR as a result of this review.
19 The CE methodology uses TRACG code to 20 calculate the mass energy release for the RPV into the 21 containment and GOTHIC code to calculate the resulting 22 containment hydraulic response.
23 Before we describe the TRACG and GOTHIC 24 parts of the review it's worth explaining that the 25 applicant submitted two stand-alone or decoupled NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 1 models that were based on the TRACG code or mass 2 energy release calculation from the RPV and GOTHIC for 3 containment pressure and temperature response 4 calculations. The NRC staff used TRACE and MELCOR 5 codes to develop confirmatory models by using the 6 input specifications that will directly derived from 7 the TRACG and GOTHIC decks as submitted by the 8 applicant. This allowed the staff to ensure a direct 9 comparison between the confirmatory model results and 10 the TRACG/GOTHIC results for both large break LOCA and 11 small break LOCA cases.
12 The staff confirmatory analyses were 13 mainly performed to accelerate a regulatory review by 14 developing necessary insights to focus the staff's 15 efforts on safety significant issues in the BWRX-300 16 phenomenal review.
17 MEMBER MARCH-LEUBA: Let me interrupt you 18 there since you are breaking. I wanted to say this 19 before but I got sidetrack. A key feature of the 20 BWRX-300 is the isolation valves that you assume they 21 work for your analysis, and rightly so for Chapter 15-22 type analysis. But when we move into the risk 23 analysis at Chapter 19-type analysis, there will have 24 to be an evaluation by the staff of why we have so 25 much confidence that those valves will close on time NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 1 and the pipe upstream of it won't break. Just put a 2 marker there that when we do that, let's not forget to 3 review that because it is a key. If those valves 4 don't work, the concept does not work. So we believe 5 it does. It will. But we need to cross our Ts and 6 dot our Is. Okay. Thank you.
7 MR. HAIDER: Thanks. I really appreciate.
8 So now I will turn over the presentation 9 to my NRR colleague Carl Thurston to present a summary 10 of the TRACG modeling of the mass energy release for 11 BWRX containment.
12 MR. THURSTON: My name is Carl Thurston 13 and I am the primary reviewer for the TRACG code 14 calculation of mass and energy release.
15 TRACG code has a long history of review by 16 the NRC staff due to several previous topical reports 17 submitted for SBWR, for ESBWR, and for the existing 18 fleet of BWR/2 through 6 plants.
19 The staff focused its review of the code 20 on changes made since the ESBWR timeframe, which is 21 around 2008-2009. The latest TRACG code was used for 22 this analysis and the staff confirmed that there were 23 no significant changes made since ESBWR.
24 The RPV model and internal components are 25 pretty much scale from the ESBWR design. The NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
93 1 methodology uses what we call a decoupled approach 2 though the containment is allowed to blowdown 3 continuously into atmospheric pressure. So the 4 boundary condition for the TRACG code is fixed at 5 atmospheric pressure and that maximizes the amount of 6 mass and energy release that will go into the GOTHIC 7 code.
8 The methodology relies heavily upon 9 previous TRACG topical reports and their compatibility 10 with the BWRX-300. This is particularly true for the 11 amount and the extensive qualifications done for the 12 ESBWR design. So the ESBWR qualification was extended 13 to BWRX-300 such that the ESBWR PIRT and the modeling 14 basis are applied for the RPV and its internals. The 15 BWR/2 through 6 methods and topical reports are 16 invoked since some events result in core uncovery.
17 The RCS, the isolation condenser have a significantly 18 more important safety function than for ESBWR and is 19 modeled in considerably more detail. Staff reviewed 20 these modeling features and deemed them to be adequate 21 for M&E release calculations with the limit and 22 conditions applied.
23 Next slide. The next slide reviews unique 24 features of the BWRX-300 design as compared to ESBWR.
25 And as was mentioned, the large break LOCA isolation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
94 1 valves are a significant change and are significant in 2 allowing the pressure to be limited because these 3 valves close very quickly once a pipe break occurs.
4 In addition, said several times, there's 5 no suppression pool and the isolation condensers are 6 the primary method for decay heat removal. So they 7 have to function at rated conditions or better in 8 order for the plant to cool down.
9 The RPV isolation valve limits break flow 10 and M&E release for large piping, but the small breaks 11 are un-isolated and they can continue to blowdown for 12 the full time of the transient. The methodology 13 assumes that one of the ICS trains are inoperative due 14 to single failure.
15 The staff reviewed the conditions assumed 16 and found that the inputs were conservative, related 17 to initial power, power history, scram time, tilt flow 18 modeling, atmospheric pressure for the bounding 19 conditions indicated, and the operating conditions 20 assumed for the plant, at generally 102 percent.
21 MEMBER MARCH-LEUBA: Do you remember what 22 was the target pressure for the -- target peak 23 pressure for containment? Is it like 50 psi?
24 MR. THURSTON: I don't recall if we had a 25 target pressure.
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95 1 Syed, you can chime in.
2 MR. HAIDER: It's -- you mean predicted 3 heat containment pressure? Yes, 61 psi. Yes.
4 MEMBER MARCH-LEUBA: Yes, approximately 5 60. So I'm going to say pressure boundary condition 6 is probably not very conservative at the beginning 7 when you're 1000 psi inside, but eventually it will 8 become conservative.
9 MR. THURSTON: Sure. Yes, until you 10 ignore the pressure in the containment, right, and 11 force as much mass and energy out of the reactor that 12 you can.
13 MEMBER MARCH-LEUBA: Yes, that's clearly 14 one of those obviously conservative assumptions.
15 MR. THURSTON: Correct. Yes.
16 Next slide, Jim.
17 So next we will review the findings, the 18 limiting conditions that the staff had related to the 19 BWRX-300 topical report methodology.
20 The first issue was related to radiolytic 21 gas accumulation. The staff determined that the TRACG 22 model did not adequately account for the accumulation 23 of radiolytic gases in the IC tubes and its subsequent 24 effect on heat transfer performance. Staff issued an 25 RAI and GEH responded that a design change would be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
96 1 implemented to limit build-up of non-condensable gases 2 to negligible amounts.
3 MEMBER MARCH-LEUBA: For the deflagration 4 margin had you guys given any consideration to the 5 fact that there really is no oxygen there, that if any 6 gas were to enter it would be nitrogen? Would you 7 grace for that or you say however much hydrogen you 8 have assume the oxygen miraculously shows up?
9 MR. THURSTON: Where they delve in those 10 details Shanlai is -- we can --
11 MR. HAIDER: It's actually a little bit 12 proprietary information, so we can definitely provide 13 discussion on that if you --
14 MEMBER MARCH-LEUBA: No, I was just 15 curious.
16 MR. HAIDER: Yes, because there is a 17 number there.
18 MEMBER MARCH-LEUBA: Yes, my question is 19 what if GE comes up with an argument, would you be 20 willing to consider it?
21 MR. LU: Yes. I can provide you that, at 22 least the open session answer. And I think during the 23 normal operation there is a purge line through the top 24 of the isolation condenser, continuous purging the 25 non-condensable gas there. So therefore, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
97 1 consideration for that one is very minimal in terms of 2 the existing oxygen pressure inside of the isolation 3 condenser. During normal operation it's completely 4 flooded anyway. So I think that's a good question and 5 if you want to talk about a little more detail, I 6 think that's a proprietary session.
7 MEMBER MARCH-LEUBA: I know you're 8 proprietary answer because I was in the Subcommittee, 9 so I don't need it.
10 MR. LU: Thank you. Oh, this is Shanlai 11 Lu from the staff.
12 MARCH-LEUBA: Carl?
13 (Simultaneous speaking.)
14 MR. THURSTON: -- any more questions 15 related to the first one, then move to the second 16 issue.
17 So our next issue was related to reverse 18 flow of steam in the ICs return line. The staff noted 19 that the design does not include a trap to prevent 20 reverse flow of steam and radiolytic gases from the 21 RPV back into the ICs and ICs tubes.
22 The staff issued an RAI and GEH indicated 23 that in the final design a trap would be included to 24 prevent this type of reverse flow.
25 MEMBER MARCH-LEUBA: And is their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
98 1 commitment part of Revision 2, or is still on the RAI?
2 MR. THURSTON: Huh?
3 MEMBER MARCH-LEUBA: The commitment is 4 part of Revision 2 or it's still in the RAI, the 5 commitment to include a loop seal?
6 MR. THURSTON: Yes, so it's included as a 7 limitation, so we expect them to --
8 MEMBER MARCH-LEUBA: Okay.
9 MR. THURSTON: -- comply.
10 MEMBER MARCH-LEUBA: Just part of the SER?
11 MR. THURSTON: Yes, sir.
12 MEMBER MARCH-LEUBA: I think the SER says 13 include a loop seal or analyze it properly.
14 MR. THURSTON: That's correct. It does 15 say or, but we are --
16 MEMBER MARCH-LEUBA: But we --
17 MR. THURSTON: Yes, we don't want to 18 depend on something for analysis if we know that 19 design can do -- and minimize any type of extensive 20 analysis.
21 MEMBER MARCH-LEUBA: I think you hit the 22 nail on the head. When we deal with an operating 23 reactor that is already built, making modifications 24 like this is very difficult, very costly and the 25 operator -- the ignition that have to go there will NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
99 1 get a significant dose. So we tend to rely on 2 calculations, more and more accurate calculations, 3 expensive calculations to show it can happen.
4 At the design stage I applaud the approach 5 you guys have taken to modify the design. So a simple 6 calculation shows you cannot have backflow.
7 MR. THURSTON: That's right.
8 MEMBER MARCH-LEUBA: And at the design 9 time it doesn't cost money. It doesn't cause a lot of 10 considerations. This is the time to do it. Please 11 continue. Continue both with your presentation and 12 with the good work.
13 MR. THURSTON: Now I'll turn it back over 14 to Syed.
15 MR. HAIDER: Okay. Thanks, Carl. This is 16 Syed Haider. This slide presents an overview of the 17 GOTHIC code as reviewed by the staff for the BWRX-300 18 containment response modeling in the LTR application 19 submitted for the NRC approval.
20 GOTHIC code is a commission-developed 21 computer code that has been widely used in the 22 industry for containment thermal hydraulic response 23 analyses. This is a continuously improved code that 24 is compliant with the pertinent NRC regulation 10 CFR 25 Part 50, Appendix B on quality assurance. The NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
100 1 applicant has used the latest GOTHIC version, 8.3, for 2 the BWRX-300 analyses demonstrated in the LTR. They 3 also clarified that they even follow development NRC 4 process should there be a need to use a different 5 GOTHIC version at the licensing stage. As far as the 6 GOTHIC-based application in the past, our concern 7 GOTHIC code has been previously approved by NRC for 8 containment response analyses.
9 The staff also reviewed the BWRX-300 10 containment PIRT phenomenology and found it covered by 11 the available GOTHIC code functionalities.
12 The staff also reviewed the GOTHIC 13 benchmarking against the CVTR test data provided in 14 the LTR for the BWRX-300 containment methodology 15 qualification and found it acceptable.
16 The information provided in the LTR and 17 RAI responses demonstrated that GOTHIC is qualified 18 for modeling thermal stratification and multi-19 dimensional flow. These features are important for 20 evaluating the nodalization-based containment design 21 that employs distributed PCCS units.
22 Next slide, please? This slide captures 23 the key features of the GOTHIC model development and 24 the containment response calculation methodology in 25 the LTR. The BWRX-300 containment evaluation method NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
101 1 has been developed following the applicable element of 2 Reg Guide 1.203. Reg Guide 1.203 presents a multi-3 step process for developing an acceptable evaluation 4 method to analyze transient nuclear power plant 5 response for reg-defined figures of merit during the 6 postulated design basis events.
7 The BWRX-300 containment evaluation 8 methodology is based on a decoupled calculation 9 between the TRACG and GOTHIC codes where TRACG model 10 first calculates the mass energy release from the RPV 11 into the containment assuming it remains at the same 12 initial atmospheric pressure. This way the TRACG mass 13 energy release calculations do not account for the 14 increasing containment back pressure.
15 Using this decoupled approach the stand-16 alone GOTHIC containment model independently uses the 17 mass energy release calculated by TRACG as a 18 containment boundary condition for calculating the 19 containment pressure and temperature response. The 20 staff agrees that ignoring back pressure ensures a 21 conservative mass energy release into the containment 22 with respect to containment pressure and temperature 23 response.
24 As described in the LTR the containment 25 evaluation method is based on a four-component GOTHIC NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
102 1 model that include a nodalized main containment, a 2 nodalized dome, a nodalized PCCS, and a lumped reactor 3 cavity water pool within the space. A conservative 4 diffusion layer model is used and justified for 5 modeling condensation heat transfer on the containment 6 shell and PCCS surface. The DLM condensation modeling 7 option is available as a functionality in the GOTHIC.
8 DLM is a mechanistic model that represents the 9 underlying condensation heat transfer phenomena which 10 is based on heat and mass transfer analogy as opposed 11 to a curve fit of the test data.
12 The GOTHIC modeling approach accounts for 13 thermal stratification inside the containment, which 14 is important due to the containment and PCCS 15 nodalization used in the BWRX-300 containment 16 evaluation methodology.
17 This slide summarizes the review NRC staff 18 performed of the BWRX-300 GOTHIC containment response 19 methodology. The staff reviewed the important 20 physical phenomena again defined in the GOTHIC PIRT 21 table for the BWRX-300 design and found them 22 acceptable. The applicant submitted GOTHIC decks for 23 the base and conservative cases for both large break 24 and small break LOCAs. The staff reviewed the GOTHIC 25 input models and used the submitted GOTHIC decks to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
103 1 gather information to develop the TRACE and MELCOR 2 models for confirmatory analyses.
3 Key modelings, uncertainties, and 4 conservative biases were reviewed to evaluate the 5 overall GOTHIC model conservatism. We reviewed 6 information about the nodalization sensitivity studies 7 for the containment and PCCS. We also reviewed the 8 GOTHIC code benchmarking as documented in the LTR 9 against the CVTR test data that's applicable to BWRX-10 300 containment design as well as the justifications 11 for bounding the containment condensation and natural 12 convection heat transfer correlations by using the 13 opening test data.
14 The staff also made sure that the BWRX-300 15 containment response analyses for small break and 16 large break LOCA duly account for the appropriate 17 break location and break flow direction sensitivities.
18 In this process we identified a small un-isolated 19 liquid break as the limiting small break LOCA.
20 The staff also reviewed the modeling of 21 the PCCS that is used to reduce and mitigate the 22 containment pressure in the long term. Containment 23 mixing was also reviewed for potential deflagration 24 and detonation due to combustible gases accumulation.
25 Now this slide summarizes the key typical 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
104 1 issues that the staff encountered on the GOTHIC 2 containment side of the review. We present only the 3 non-proprietary details about the issues or 4 resolutions here. Any prop information on 5 confirmatory analysis results will require a separate 6 session.
7 First, the originally-submitted revisions 8 of the LTR had not performed any break location or 9 breakthrough directional study while identifying 10 limiting breaks with respect to containment pressure 11 and temperature. The staff also noticed that the 12 limiting small break LOCA was a steam break while no 13 containment analysis was performed for a small liquid 14 break. So RAIs were issued and their response did 15 show sensitivity to break location and break flow 16 direction which modified the limiting large break 17 location and break flow orientation that ended up 18 increasing the limiting containment pressure. The 19 response also showed that liquid rather than a steam 20 break was the limiting small break LOCA with respect 21 to peak containment pressure. The entire LTR was duly 22 analyzed.
23 The staff also noticed that even though 24 the applicant had presented a containment nodalization 25 study for large break LOCA with four different 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
105 1 nodalization schemes most of that sensitivity study 2 was performed for the small break. So and I've 3 already showed the resulting small break LOCA 4 nodalization study revealed that even though the peak 5 containment pressure for a small break LOCA was 6 sensitive to nodalization scheme, it was always 7 smaller than the limiting peak containment pressure 8 that occurs during the large break LOCA for the 9 containment design used for the demonstration 10 analyses.
11 The staff noticed that the sensitivity 12 small break LOCA nodalization diminished in the long 13 term and the methodology bounds the containment 14 pressure by delimiting RPV pressure after the RPV in 15 containment pressure equalize. However, using both 16 the limiting RPV pressure and containment pressure for 17 SB LOCA.
18 The long-term containment response 19 evaluation raises a distinct possibility of break flow 20 reversal from the containment back into the RPV with 21 no -- with non-condensable gases from the containment 22 entering the RPV and degrading the isolation 23 condensers. This has led to the development of 24 limitation with condition No. 3 that essentially 25 requires that the BWRX-300 design has to be 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
106 1 demonstrated for no safety-significant break flow 2 reversal.
3 There were soil and water containment heat 4 transfer modeling shows that involve significant 5 differences between GOTHIC-predicted dome and PCCS 6 heat transfer reserves and the confirmatory analyses 7 reserves and the justification that was needed for the 8 flow direction-dependent condensation and natural 9 convection heat transfer biases used in the model.
10 In order to address these concerns the 11 applicant deep dived into the GOTHIC decks and 12 identified immediate error in the PCCS condensation 13 heat transfer modeling. This was a significant 14 correction to a large number of plots in the LTR that 15 led to the revision of LTR Revision 1 to LTR Revision 16 2.
17 The staff also issued an RAI on PCCS 18 modeling and performed a PCCS confirmatory study to 19 validate the PCCS sensitivity study presented in the 20 LTR. In the RAI response the applicant did mention 21 that even though they have presented the demonstration 22 analyses with a specific PCCS configuration they may 23 use a different PCCS configuration for the final 24 design. In this backdrop the staff has introduced 25 Limitation and Condition No. 4 for the applicability 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
107 1 to the final PCCS design for licensing basis.
2 So far we have described the TRACG and 3 GOTHIC codes and the evaluation methodology the 4 applicant used for the BWRX-300 design. Now this 5 slide shows a schematic of the overall approach by the 6 NRC staff followed for using the TRACE and MELCOR 7 codes to develop the confirmatory models. We 8 evaluated the methodology for calculating the 9 containment pressure and temperature response.
10 In this application the applicant has used 11 TRACG to model the reactor systems and used the 12 atmosphere as a pressure boundary condition relevant 13 to the containment vessel to calculate the mass energy 14 release through the break. The mass energy release is 15 then used as a flow boundary condition in the stand-16 alone GOTHIC containment model. Basically it's a 17 decoupled calculation between GOTHIC and TRACG codes.
18 However, to investigate the detailed 19 response of GOTHIC calculation the staff developed 20 several containment models using TRACE and MELCOR 21 codes using the same TRACG mass energy release as 22 boundary condition. The staff calls these models as 23 stand-alone models. These models include both lumped 24 models such as TRACE-CONTAIN and MELCOR-based models 25 and 3D fine-node models that is based on TRACE also.
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108 1 So there were three types of stand-alone confirmatory 2 models that we developed.
3 The staff also developed the TRACE reactor 4 system model balanced to the applicant's TRACG model.
5 Like the applicant's TRACG model the reactor system in 6 the staff's TRACE model interfaces with atmosphere and 7 calculates M&E for a stand-alone containment model.
8 The staff calls this as decoupled model. However, 9 investigate the feedback of containment pressure on 10 M&E during the LOCA transient the staff directly 11 interfaced the TRACE reactor system model with base 12 containment model which we call the decoupled model.
13 So in case of TRACE we have coupled model that we can 14 use to study any feedback that the containment 15 pressure may have for the RPV.
16 So with these three different approaches; 17 that is, the stand-alone, the decoupled, and the 18 coupled calculation, the staff was able to develop 19 insights and validate important findings that are 20 summarized on the slide.
21 In summary, the confirmatory analysis 22 showed that the proposed CE methodology based on TRACG 23 and GOTHIC codes is converted and confirmed that the 24 containment response is sensitive to nodalization, 25 break location and orientation. It demonstrated that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
109 1 the accumulation radiolytic gases in the isolation 2 condensers is possible during LOCA which has led to 3 Limitation and Condition No. 1.
4 Confirmatory analyses underscored the need 5 to have an isolation condenser return line water trap 6 which led to Limitation and Condition No. 2.
7 The confirmatory analyses also showed that 8 flow reversal is possible from the containment to RPV 9 if the PCCS heat removal capacity is not sufficient.
10 It led to the development of Limitation and Condition 11 Nos. 3 and 4.
12 MEMBER HALNON: So, on limitation four, 13 you said earlier it was a significant error. Did that 14 call into question their QA as another thing that 15 might be looking at to have a significant error 16 provided to the staff, and you guys catch it, which is 17 good, but not good on their part. Did you guys look 18 in to make sure that their QA for establishing these 19 calculations in the first place were good, or do we 20 know how the error was made?
21 MR. HAIDER: We understand what the error 22 was, and how for us to correct. And also the error 23 was a concern with -- they were under stimulating the 24 condensation heat --
25 MEMBER HALNON: Okay, but an error is 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
110 1 still an error.
2 MR. HAIDER: Yes, that's right.
3 MEMBER HALNON: So, it could have easily 4 gone the other way if it was an unknown thing. So, 5 did you ask the question how did this occur, and what 6 kind of corrective actions have you made?
7 MR. HAIDER: Yes, we did have the 8 discussions --
9 MEMBER MARCH-LEUBA: Okay Charlie, you can 10 come ahead, and get the microphone. Charlie Heck, say 11 your name.
12 MR. HECK: Yes, this is Charles Heck at 13 GEH. I want to say that this error was discovered as 14 a result of us continuing to dig into what we found as 15 sort of a discrepancy in the results. So, I think 16 it's mischaracterized to say that it was completely 17 discovered by the staff. We were both questioning 18 during this long part, we were both questioning, and 19 looking at the results.
20 But the error itself was actually 21 discovered by us as we've been trained to do, 22 questioning adding to what the results were telling 23 us.
24 MEMBER HALNON: Okay, discovered 25 nevertheless, did you guys enter this into your 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
111 1 corrective action program, and establish how it was an 2 error in the first place, and correct any process, or 3 personal performance that occurred during this 4 timeframe?
5 MR. HECK: The corrective action program 6 has to do with -- at least has to do with errors that 7 escape our release, which was not in that category.
8 MEMBER HALNON: Okay, did you still figure 9 out why it happened, and who incorrected, even though 10 you didn't have to put a corrective action program?
11 MR. HECK: Yes, we did correct -- we had 12 verification processes ongoing, and we corrected 13 those, and we continued to verify. Those results 14 ended up in the Reg 2.
15 MEMBER HALNON: All right. I'm left 16 wanting, but I'll defer to the chairman.
17 CHAIRMAN REMPE: What was the root cause 18 for the discrepancy that was an error Charlie?
19 MR. HECK: I'm going to let Necdet address 20 that, because I think he can do it better than me.
21 So, please Necdet.
22 MR. KURUL: This is Necdet Kurul, GEH. I 23 can describe the error in detail in a closed session.
24 MEMBER HALNON: I'm interested in the 25 organizational aspect, not necessarily the specific 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
112 1 detail of it.
2 MR. KURUL: It was a very subtle feature 3 in the code that we are using, we've got it. It was 4 not obvious which models use what inputs, and so it 5 was overlooked, and the results were looking fine, it 6 was more limited than we expected to see. And so the 7 results were more limiting than we expected to see 8 what they were. And so we took those results as they 9 were.
10 MEMBER HALNON: So, I'm going to just put 11 my management hat on, did you go back and check, is 12 there other features in the code that could be 13 overlooked because of subtleties, or experience of the 14 users, or any of that situation?
15 MR. KURUL: Yes, we have been looking at 16 those all along actually. This was just one that was 17 difficult to detect actually, it took us some time to 18 find out.
19 MEMBER HALNON: Okay, we can leave it 20 there, the message, I hope, has been --
21 MEMBER MARCH-LEUBA: No, let's follow up.
22 I'm thinking here, the whole industry, us included, 23 are trying to minimize costs, and maximize the 24 sureness of the time to design reactors. But was part 25 of the error the fact that you were rushed into NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
113 1 submitting this LTR? Because what you're telling me 2 is you submitted an LTR, and in power you kept 3 looking. Is there something that you can say in that 4 respect at the high level?
5 MR. KURUL: Well, no, actually we have 6 taken the time to submit the LTR. Even before we 7 submitted the LTR, we have spent a lot of time 8 developing. The reason why we didn't catch it in the 9 first place was that the results were very 10 conservative. More conservative than what we could 11 calculate by our hand calculations, and such.
12 MEMBER MARCH-LEUBA: So, you were happy 13 with the results, and didn't look further because you 14 were happy with the results, happy in the sense that 15 they were conservative. The code was calculating 16 something worse than you had expected?
17 MR. KURUL: That's correct. It's 18 basically the reason why we didn't keep thinking in 19 the first place.
20 MEMBER HALNON: I would argue the error 21 was not conservative, because it ended up in having a 22 design change. Why would we have a limitation 23 condition if it was --
24 MR. THURSTON: So -- Carl Thurston. So, 25 that particular issue was not related to design 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
114 1 change.
2 MEMBER HALNON: Okay, well it ended up in 3 a limitation.
4 MR. THURSTON: It did end up in a 5 limitation.
6 MEMBER HALNON: So, the design change 7 aside, I guess the point is when you overlook 8 something, the root cause is either skill, experience, 9 going too fast, not enough reviews, that sort of 10 thing. And I would hope that you've gone back to your 11 process, and figured out which one of those, or 12 combination of those caused this overlooking. And if 13 it was because we were happy with our results, that's 14 not a questioning attitude.
15 So, at least insufficient. So, those are 16 the types of management things I was hoping to hear as 17 we went back, and looked at it. Instead, I think all 18 we got was we corrected the error. So, I just wanted 19 to sum that message up, we don't need to continue on, 20 thanks.
21 MEMBER MARCH-LEUBA: And a message for the 22 staff, that even though the applicant is under a PSSP 23 and they have to do the reviews on the peer reviews, 24 and the reviews of the peer reviews, doing something 25 last line of defense, the staff does the final peer 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
115 1 review. So, please keep all this in mind, and keep a 2 questioning attitude.
3 CHAIRMAN REMPE: I have a question before 4 you leave, it goes along with what Greg said. It 5 sounds like -- correct me if I'm wrong, but this is a 6 certain model you had not used in the code until you 7 were trying to apply it to this design, is that a true 8 statement?
9 MR. KURUL: Yes, that's a true statement.
10 CHAIRMAN REMPE: So, it was being 11 unfamiliar with that particular model is, again, 12 probably a root cause.
13 MR. KURUL: As we mentioned, there are 14 several causes, that might be one of the causes.
15 CHAIRMAN REMPE: Perhaps, then, there 16 isn't a concern this could -- although you may have 17 other aspects of what you're trying to apply here that 18 you're not familiar with, you don't have to worry 19 about well we may have been incorrectly analyzing our 20 isolation condensers in the operating fleet, this is 21 something limited because of this application to this 22 design?
23 MR. KURUL: Yes, it is actually limited to 24 PCCS, not even the containment surface, condensation 25 on containment surfaces is not related to that either.
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116 1 CHAIRMAN REMPE: Okay, thank you.
2 MR. HECK: This is Charlie Heck, I'd like 3 to also interject that there was a component to this 4 related to how you do the nodalization of the 5 containment in how the PCCS units are placed within 6 that nodalization. So, the error that was discovered 7 manifests itself differently depending on the 8 nodalization. So, that's another subtlety that made 9 it difficult to find.
10 MEMBER MARCH-LEUBA: Okay, I think we've 11 peeled the topic enough. So, please continue.
12 MR. HAIDER: Another thing that the staff 13 would like to add is that after they made the 14 correction, they resubmitted the reserves, and they 15 communally brought the predictions, the GOTHIC 16 predictions. So that's something that I would like to 17 bring into notice. That confirmatory analysis did 18 play a role in finalizing this issue, around this 19 issue.
20 MEMBER MARCH-LEUBA: And not finding it, 21 but confirming the solution was likely correct.
22 MR. HAIDER: That's right, because it was 23 the confirmatory analysis in the first place that had 24 led the staff to think that there was something that 25 was not right, because the heat transfer and PCCS were 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
117 1 not matching between the confirmatory predictions, and 2 the GOTHIC predictions, but then they came closer.
3 MEMBER MARCH-LEUBA: So, now I will ask 4 Carl Thurston to describe the first two RPV specific 5 limitation conditions.
6 MR. THURSTON: Yeah, so this is Carl 7 Thurston again. So, these limitation conditions were 8 added to the SER, and I explained the reasons why we 9 added them. I don't think we need to go through word 10 by word, but the first one again is related to the 11 modeling of radioactive gases. And again, I reiterate 12 that the isolation condensers are the primary, the 13 only method of K heat removal.
14 So, they have to work, and any degradation 15 of the ICS performance is a significant issue. The 16 second issue is --
17 MEMBER MARCH-LEUBA: I know we're a little 18 late, but I'm going to make it even later. When you 19 said the ICS is the only method of K heat removal that 20 we take credit for in passive mode. In reality the 21 glow was open and valves and used the main condensers, 22 right? And especially when we look at risk analysis, 23 we'll have all the branches of the variations.
24 MR. THURSTON: That's correct. But for 25 our analysis because we used the safety systems --
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118 1 MEMBER MARCH-LEUBA: Chapter 15, you're 2 100 percent correct. And one concern about the 3 Chapter 19, you have to do everything.
4 MR. THURSTON: Right. So, like I say, 5 both of these limited conditions are based on 6 potential degradation of heat transfer to the ICS.
7 And so the second one again is related to the piping 8 design for the ICS return line to make sure again, 9 that the gases are not able to escape, and go back to 10 the seam, back into the ICS tubes. So, with that I'll 11 turn it -- go to the next slide, and turn it over to 12 Syed.
13 MR. HAIDER: Thank you. So, this slide 14 provides the whole containment specific limitation 15 condition number three, and four that the staff had 16 imposed on the application. And the VWR standard 17 containment evaluation methodology as observed on the 18 staff review, limitation condition number three 19 emerged out of the potential degradation of the 20 isolation condenser heat unit for performance should 21 there be a reverse flow with non-condensables from the 22 containment to the RPV.
23 Such a flow reversal could take place in 24 the PCCS if not properly sized to deal with the most 25 limiting flow reversal conditions. However, at the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
119 1 licensing stage, the applicant would need to 2 demonstrate that either no reverse flow could occur, 3 or any reverse flow that occurs under the most 4 bounding flow reversal conditions is not safety 5 significant with respect to the methodology's 6 acceptance criteria.
7 And limitation number four requires that 8 if any item of PCCS, our design configuration and 9 placement is used on the VWRX300 design at the 10 licensing state other than the one presented in this 11 LTR, the applicability of this rapid, and the PCCS 12 modeling approach would have to be reviewed, and 13 approved by the NRC.
14 MEMBER MARCH-LEUBA: And you don't mean by 15 this that sizing the PCCS a little larger with the 16 same configuration type, that's not what this 17 limitation condition applies. If you go with an air 18 cooled system, completely different, is that correct?
19 MR. HAIDER: Yes, that's my understanding, 20 or a different geometric design. Next slide, please.
21 In this presentation we have discussed the later 22 design issues, and compare records involved in the 23 staff review of the methodology. And finally we 24 present the staff conclusions for our LTR review. In 25 summary, the proposed VWRX chamber in analytical 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
120 1 approach, and the type G and GOTHIC modeling of the 2 mass and energy released and containment response 3 are acceptable with the appropriate conservative 4 biases used in the LTR.
5 With the four limitations, and conditions 6 specified in the SER, evaluation methodology presented 7 in the LTR revision two is acceptable for the VWRX 8 standard heat containment, pressure, and temperature 9 analysis, for the design, this is excellent. The NRC 10 staff will be required to evaluate the regulatory 11 compliance of the final VWRX standard containment 12 design using the containment evaluation methodology in 13 future licensing activities.
14 In accordance with 10 CFR Part 50, or 10 15 CFR Part 52 as applicable in the application. This 16 concludes the staff presentation, thank you very much 17 for your attention. Now the staff would like to 18 address any additional questions the committee members 19 might still have.
20 MEMBER MARCH-LEUBA: Yes, I have one 21 question above your pay grade. I like the approach of 22 going with a topical report, so by the time we're done 23 with the topical SERs, all the hard questions have 24 been answered. So, we are going to -- I suspect we're 25 going to have a very accelerated review of the SAR.
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121 1 Is there any planning -- are we going to have a six 2 month review of an SAR? Or it's certainly not going 3 to be six years.
4 MR. HAIDER: I would like to defer this 5 question to Becky, or --
6 MS. PATTON: I was going to defer the 7 question to Mike --
8 MEMBER MARCH-LEUBA: Becky, state your 9 name.
10 MS. PATTON: Rebecca Patton. Mike, do you 11 want to take that one, or do you want me to? Okay, 12 all right. So, I will say that there continues to be 13 a working dialogue I think, between -- within the 14 projects branch, and we obviously in the tech employee 15 area have been put into that as well. And it takes 16 into account the applicant's needs. But it's my 17 understanding none of that has been set, or decided 18 on.
19 And obviously we don't determine a review 20 schedule until an application is actually tendered.
21 So, there's some initial planning based on scope of 22 changes for budgeting, and other considerations. But 23 the actual schedule doesn't get set until after then.
24 MEMBER MARCH-LEUBA: But some interaction 25 with the projects branch, or section, or whatever they 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
122 1 are, is really worthwhile, because if the technical 2 staff is expecting that they would need to perform 3 confirmatory of the final number crunching, you cannot 4 do a three month review.
5 MS. PATTON: Yeah, absolutely. Mike 6 Dudek, and I talk very frequently, as well as the 7 other branch chiefs, and --
8 MEMBER MARCH-LEUBA: Unfortunately you've 9 been on the receiving end of this, that products tells 10 you you've got until October, and I want your SER 11 then. And it'd be best to be productive, if you 12 anticipate something to have a backup, to have 13 planning for the review.
14 MS. PATTON: Right, we do that even years 15 in advance right, when we first hear it comes down.
16 I'm just letting you know, that does happen on a very 17 frequent basis where we have dialogue.
18 MEMBER MARCH-LEUBA: Excellent, anymore 19 questions from the members?
20 MEMBER HALNON: Yeah, just a comment.
21 Notwithstanding my rant earlier, the topical report, 22 and the SER were very well done. I want to say you 23 guys, something that I could understand, which was not 24 easy, that it's written, from my perspective, 25 comprehensively.
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123 1 MEMBER MARCH-LEUBA: Thank you.
2 VICE CHAIRMAN KIRCHNER: Just an 3 observation Jose, that when the actual application 4 comes in, one would want to look at a spectrum of 5 break locations, and sizes such that you could ensure 6 that no ingestion of nitrogen from this inert 7 containment degrades the performance of the ICS.
8 MEMBER MARCH-LEUBA: So, the staff 9 recognize, I think we all recognize that the ingress 10 of non-condensable nitrogen into the vessel would be 11 really bad. So, that's again, what I was talking 12 about, about reserving some time for the review.
13 Because the temptation will be to run only the 14 limiting local that they've already found, where the 15 final design is still doing the whole break spectrum 16 of scenarios.
17 And it's something that you guys, now that 18 you have the review fresh in your mind, it would be 19 good if you provide yourself a to do note, and tell 20 Mike this is going to take this many months. Anymore 21 comments, questions? While you think about it, I'm 22 going to open up the line for the public, members of 23 the public. Anybody wants to make a comment, not 24 questions, but comments.
25 Please do so this moment. If you are in 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
124 1 teams, just unmute yourself. If you are on the phone 2 line, you need to use star six. Five second rule, we 3 have no questions. Chairman, the floor is yours.
4 CHAIRMAN REMPE: Thank you. At this point 5 we're going to go off the records, and we'd like the 6 court reporter to come back for the Point Beach 7 subsequent license renewal application. And how about 8 we take a nine minute break, and come back at 2:15 9 east coast time to read in your draft letter? Okay, 10 so we're going to recess.
11 (Whereupon, the above-entitled matter went 12 off the record at 2:07 p.m. and resumed at 3:00 p.m.)
13 CHAIRMAN REMPE: Okay, it's 3:00 p.m. here 14 on the east coast, and we're back in session, and on 15 the record. I'd like to call on Member Matt Sunseri 16 to lead us through the Point Beach subsequent license 17 renewal discussion today. Matt?
18 MEMBER SUNSERI: Thank you Chair Rempe, 19 and good afternoon members, and attendees. This 20 session is for the Point Beach subsequent license 21 renewal application review. I'm happy to say this is 22 the first time that we are exercising our one step 23 versus two step review process for subsequent license 24 review applications that don't have any open items, 25 confirmatory items, or other issues requiring a more 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
125 1 in depth inquiry.
2 The way this will flow, is today we will 3 hear from the applicant NextEra, and staff on their 4 safety review. Following the presentation, we'll ask 5 for any public comments, then the committee will 6 discuss any significant elements that should go into 7 our letter report. At our May full committee meeting, 8 I will present a draft letter report for the final 9 deliberation by the committee.
10 So, are there any questions about the way 11 this will proceed today? All right, two quick points 12 before I turn the floor over to VNR, excuse me. First 13 off, I want to thank the staff, and the applicant for 14 the thorough, and comprehensive reports, and material 15 that they've prepared for this review, and this 16 meeting. Clearly the experience base with subsequent 17 license renewals is continuing to mature, and the 18 quality of the material is excellent, and I'm really 19 looking forward to today's presentations.
20 Secondly, due to some outside work that I 21 do with a private company, I will recuse myself from 22 any deliberations on material analysis of the primary 23 system components associated with this review. At 24 this time, I now turn to Mr. Brian Smith, division 25 director for New and Renewed Licenses for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
126 1 introductions, and the presentations. Brian, the 2 floor is yours.
3 MR. SMITH: All right, thank you Member 4 Sunseri. And good afternoon Chairman Rempe, and 5 members of the ACRS. As Member Sunseri said, I am 6 Brian Smith, the director of the Division of New and 7 Renewed Licenses in the Office of Nuclear Reactor 8 Regulation. We appreciate the opportunity today to 9 present to the ACRS the results of the staff's review 10 on the fifth application for subsequent license 11 renewal.
12 This application was submitted by NextEra 13 Energy for the Point Beach Nuclear Power Plant Units 14 1 and 2, located in Two Rivers, Wisconsin. By way of 15 the background, Point Beach Units 1 and 2 received 16 approval for their initial license renewal from the 17 NRC on December 22nd, 2005. The NRC review at that 18 time was performed using the guidance contained in the 19 initial issuance of the generic again lessons learned 20 report, or the GALL report.
21 The NRC guidance for license renewal, and 22 subsequent license renewal over the years has evolved 23 through enhancements, and improvements based on the 24 lessons learned from NRC application reviews, and from 25 consideration of both domestic, and international 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
127 1 industry operating experience. The initial GALL 2 report for license renewal went through two revisions.
3 Additional interim staff guidance issued following 4 revision two.
5 GALL revision two, along with the ISGs in 6 place at that time was used in the activities to 7 develop the guidance for subsequent license renewal 8 that's contained in the GALL SLR report. In addition 9 to the previous license renewal guidance, the GALL SLR 10 report included additional focus on aging management, 11 and time limiting aging analyses for operation in the 12 60 to 80 year time period.
13 The staff recently initiated work to 14 develop the first revision of the GALL SLR report. We 15 will incorporate the current four ISGs that have been 16 issued. We'll include updated new, or clarified staff 17 positions, and we'll add improvements that have been 18 identified during previous reviews. We will interact 19 with the ACRS as we proceed throughout this process.
20 The NRC project manager for the Point 21 Beach subsequent license renewal application review is 22 Bill Rogers. Bill will introduce the staff who will 23 be presenting, and addressing the questions regarding 24 the statute review. Part of the management team here 25 with me today is Lauren Gibson, chief of the License 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
128 1 Renewal Projects Branch, and branch chiefs for the 2 staff involved in the technical review, including Joe 3 Colaccino, Steve Bloom, and Angie Buford.
4 Dr. Allen Hiser will also be available to 5 address questions from the committee. In addition, 6 John Bozga, Region III senior reactor inspector will 7 discuss the regional inspection activities. I'd like 8 to note that the staff completed its review with no 9 confirmatory, or open items in the safety evaluation.
10 Finally, we will address any questions you may have on 11 the staff's presentation, and we look forward to a 12 productive discussion today with the ACRS.
13 At this time I'd like to turn the 14 presentation over to Mr. Bill Maher, NextEra director 15 for nuclear licensing projects to introduce his team, 16 and commence the presentation. Mr. Maher?
17 MEMBER SUNSERI: You're on mute.
18 MR. MAHER: Trolling. All right, Steve 19 had a couple of buttons, I was told I'm muted, sorry.
20 All right, I appreciate the time for both ACRS, and 21 staff enable for us to be able to prevent the results 22 of the staff's review on the subsequent license 23 renewal for Point Beach. Next slide. So, we're on 24 slide two. So, we'll go through introductions as we 25 go, once I go through the agenda here, and we'll get 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
129 1 started.
2 So, next will be a performance philosophy 3 with Mike Strope, who is online also, as the site vice 4 president he'll present that to ACRS. Then Steve 5 Franzone will go through the site overview, and then 6 that'll turn over to Steve Hale, who will do the 7 process that we used for subsequent license renewal 8 evaluation. Also, there'll be specific topics of 9 interest that we'll talk through, and then we'll go 10 through closing remarks.
11 Again, I'd like to appreciate the staff's 12 time, and the thorough review that they've done for 13 us. What we're going to present to you is the results 14 of their review, as well as what Point Beach is doing 15 to maintain adequate safety margins for the subsequent 16 period of extended operation. So, my name's Bill 17 Maher, I'm the senior director of licensing for 18 nuclear projects for NextEra Point Beach. Mike, would 19 you like to introduce yourself?
20 MR. STROPE: Good afternoon. I'm Mike 21 Strope, I'm the site vice president at Point Beach.
22 MR. FRANZONE: And I'm Steve Franzone.
23 I'm the licensing manager for subsequent license 24 renewal for Point Beach Units 1 and 2. And then Steve 25 Hale?
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130 1 MR. HALE: Hi, my name is Steve Hale, I'm 2 one of the technical leads for ENERCON for the license 3 renewal process. I had the same role for the Turkey 4 Point SLR. Prior to that I was at NextEra for over 46 5 years, had several roles including the engineering 6 director at Turkey Point, and I was directly involved 7 with Turkey Point, and St. Lucie original license 8 renewal, as well as the Point Beach ECU.
9 MR. FRANZONE: Thank you. Maribel Valdez?
10 MS. VALDEZ: Yes, good afternoon, my name 11 is Maribel Valdez, and I am the reactor vessel 12 internals program owner.
13 MR. FRANZONE: Thank you. And Anees 14 Udyawar?
15 MR. UDYAWAR: Good afternoon, my name is 16 Anees Udyawar, I work at Resting House on fraction 17 mechanics, and I'll make a brief presentation on the 18 RDO structural steel supports, thank you.
19 MR. FRANZONE: Okay, and back to you Bill.
20 MR. MAHER: Thank you. And I'd like to 21 introduce Mike Strope to give some opening remarks, 22 and we'll go to slide three.
23 MR. STROPE: Good afternoon everyone, 24 appreciate the opportunity to speak to the committee.
25 Point Beach is one facility of the NextEra fleet we 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
131 1 operate for our nuclear excellence model. We have the 2 desire to be a team that delivers consistent, and 3 excellent performance on a sustainable basis. We 4 execute on our values based on our core principles, 5 and we do so in four strategic focus areas of 6 operational excellence, organizational effectiveness.
7 Generation reliability, and effective 8 business, and financial performance. At Point Beach 9 specifically, we use the slogan that Point Beach is a 10 team, it's not a place. And we execute on those 11 strategic focus areas through three site focus areas, 12 those being people, behaviors, and never satisfied.
13 We want to engage our people, and build the best team 14 that we possibly can.
15 We're very focus on behaviors, both 16 meeting our standards, and also as well as being never 17 satisfied on what we do, and knowing that we can 18 always do it better. Next slide, please. The 19 information here I will not read to you, but our plant 20 performance, based on our plant capacity factor, has 21 been sound. The areas with lower percentages are 22 related to when we have had planned refueling outages.
23 Our last reactor plant trip was associated 24 last summer. Other than that we've had breaker to 25 breaker runs for the last two outages on both units.
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132 1 Our regulatory status is as listed, and noted in 2 previous conversations when we're introducing the 3 presentation. Thank you.
4 MR. FRANZONE: Yes, good afternoon again, 5 this is Steve Franzone. Again, I'm the licensing 6 manager for subsequent license renewal. Just to note, 7 please stop us at any time during the presentation if 8 you have questions, and as a reminder for our team, 9 please introduce yourselves when you speak. This 10 slide, as well as the next few slides provides some 11 facts, figures, and photos which will provide a brief 12 overview of the site.
13 Turn to slide six now. Point Beach Units 14 1 and 2 are located on the shores of Lake Michigan, 15 near the city of Two Rivers, Wisconsin. We are one of 16 the largest single sources of electricity in 17 Wisconsin. For example, we generated about 14 percent 18 of all the electricity created in Wisconsin in 2019, 19 and powered 950000 homes, and businesses. The star in 20 the center of the circle marks the location of the 21 site.
22 It is located approximately 29 miles 23 southeast of Green Bay, Wisconsin. Point Beach has 24 two Westinghouse two loop NSSS systems, as well as 25 Westinghouse electric turbine generators. The cooling 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
133 1 water is pulled from Lake Michigan. The black dashed 2 circle represents the 50 mile radius from the plant, 3 and finally, as you can see on the figure, that the 4 decommissioned Kewaunee Power Station is located due 5 north of the site.
6 Please turn to slide seven. This photo 7 provides a view of the entire site. North is to the 8 top of the figure, Lake Michigan is on the right hand 9 side of the photo.
10 A little history about the site. Site 11 construction started in 1967 and 1968, Units 1 and 2, 12 respectfully. Unit 1 received its operating license 13 in October of 1970, followed by Unit 2 in March of 14 1973. The Unit 1 steam generators were replaced in 15 1983; likewise, the Unit 2 steam generators were 16 replaced in 1996.
17 This figure is taken from the 18 environmental report. The red line identifies the 19 exclusion area boundary, which coincides with the site 20 boundary. With the dotted line in the inside circle 21 marks the protected area boundary.
22 The site buildings are located in the 23 light red, the Point Beach independent spent fuel 24 storage installation is located northwest of the 25 units, as you can see. I'll turn to slide eight.
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134 1 Point Beach initial license renewal was approved in 2 December of 2005, and was based on GALL Rev Zero. One 3 of the lessons learned from our experience at Turkey 4 Point was to conduct the NEI aging management 5 effective review early.
6 In fact, the site completed the effective 7 review in 2018, which helped our project team complete 8 our feasibility study in 2019. At the risk of 9 stealing Steve Hale's thunder, on slide 18 the review 10 concluded that all AMPS continue to be effective with 11 no failed elements. He will also be discussing 12 results of the phase four inspection on the same 13 slide.
14 One challenge we will be facing is, given 15 that the current license for Unit 1 expires in 2030, 16 Unit 1 will already be in a few years into the 10-year 17 pre-window of the subsequent period of extended 18 operation, if we obtain approval of the subsequent 19 license renewal application in the next couple of 20 years.
21 I'll turn to slide nine. Slide nine and 22 ten provide a listing of modifications and upgrades 23 for the plant since the first license renewal.
24 Although not a comprehensive list, we identify some of 25 the major modifications to the unit. As you can see 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
135 1 by these two slides, we have made significant 2 investment in various plant systems, and components to 3 allow the safe, and efficient operation of the units 4 into the future. For example, we added two new motor 5 driven auxiliary feed water pumps, but turned the 6 existing auxiliary feed water motor driven pumps into 7 additional non-safety backups.
8 Thereby improving system reliability, they 9 are now referred to as the standby feed generator 10 pumps. I'll pause a moment for questions before we go 11 to the next slide.
12 MEMBER HALNON: Steve, this is Greg 13 Halnon, you mentioned the AMPS continue to be 14 effective. I was struck by how many revisions to the 15 AMPS were made in the table. And many of those 16 revisions look like they already should have been into 17 a decent, good program. Can you help me with that 18 perception, is that incorrect, or? Maybe you can 19 answer the more direct question, why were there so 20 many revisions required?
21 MR. FRANZONE: I'm not sure I quite 22 understand the question, are you saying --
23 MEMBER SUNSERI: It's exceptions is what 24 you're talking about Greg, right?
25 MEMBER HALNON: Right.
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136 1 MEMBER SUNSERI: There's a lot on here 2 with exceptions.
3 MEMBER HALNON: And just the table, I 4 guess in the SER.
5 MR. FRANZONE: We have a slide, I can go 6 to that slide, or if you want we can hold until we get 7 to that slide, and we can follow up.
8 MEMBER HALNON: We can hold, that's my 9 lingering question. We can hold, if it seems 10 appropriate I'll ask it later.
11 MR. FRANZONE: Yeah, when we get to that 12 slide we can talk to that.
13 MR. HALE: Yeah, I will address that as 14 well.
15 MEMBER SUNSERI: So, Steve, one of the 16 things that was mentioned earlier is you replaced the 17 steam generators, can you go over those dates again, 18 when the steam generators were replaced?
19 MR. FRANZONE: Yes. The steam generators 20 were replaced -- Unit 1 was replaced in 1983, while 21 Unit 2 was replaced in 1996. The Unit 2, actually, at 22 the time in 1983, they did a sleeving, I guess, versus 23 a replacement at the time. And that allowed them to 24 continue operating until 1996. And overall, the steam 25 generators have very few -- two plugs for each unit.
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137 1 I want to say it's like 14 total on Unit 1, and I 2 think five or six total on Unit 2, total.
3 MEMBER SUNSERI: Thank you.
4 MR. FRANZONE: Yeah, five total on Unit 2, 5 I'm looking at the numbers now. And we'll go to slide 6 11 now. Yeah, slide 11. The extended power up rate 7 was a multi-year process which culminated with an 8 approximately 17 percent thermal power up rate in 2011 9 for both units. As you can see, we had quite an 10 extensive list of modifications.
11 In fact Steve Hale, one of the ENERCON 12 technical leads, was actually our NextEra lead for the 13 Point Beach extended power upgrade, and is familiar 14 with many of the issues associated with such a large 15 effort. And we'll go to slide 11. Preparation for 16 the application started in 2019 with a detailed 17 feasibility study, and we submitted the application in 18 November of 2020.
19 The project team has many years of license 20 renewal, and Point Beach experience. This is 21 essentially the same team which successfully completed 22 the Turkey Point subsequent license renewal.
23 Additionally, our team had our third application for 24 St. Lucie Plant accepted for review last year, and we 25 have just finished our break out session for that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
138 1 project.
2 The multi discipline team consists of 3 ENERCON as the lead preparer for the submittal, as 4 well as Westinghouse, Framatome, and Structural 5 Integrity Associates in supporting roles. Also the 6 project team was supported by fleet, and site program 7 owners, as well as various subject matter experts, 8 such as Scott Boggs, who is the owner's group 9 representative.
10 Or Maribel Valdez, who is our reactor 11 vessel internals program owner. Every aging 12 management program for SLR was actually assigned a 13 program owner to support a portion of the application 14 preparation in NRC review. The project team generated 15 over 100 reports which supported the application.
16 These reports not only provide the next level of 17 detail for various aging management programs, and 18 other parts of the application.
19 They will also provide a way to ensure the 20 knowledge will be passed on to personnel who will need 21 it in the future. As part of our implementation 22 process, these reports will be incorporated into the 23 plant's licensing basis as control documents. The 24 technical leads for our project partner, ENERCON, are 25 both former NextEra employees with almost 80 years of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
139 1 combined experience with NextEra.
2 Both developed the original license 3 renewal applications for Turkey Point and St. Lucie.
4 In addition, we have to establish a subsequent license 5 renewal lane of opposition now in order to ensure we 6 will successfully transition to the subsequent period 7 of extended operation. I will turn the presentation 8 over to Steve Hale, who is one of the technical leads 9 I just talked about. We'll go to slide 12.
10 MR. HALE: Thanks Steve, this is Steve 11 Hale again here for ENERCON. For the Point Beach 12 subsequent license renewal application, we followed 13 the guidance of NEI 17-01, which was developed 14 specifically for subsequent license renewal. We also 15 reviewed REIs, and responses from first three SLR 16 reviews, which included Turkey Point, Surry, and Peach 17 Bottom.
18 Other activities we conducted to ensure 19 quality, subsequent license renewal application 20 included extensive interviews with that AMP owners 21 both on site, and at the fleet level, and in several 22 pre-application meetings with the NRC. As Steve 23 indicated, the input from the program owners is very 24 important to ensuring we have an informed review of 25 the operating experience at the site, as well as 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
140 1 ongoing efforts with regards to AMP implementation.
2 Our approach going in was to comply with 3 NUREG 2191, and 2192 to the greatest extent possible, 4 and we believe we have been able to accomplish that.
5 We, along with the NRC, worked diligently to ensure we 6 met the SLR goal of an 18 month safety review 7 schedule, and this was accomplished with issue of the 8 SER without any open items in February. Go to slide 9 13 Steve.
10 Having been involved with original, and 11 now subsequent license renewal, I have a unique 12 perspective as to what is involved with the integrated 13 plan assessment, or IPA for both efforts. We thought 14 the best way to present our methodology for Point 15 Beach was really to discuss the differences between 16 original license renewal, and subsequent license 17 renewal.
18 For scoping, and screening, there were 19 minimal challenges, because the criteria in 10 CFR 20 Part 54 really hasn't changed a lot. We do have to 21 address the modifications, as you saw in the slide 22 Steve presented were quite extensive, and we also 23 needed to do some updates relative to what we call the 24 A2 scoping, and screening criteria that has to do with 25 non-safety, which can affect safety.
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141 1 And this has to do with the fact that when 2 Point Beach originally got their license renewal 3 approval, there was an absence of specific guidance 4 documents related to alpha two scoping, and screening.
5 As we moved into the AMRs, and AMPs, you really see 6 where the differences between license renewal, and 7 subsequent license renewal reside. Point Beach, as 8 Steve mentioned, was a GALL Rev Zero plant.
9 And with the issue of GALL Rev One, Gall 10 Rev Two, subsequent ISGs, and GALL SLR, there are a 11 number of aging effects we needed to address beyond 12 what was addressed originally. The most significant 13 differences however are in the aging management 14 programs. Point Beach originally had 27 aging 15 management programs, and for subsequent license 16 renewal, we're up to a total of 48.
17 If you go to slide 14 Steve, you'll 18 provide some of the specifics there. With regard to 19 our consistency with GALL, we were over 98 percent 20 consistent with the over 2500 line items in the GALL 21 AMR tables. And of the 48 aging management programs 22 for SLR, 9 are new, and 39 are existing, and all were 23 evaluated against the GALL AMP requirements on an 24 element by element basis.
25 Differences were addressed either through 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
142 1 proposing enhancements that were captured as 2 commitments, or exceptions if necessary. There are 3 ten AMPs with exceptions to GALL, all of which have 4 been reviewed, and accepted by the NRC. Most have to 5 do with specific design features, or characteristics 6 at Point Beach, which require taking an exception, and 7 I'll get into that in the next couple of slides.
8 There were no plant specific AMPs. One of 9 the things we did, as I mentioned earlier, was 10 addressing RAIs. In each of our AMP basis documents, 11 which are the technical documents that support each 12 AMP, we have a separate section specifically 13 addressing the RAIs for Turkey Point, Peach Bottom, 14 and Surry. And those documents were available to 15 staff to assist them in their technical review.
16 We'll go to slide 15 Steve.
17 MEMBER SUNSERI: Let me ask a question 18 here. I think Greg, is this where you had your 19 question earlier?
20 MEMBER HALNON: Yeah, I guess in the 27, 21 if you go back one slide, 27 programs consistent with 22 enhancement, it's not told whether that number is the 23 significant number of enhancements required on some of 24 the programs. And some of them just struck me as 25 something that should have already been there. For NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
143 1 example, the fire protection AMP, one of them is 2 enhanced plant procedures that indicate that personnel 3 performing fire protection inspections would be 4 qualified to do so.
5 That just seems rudimentary, that it 6 should have already been there. Another one is, you 7 specify that penetration seals will be inspected for 8 indications of increased shrinkage, and lesser strain, 9 it seems like some of those should still just be 10 there. And there's some cross talk between this one, 11 and the external services. It just struck me that 12 there was -- I guess my perception was it felt like 13 the previous AMPs were not sufficient on some of the 14 most rudimentary type issues.
15 And that was my perception, I'm not saying 16 that I did enough research to obviously come to a 17 conclusion, but I was hoping that you could help me 18 with that perception.
19 MR. HALE: Yeah, and what you find that 20 not all programs have all the prescriptive 21 requirements stipulated in GALL. When we do our 22 reviews on an element by element basis, we delve into 23 what the GALL is telling us we need to do, and review 24 it against the specific implementing procedures to 25 make sure those particular aspects are captured. In 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
144 1 some cases, if it's not rigorous, or not specifically 2 called out for, we flag that as an enhancement to 3 ensure that's captured as a commitment for subsequent 4 license renewal.
5 So, the programs, as they are currently at 6 the site, existing programs, of course were never 7 lined up against the specific GALL requirements. In 8 some cases we may be doing that, but it's not 9 specifically detailed enough, we felt, and the program 10 owners felt in the program commitments. We wanted to 11 make sure that they were done, and they're good for 12 the rest of the plant operation life.
13 MEMBER HALNON: So, repeating back in 14 short hand, the change from GALL zero, to subsequent 15 license renewal GALL, plus the ISGs, plus just doing 16 a lot more rigorous job on it.
17 MR. HALE: That is correct, and it really 18 provides a benefit to us, because we have the AMP 19 guideline that we need to address ourselves to. And 20 in a lot of cases, some of these aging management 21 programs we call existing may not have had a GALL 22 program description beforehand. And so that really 23 gives us the guideline that we need to follow, and to 24 make sure that the commitments we capture are good 25 from here on out.
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145 1 MEMBER HALNON: Okay, that's fair, thank 2 you.
3 MR. HALE: Okay, next slide Steve, I 4 believe that's 15. I just wanted to show these, these 5 are what we are calling the new AMPs, because these 6 specific AMPs were not around when Point Beach went 7 through license renewal originally. And you'll notice 8 there are, a majority of them electrical. Because 9 when we went into license renewal originally, there 10 was very little in the electrical area.
11 So, this has highlighted a number of 12 activities, and new AMPs that we'll be implementing 13 for the SBEO. Now to slide, I believe it is Steve, 14 slide 16. Now, to address your comments with regards 15 to exceptions to GALL, we've highlighted some of the 16 specifics here. For example reactor head closer sub 17 bolting, there is a preventative action in GALL that 18 says the bolting should not be high strength from a 19 concern over stress corrosion cracking.
20 Because the records are not specifically 21 clear as to what the ultimate strength of the studs 22 were, we just defaulted to the conservative, and 23 assumed they were all high strength. And these bolts 24 are looked at under section 11, every outage 25 regardless. However, because the GALL calls out that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
146 1 preventative action, we took the exception. Another 2 one that may be interesting to you, if you look under 3 water chemistry.
4 Because we have steam heating around the 5 plant, and that's fed by a boiler to ensure that 6 chemistry is maintained appropriately, and this has to 7 do with a two over one concern, and there is no 8 specific requirements regarding boiler standard 9 heating chemistry. We took an exception to the 10 specific requirements in the EPRI document to 11 secondary steam quality, and ensured that we were 12 following the appropriate boiler standards.
13 This was also consistent with something 14 that Peach Bottom did. Any other questions here?
15 This is five of the AMPs with exceptions? If there 16 are none, the next slide has the remaining exceptions.
17 One of these is fuel oil chemistry, ENERCON has had 18 quite a bit of extensive experience with 19 implementations at the various sites. And one of the 20 challenges with fuel oil chemistry is some of the 21 smaller tanks.
22 Like the skid tanks on the diesel cannot 23 be specifically drained fully, and to do a complete 24 internal inspection. So, based on our experience 25 there, we've requested an exception to that. However, 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
147 1 we are doing the fullest extent possible. And then, 2 under the GALL requirements for reactor vessel 3 material surveillance, if you have to change your 4 capsule removal schedule, that's an automatic 5 exception to the reactor vessel material surveillance 6 aging management program.
7 Now, I hope I've addressed our questions 8 regarding the AMPs exceptions. Typically it has to do 9 with the certain design features at the plant. And 10 when you line up yourselves with the GALL 11 requirements, it's just not possible to comply with 12 them. Any other questions there?
13 MEMBER SUNSERI: Members, any?
14 MR. HALE: Okay. Moving on to slide 17 --
15 I'm sorry slide 18, right? When you look at the 16 override, and Steve said he was stealing my thunder 17 with this, we did conduct an effectiveness review at 18 the site of all the AMPs in May of 2018. This was the 19 lessons learned at Turkey Point, that before you 20 embark on a subsequent license renewal effort, you 21 really need to do the effectiveness review.
22 It gives you -- I'll call it a benchmark 23 as to how well you're doing. And if there are some 24 issues, you can identify those before you go ahead 25 with a subsequent license renewal project. There is 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
148 1 guidance out there from NEI, it's 14-12. All the AMPs 2 were reviewed regarding those requirements, and all 3 were determined to be effective, with none of the ten 4 elements under each AMP -- none of the elements were 5 considered to be failed.
6 Also beneficial is the NRC performs a 7 phase four inspection, that's typically done after 8 five years of operation into the period of extended 9 operation. The NRC came in, and they looked at seven 10 of the specific AMPs at Point Beach, and no findings 11 were identified with that inspection. Slide 19 Steve?
12 I'm sorry, slide 20. No, 19, I'm sorry. Okay, for 13 commitments.
14 We have 51 total for Point Beach. These 15 will be maintained separate from current license 16 renewal, and that's really for clarity, and to avoid 17 confusion. There are two license conditions 18 specifically related to Point Beach in the SER. Those 19 are to incorporate a supplement into the UFSAR, that's 20 already prepared, and ready to go in once we get the 21 subsequent renewed licenses.
22 And to implement all of the programs, and 23 complete activities including pre-FPO inspections 24 prior to the subsequent period of extended operation.
25 There will be a new chapter, 16, to the UFSAR, which 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
149 1 as I've indicated previously, will be maintained 2 separately. And included in that new chapter is a 3 complete table of all the commitments which will be 4 maintained until they're all implemented.
5 The project team, both ENERCON, and 6 NextEra have a lot of experience with commitment 7 management, and expectation, so there's a high level 8 of expectation that these will all be accomplished 9 within the required time period. Slide 20. Now, 10 that's all I was going to discuss with aging 11 management programs. Are there any questions?
12 MEMBER SUNSERI: Members, any questions 13 with the aging management? No, please go ahead.
14 MR. HALE: Okay, I'll now move into the 15 time limited aging analyses. We did have the 16 dispositions of I, II, and III, some of them did 17 change for Point Beach. And this primarily has to do 18 with the fact that they've issued several GALL AMPs to 19 address specifically some of the TLAAs, like fatigue.
20 And as a result of moving that way some of your 21 analytical dispositions have moved to more of an aging 22 management disposition.
23 One of the challenging efforts we have is 24 addressing environmentally assisted fatigue. There 25 were a number of vendors involved with that, and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
150 1 updates there were primarily due to the fact that 2 there's a number of the guidance documents that have 3 been issued since original license renewal, and we had 4 to adjust ourselves to those updated requirements. We 5 followed the same process of searching the current 6 licensing basis to establish the time limited aging 7 analyses for Point Beach.
8 We did not identify any new ones for 9 subsequent license renewal. Can you move to slide 21 10 Steve? The next two slides present all of the TLAAs 11 for Point Beach for SLR. Rather than going into them 12 specifically, we thought that we would leave it up to 13 the committee to pose any questions they might have, 14 and please recognize as you can see, there were a 15 number of vendors involved.
16 And some of their specific reports are 17 considered proprietary information. So, if we delve 18 into a high level of detail into some of these, we 19 might require a special session for the proprietary 20 information.
21 MEMBER SUNSERI: I think we're good with 22 this. You have a selected set of topics of interest 23 that I think are going to be most interesting to the 24 committee. So, if we can kind of speed up a little 25 bit, and get to those topics, I think it would be 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
151 1 beneficial.
2 MR. HALE: Okay. So, with that I guess 3 I'll turn it back over to you Steve.
4 MR. FRANZONE: Thank you Steve. So, now 5 we have four topics that we selected, reactor vessel 6 internal, and then irradiated concrete, and supports 7 are kind of like one, and the same topic. And then 8 the last is the epoxy resin based grout. And so at 9 this point I'll turn it over to Maribel Valdez, who 10 will talk about reactor vessel internals. Maribel?
11 MS. VALDEZ: Thank you Steve. Good 12 afternoon again, my name is Maribel Valdez, and in 13 this slide I'll discuss the proposed analysis 14 regarding the gap analysis for the reactor vessel 15 internals aging management program. The site RVI 16 program was recently implemented to implement MRP-227 17 Revision 1-A, which is the latest NRC approved 18 revision of the guideline.
19 The GALL SLR allows the use of the 20 existing 60 year RVI AMP if supplemented by a 60 to 80 21 year gap analysis with MRP-227A as the starting point.
22 When Point Beach was revised, the SLRA was revised to 23 use the NRC interim staff guidance. The ISG that's 24 listed on this slide, on the third bullet, and the 25 fifth bullet. And that allows us to use MRP-227 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
152 1 Revision 1-A as the starting point, as opposed to MRP-2 227-A.
3 Also showing on the slide, you'll see the 4 two industry guideline documents that we use to 5 complete our analysis, and the gap analysis. And 6 these two documents, MRP-191, and MRP-2018-022, there 7 pertain to screening, categorization, and ranking of 8 the reactor vessel internals components for a period 9 of 80 years of operation. In the latter part of the 10 slide, you'll also see that we are going to continue 11 to use industry operating experience to inform the 12 program.
13 And NextEra Energy will continue to 14 actively participate in the joint industry issue 15 programs, and update the RVI aging management program 16 as needed. So, I'll pause here to see if there are 17 any questions. With that, I would like to turn it 18 over back to Steve, thank you.
19 MR. FRANZONE: We had to move the 20 microphone over. The design configuration of the 21 reactor cavity area at Point Beach consists of a three 22 foot two inch thick biological shield wall, which I'll 23 refer to as the BSW, and it surrounds the active fuel 24 region. This BSW is integral with a six foot six inch 25 primary shield wall, which I'll refer to as the PSW.
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153 1 The BSW provides radiation shielding for 2 the PSW, but it does not perform a license renewal 3 structural intended function. All of the concrete in 4 this area for Point Beach is 4000 PSI concrete 5 strength. Go to slide 26. This is a diagram of the 6 reactor cavity area at Point Beach. What I've 7 highlighted in beige here is the BSW, or biological 8 shield wall.
9 And in blue, is the primary shield wall, 10 or the PSW. The reactor is shown here in phantom.
11 Above the beige area, along the blue wall is where the 12 horizontal supports attach to the PSW, and at 13 elevation ten foot is where the column supports are 14 anchored into the concrete. Next slide, slide 27 15 Steve. Based on the calculations, which were 16 performed by Westinghouse for 72 effective full power 17 years.
18 Which equates to the 80 year life of the 19 plant for SLR, the neutron fluence, and gamma dose 20 exposures on the biological shield wall at the reactor 21 vessel belt line were both above the NURER-2191, and 22 2192 thresholds. As a result of that, you are 23 essentially required to go back, and take a look at 24 what the impacts are on the concrete strain.
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154 1 has been using provided by EPRI, the effects of those 2 exposures on the biological shield wall are as 3 follows. And these assumptions were made as part of 4 our structural evaluation. It's assumed that there's 5 zero remaining strength for the first 3.9 inches 6 horizontally into the biological shield wall, and we 7 apply that to the entire active fuel length.
8 Even though you do get some drop off, and 9 it's more like a curve, which drops off at the upper, 10 and lower ends of the active fuel region. And this 11 accounts for the neutron fluence effect, including 12 radiation induced volumetric expansion more commonly 13 referred to as RIVE. For an additional 20 inches 14 horizontally into the biological shield wall, we 15 assumed there was 80 percent remaining strength into 16 the BSW for the whole active fuel length.
17 And again, that's a conservative 18 assumption, since you're applying that reduction in 19 strength both horizontally, and vertically with that 20 full loss. The evaluation confirmed that the BSW will 21 maintain its structural integrity, and not impact the 22 PSW under design basis loading conditions. I would 23 like to point out that as I identified where the 24 supports were actually attached to the concrete, those 25 areas are both left in the NUREG-2091, and NUREG-2192 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
155 1 thresholds.
2 Those are conclusions based on these 3 evaluations that we performed, that there were no 4 further actions beyond what is currently done as part 5 of the structure's monitoring program. And with that, 6 I'll open it up for any questions you have for the 7 concrete evaluations.
8 MEMBER HALNON: So, what does the failure 9 look like, is it when it loses structural integrity, 10 does it just crumble, or fall, what is that?
11 MR. FRANZONE: One design feature I'd like 12 to point out is all of the surfaces of both the BSW, 13 and PSW have a quarter inch liner plate that's 14 anchored into the concrete. So, what we do is we 15 actually look at the strain within the concrete, and 16 when -- this is using the conservative methodology, so 17 the concrete, although it's assumed to have zero 18 strength, the fact that we have a liner plate there 19 would contain any crumbling, or cracking you may have 20 underneath it.
21 MEMBER HALNON: Okay, so back to the 22 question, what does failure look like? Is it even a 23 consideration that we have to worry about?
24 MR. FRANZONE: No, because if you'll see, 25 the buttresses, and the steel reinforcing that goes 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
156 1 into the BSW, you will not get a failure. You'll have 2 a reduction in strength of the evaluation we conducted 3 confirm that you would not have a failure.
4 MEMBER HALNON: Okay, you got it, thank 5 you.
6 MR. FRANZONE: Yeah.
7 MEMBER SUNSERI: So, before you go on, I'd 8 just like to remind everybody listening in, if you're 9 not actively speaking, there's some tapping background 10 noise coming from some place, so appreciate that, 11 thanks.
12 MR. FRANZONE: Okay, so I believe now it's 13 going back to Anees Udyawar, and he will talk about 14 the radiation evaluation done by Westinghouse of the 15 fuel reactor vessel for us.
16 MR. UDYAWAR: Thank you Steve, and 17 probably that tapping was when I turned on the mic, so 18 my apologies. My name is Anees Udyawar, I'm from 19 Westinghouse, and I work on various fraction 20 mechanics, and ASME Section 11 small evaluation 21 projects. The next set of slides will be on the Point 22 Beach steel support assessments. And the next two 23 slides will provide the basic configuration, and 24 geometry of the supports.
25 But let me first give a brief background 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
157 1 on the assessment. As part of the SLR program, the 2 Point Beach structural steels were analyzed based on 3 a fraction mechanics evaluation to determine whether 4 the complements are structurally safe with 5 consideration of decreased fracture toughness duly 6 written for 80 years.
7 The objective was to demonstrate that 8 there is a sufficient level of flow tolerance to 9 justify continued, or current tenure ASME Section 11 10 in service examinations into the future. The analysis 11 is consistent with the methodology previously accepted 12 by the industry, and the NRC in NUREG-0933, 1509, and 13 NUREG-5220, which had all previously determined that 14 the supports were in general, structurally safe for 40 15 years, and for 60 years of radiation.
16 So, on this slide, we see the Point Beach 17 steel structure, which has a six sided structural 18 steel ring, girder, supported at the top by a 19 foot 19 long steel column, which extends down into the 20 interior concrete structure that is below the RPV.
21 The columns are about 12 inches diameter steel 22 cylinders, and those columns support the box we're 23 anchoring.
24 Let's go to slide 29, thank you. The 25 center of each of the segments of the ring girders 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
158 1 provides lateral, and rotational restraints by the I-2 beam structural members, which are embedded in the 3 surrounding primary shield wall concrete, and that's 4 another figure on the left here. The six columns of 5 the support structure are bolted at the top of the 6 anchor, and pinned at the bottom to the floor anchors, 7 as you can see in the figure on the right there.
8 The RPV has six support pads, one of each 9 of the four nozzles has a support pad, and then there 10 are two additional gusset-braced support pads, which 11 are welded directly to the vessel. These vessel 12 supports bears on the support shoe, which is fastened 13 to the support structure. The support shoe is 14 designed to restrain the vertical, lateral, and 15 rotational movements of the RPV, but allow for thermal 16 grill by permitting radial sliding on the barrier 17 plates at each of the supports.
18 Let's go to slide 30. Now, this slide 19 shows the column supports, these are the vertical 20 columns that go from the ring girder to the bottom of 21 the RPV. This particular support is surrounded by 22 concrete, and this model is coming from the neutron 23 transport analysis, which was used to generate the 24 iron DPA. The industrial guidance requires us to use 25 iron DPA neutron energies above .1 MMEs.
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159 1 But the evaluation period considered iron 2 DPAs both above, and below .1 MME for a very 3 comprehensive analysis. Shown here is the maximum 4 DPA, 5.4 times 10 to the -4 in the column. But we 5 also calculated other iron DPAs along the entire 6 length of the column, and the ring girder, and to the 7 bottom of the columns to be a very comprehensive 8 analysis.
9 Both the Point Beach units were similarly 10 evaluated to determine the DPAs. Next slide, slide 11 31. Some of the columns are not surrounded by 12 concrete, so the DPAs were calculated as well for 13 these column points since the DPA were slightly higher 14 than the columns which were encased in concrete.
15 Based on the iron DPA, the shift, and build up to the 16 temperature was calculated based on regulations 17 provided in NUREG-1509.
18 And these were previously also used in the 19 familiar investigation. Next slide, slide 32.
20 Lastly, following the guidance of NUREG-1509, the 21 critical source size for each of the ten RPV support 22 complements were determined by setting the applied 23 stress intensity factor equal to the fracture 24 toughness of the material. All the loading conditions 25 were considered, such as normal, upset, emergency, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
160 1 faulted.
2 The critical flaw sizes were then compared 3 with the ASME Section 11 allowable flaw sizes, which 4 is the recommendation for NUREG-1509. In many cases 5 the critical flaw sizes were larger than section 11 6 allowables. For other cases, we could compare the 7 critical flaw sizes to the initial fabrication 8 requirements of the design specification.
9 Since there is no significant transient, 10 or thermal cycling at the supports, there will be no 11 significant crack growth over time. Thus the initial 12 fabrication flaw sizes on the order of the critical 13 flaw sizes would have been repaired, or the concrete, 14 and lead would be replaced. Thus the Point Beach 15 supports were evaluated to be structurally safe, and 16 stable, and the flawed tolerance constraint 80 years 17 of radiation embrittlement was considered.
18 It should be noted that per NUREG-1509, 19 physical examination of the complement is also 20 essential to assess the overall condition of the 21 support structure. Thus, in accordance with ASME 22 Section 11 IWF requirements, the Point Beach Unit 1 23 supports were examined 2005, 2007, 2016. And the Unit 24 2 supports were inspected in 2006, 2008, and 2015.
25 And based on all these latest examinations, there are 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
161 1 no relevant conditions identified that were applicable 2 for the fraction mechanics assessment.
3 So, in conclusion no further action is 4 recommended beyond doing the current ASME Section 11 5 IWF inspection requirements. This concludes the 6 discussion on the structural steel supports, and if 7 there's any questions, I can field them. Thank you, 8 and I can pass this back to Steve Franzone for 9 discussions on epoxy residue grout. Thank you.
10 MR. FRANZONE: Yes, thank you Anees. All 11 right, so the next couple of slides, we want to cover 12 several questions from --
13 MEMBER BROWN: This is Charlie Brown, can 14 you go backwards for a minute? I'm not a metaller, so 15 I'll ask a question. You say in most cases the 16 critical flaw sizes are larger than the allowable flaw 17 sizes by a large margin. And then you said but 18 they're okay out to 80 years, going from 60, to 80 19 years. Is there a basis for that? Like the stresses 20 are so low that they're not going to increase?
21 Or the temperature variations are so 22 minor, that you don't get any structural stressing 23 from that standpoint?
24 MR. UDYAWAR: Yeah that's -- sorry to 25 interrupt. Yeah, that's a good question. Certain 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
162 1 components, the stresses were very low. Some of them 2 were less than six KSI, or even lower, or some of them 3 were compressive. So, in those cases, we were not 4 getting any crack growth on any kind of critical flaw 5 sizes.
6 In other parts of the complement, if 7 they're away from the belt line region, the fluence 8 embrittlement was also very, very slow. So, also for 9 those considerations, the critical flaw sizes were 10 much lighter. So, you're correct, it's either small 11 stresses, or minimum level of iron DPA for certain 12 locations.
13 MEMBER BROWN: What's a DPA?
14 MR. UDYAWAR: Displacement per atoms, it's 15 --
16 MEMBER BROWN: Okay, thank you, I had 17 never heard that one before. Thank you, appreciate 18 it.
19 MR. UDYAWAR: Thanks.
20 MR. FRANZONE: All right, are we ready to 21 go on? All right. Again, so the next couple slides 22 we hope to cover several questions for epoxy resin 23 based grout. What is it? How, and where is it used 24 at Point Beach? Why do we care? And what did we do?
25 So, let's dive right into it. Epoxy resin based grout 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
163 1 consists of a resin, a hardener, and an aggregate. It 2 is designed for a high dynamic load capacity, and 3 improved load transfer, which improves stability, and 4 alignment, and minimizes vibration.
5 Here at Point Beach, the use of epoxy 6 resin based grout is controlled by the engineering 7 change process. Our engineering change process 8 controls how various engineering products, and 9 activities are developed, and control changes to the 10 plant's configuration. For example, it provides 11 engineering with a vehicle to communicate to the 12 craft, the necessary details to control for instance 13 the surface preparation, and mixing procedures.
14 And it includes hold points for 15 engineering a quality control verification. Full 16 scale mock ups were used to validate the installation 17 procedures, and demonstrate the use of epoxy resin 18 based grout in these specific applications. The 19 engineering change also provides the proper vehicle to 20 evaluate the change from concrete, and mortar to the 21 epoxy resin based grout.
22 The use of epoxy resin based grout is not 23 permitted in areas where normal temperatures exceed 24 120 degrees Fahrenheit, and is not used in areas of 25 high radiation exposure. The process was used for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
164 1 both the service water -- or this process was used for 2 both the service water, and cooling water pumps. And 3 just by the way, the last component cooling water pump 4 is being replaced this outage.
5 If you'll turn to slide 34. Almost 40 6 years ago Information Notice 83-40 was sent out by the 7 NRC staff to their licensees to potential issues with 8 the use of epoxy grout. Two issues were discussed, 9 and I quote, the compounds had shown significant loss 10 of strength at temperatures above 120 degrees 11 Fahrenheit, and relatively low creep strength of 12 epoxy.
13 The science in this industry has come a 14 long way since those days, and epoxy resin based 15 grouts have proven to be very effective in certain 16 uses. Epoxy resin based grout used as an aggregate, 17 which significantly reduces the effective creep. It 18 also acts as heat sink, and stabilizer. It also 19 increases ultimate strength, and reduces shrinkage, 20 and cracking.
21 One would have to ask is it better than 22 sliced bread? Current aging management practices for 23 the structured monitoring program, a visual inspection 24 addresses the aging effect discussed on the slide with 25 the exception of creep. To address this aging effect, 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
165 1 we added requirements to conduct checks for tightness.
2 That it initial, and periodic torque checks of all the 3 anchors within the scope of license renewal embedded 4 with epoxy resin based grout.
5 The issues associated with pump alignment, 6 and vibrations have virtually disappeared. And with 7 that, does anybody have any questions on this issue?
8 Okay, thanks for the opportunity. I will now turn the 9 presentation back to Bill Maher for closing remarks.
10 MR. MAHER: Thanks Steve. So, even though 11 Point Beach was a GALL Rev Zero plant in the first 12 round of license renewal, we have adopted the SLR Gall 13 with minimal exceptions. Like Mike stated in his 14 remarks earlier, in keeping with the sustainability 15 that he spoke about, the goal now is to focus on 16 building, and maintaining the margin that we currently 17 have in order to get to 80 years of operation.
18 We have, as part of that, to oversee that 19 maintaining of margins, we have a dedicated individual 20 on site to oversee the implementation of these aging 21 management programs as we transition to the subsequent 22 period of extended operation. Any questions?
23 MEMBER SUNSERI: Members, any questions 24 for the applicant, NextEra? What about the 25 transition? Okay, so thank you very much for your 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
166 1 presentations, good job.
2 MR. MAHER: And thank you to the staff.
3 MEMBER SUNSERI: All right, now we will 4 transition back to the DNRL staff. Brian, your team, 5 are you ready?
6 MR. ROGERS: Yeah, good afternoon, we're 7 just switching slides.
8 CHAIRMAN REMPE: So, there appears to be 9 two microphones open, and we're getting a bit of an 10 echo. So, whoever is speaking should have their mic 11 open, and others please mute themselves.
12 MEMBER SUNSERI: So, Bill it's on your 13 end, if you're talking through your phone you've got 14 to mute your phone, and the computer.
15 MR. ROGERS: Okay, can you hear me now?
16 MEMBER SUNSERI: Perfect, thank you.
17 MR. ROGERS: All right, excellent, thank 18 you very much. Okay, good afternoon Chairman Rempe, 19 and members. Can you see my slides?
20 MEMBER SUNSERI: Yes.
21 CHAIRMAN REMPE: Yes we can.
22 MR. ROGERS: Okay, thank you. Good 23 afternoon Chairman Rempe, and Sunseri, and members of 24 the ACRS. My name is Bill Rogers, and I'm one of the 25 senior license renewal project managers in the Office 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
167 1 Nuclear Reactor Regulation, and we are here today to 2 discuss the staff's safety review of the Point Beach 3 Nuclear Plant subsequent license renewal application, 4 or SLRA as documented in the safety evaluation of NRC.
5 And joining me at my table today, I have 6 Angie Buford, chief of the Vessels and Internals 7 Branch, who is assisting me with the slides, and 8 presentation. We have additional technical branch 9 chiefs, and staff also on the line, and Dr. Hiser, 10 previously the license renewal SL, and now we've hired 11 a new one since in DNRL that's with me. And also 12 Lauren Gibson, chief of the Projects Branch.
13 Next slide, please. So, we'll begin 14 today's presentation with an overview of the Point 15 Beach licensing history before moving onto the Point 16 Beach aging management programs. We've already 17 discussed selective technical errors that we believe 18 are of interest to ACRS, and hear from the Region III 19 inspections, and plant material conditions before 20 assuring the conclusion of the staff's safety review.
21 Next slide, please. The Point Beach SLRA 22 described -- excuse me, one moment. The Point Beach 23 Units 1 and 2 were initially licensed on October 5th, 24 1970, and March 8th, 1973, respectively. In February 25 of 2004, the applicant submitted the initial license 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
168 1 renewal application. The initial renewed licenses 2 were issued December 2005, extending the expiration 3 dates to October 2030 and to March 2030 for Units 1 4 and 2, respectively.
5 On October 16th, 2020, NextEra submitted 6 an SLRA for Point Beach Units 1 and 2. The 7 application was accepted for review on January 8th, 8 2021, and the safety evaluation was issued on February 9 23rd, 2022 with no open or confirmatory items.
10 Next slide, please. The Point Beach SLRA 11 described a total of 48 AMPs, 39 existing programs, 12 and nine new programs.
13 This slide identifies the applicant's 14 original disposition, and these AMPS as updated in the 15 SLRA in the left column, and the final position is 16 documented in the SE in the right. All of the AMPs 17 were evaluated for consistency with the generic aging 18 lessons learned, subsequent license renewal, our SLR 19 report, and ultimately all of the AMPs were found to 20 be consistent with the broad SLR, including those with 21 enhancements, or exceptions.
22 The applicant did not introduce any plant 23 specific AMPs. So, I'd like to add a bit about the 24 work we did perform to do the review of the aging 25 management programs, and the other technical 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
169 1 information in the application. We had extensive 2 audit activities, including a review of documents that 3 were placed in a portal -- a real voluminous 4 collection of documents.
5 We had three weeks of breakout sessions 6 that addressed most all of the aging management 7 programs. We did an extensive OE review, also through 8 the portal. And in addition, the applicant was unable 9 to provide us with some current visual representation 10 of items of interest on the plant and in the plant.
11 And the reason we chose that approach was due to the 12 COVID pandemic that affected our ability to travel 13 during the review.
14 We'd like to thank the applicant for 15 providing this, an alternative method for us to see 16 things of interest. We had 14 RAI sets, 62 RAIs.
17 I'll note that we only had four second round RAIs from 18 this particular review. The applicant made five SLRA 19 supplements with an annual update. We had six public 20 meetings. The public meeting topics included the 21 irradiated reactor vessel supports.
22 We had two for reactor vessel internals.
23 We had a third on reactor vessel internals. We had a 24 fourth meeting on epoxy grouted anchors, fatigue 25 amounts and cycles, selective leeching, and cathodic 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
170 1 protection. We had a second meeting, the fifth 2 meeting also included cathodic protection, and a sixth 3 meeting on the fire protection response.
4 So, based on our review, the results of 5 the audits, and additional information provided by the 6 applicant, the staff concluded that the applicant's 7 aging management program activities end results were 8 consistent with the criteria of the standard review 9 plan, the requirements of 10 CFR Part 54. Any 10 questions on the information I presented so far?
11 MEMBER SUNSERI: No questions, please go 12 ahead.
13 MR. ROGERS: Okay, next slide, please.
14 So, this slide represents certain to specific areas of 15 the SLRA review, these are technical issues that we 16 spent some time on. The first bullet refers to 17 staff's evaluation of epoxy resin grout for the 18 embedded fasteners, which had been used to replace the 19 more common cementitious grouted anchors.
20 During review of the application, and the 21 NRC staff audit reviews, the applicant indicated that 22 epoxy resin grout was used for the embedded anchors in 23 the service water pump and the component cooling water 24 pump base plates. The staff questioned the 25 qualification of epoxy grouted fasteners for these 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
171 1 applications to better understand the potential aging 2 mechanisms.
3 The applicant provided additional 4 information that indicated that epoxy resin based 5 grout materials have been purchased as safety related 6 from the supplier, and they qualify for use in safety 7 related applications. This data is consistent with 8 the requirements of 10 CFR Part 50 Appendix B. In 9 addition the plant application was subject to an 10 engineering change process review for use in these 11 particular applications like the service water pump, 12 and CCW pump base plate.
13 As previously mentioned, it was noted in 14 the NRC Information Notice 83-40 identified the need 15 to consider the creep strength of epoxy, and 16 recommended that were anchor bolts are embedded into 17 the epoxy grout, and tensioned to any appreciable 18 preload, it might be important to periodically verify 19 that the preload has not been lost by creep in the 20 grout.
21 The applicant enhanced the aging 22 management program, establishing the structure 23 monitoring program to address the potential aging 24 effects of creep associated with epoxy grout anchors, 25 and to do so by periodically checking the anchor 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
172 1 bolts, and different epoxy resin based grout, verify 2 that preload had not been lost due to creep.
3 The remaining four bullets are those areas 4 referred to in the staff requirements SECY-14-0016, 5 ongoing staff activities to assess regulatory 6 considerations for a power reactor with subsequent 7 license renewal. Those four items are reactor 8 pressure vessel neutron embrittlement, reactor vessel 9 internals, radiation assisted stress corrosion 10 cracking, irradiated concrete containment, and 11 electrical cable qualification, and condition 12 assessment.
13 For each of these four areas, the 14 applicant provided quality information as contained in 15 the SLRA as provided during the core of the staff's 16 review. The staff was able to disposition the 17 applicant's information using the GALL SLR report, the 18 SRP SLR, and the memo on staff guidance, which allowed 19 the staff to obtain reasonable assurance, and made its 20 conclusion through the normal review process with the 21 expected level of engagement.
22 So, I know we've discussed this a bit 23 previously, but if there are any questions, or further 24 information needed in this area, we have the technical 25 staff currently available, and prepared to respond.
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173 1 MEMBER SUNSERI: Members, any questions?
2 Doesn't sound like any, go ahead.
3 MR. ROGERS: Okay, thank you. Well, at 4 this time we're going to turn it over to John Bozga, 5 Region III senior inspector. And he's going to 6 present some slides on these colonies. But to do so, 7 we're going to turn the presentation over to this 8 camera. So, we'll need to pause a moment while John 9 takes possession of the presentation. John, are you 10 ready?
11 MR. BOZGA: Yes, I am. All right, Mike, 12 thank you.
13 MR. ZIOLKOWSKI: Yeah, this is Mike 14 Zilokowski, Region III, let me know when you can see 15 the PowerPoint. I will present while John speaks.
16 MEMBER SUNSERI: All right, we have it up, 17 you might want to put it in presentation mode.
18 MR. ZIOLKOWSKI: Okay, thank you.
19 MEMBER SUNSERI: Very good, there it is.
20 MR. BOZGA: All right, thank you. Good 21 afternoon everyone, I'm John Bozga, senior reactor 22 inspector in the Region Three Division of Reactor 23 Safety Engineering Branch One. One of my inspection 24 duties is to perform license renewal inspections.
25 Also joining me on behalf of the region is my branch 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
174 1 chief Mike Zilokowski, as well as the senior resident 2 inspector for Point Beach, Tom Hartman.
3 I'm here to provide Region Three's review, 4 and assessment of the implementation of aging 5 management programs, material condition, and overall 6 regulatory assessment of Point Beach Units one, and 7 two. The license renewal inspection program, and the 8 reactor oversight baseline inspection program are both 9 used to inspect aging management activities at Point 10 Beach.
11 I'll start with the activities performed 12 under the license renewal inspection program, then 13 discuss the LOP inspections, and follow up with 14 material condition of the plant. So, in order to 15 assess the adequacy of the license renewal program of 16 the initial period of extended operation, inspection 17 procedure 71003 recommends using a four phased 18 approach to license renewal inspection. This current 19 slide you see lists the specific license renewal 20 inspections that have been performed at Point Beach.
21 The phase one, and phase two inspections 22 were performed for both units on the dates listed with 23 no findings identified. Phase three inspections were 24 not required, because no findings were identified in 25 phase one, or two. Finally, the phase four NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
175 1 inspection, which typically occurs five to ten years 2 into the period of extended operation has been 3 performed in September of 2019 for both Units 1 and 2 4 with no findings identified.
5 Next slide, please. In addition to the 6 inspections mandated by the license renewal inspection 7 program, inspectors used several ROP baseline 8 inspection procedures to evaluate the implementation 9 of aging management activities. One example is the 10 baseline inspection of the in service inspection 11 program. This inspection is performed each refueling 12 outage, and provides the inspectors the opportunity to 13 review, and assess inspections credited for aging 14 management.
15 The second example is the heat sink 16 inspection, which provides the inspectors an 17 opportunity to review the service water system, as 18 well as the ultimate heat sink. All of these 19 activities are within the scope of license renewal, 20 additionally the design basis assurance inspection 21 includes a review of aging management activities for 22 safety related structure systems, and components that 23 are selected during this inspection.
24 At Point Beach, the regional inspectors 25 have found no violations, or findings of greater than 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
176 1 green significance as a result of these inspections 2 performed using these aforementioned procedures. Also 3 of note, the triennial fire protection procedure has 4 been recently updated to review aging management of 5 the fire protection equipment.
6 Next slide. Currently, Point Beach Unit 7 1 and 2 are in the licensee response column, and have 8 all green findings in performance indicators. This 9 indicates that the licensee has been able to 10 effectively identify conditions adverse to quality, 11 and correct them in a timely manner. We have reviewed 12 all inspection findings over the last ten years to 13 gain insights related to aging components. We did not 14 identify any findings with an aging management aspect 15 related specifically to license renewal. Next slide, 16 please.
17 I will now speak to the material condition 18 of Point Beach from a regional inspector viewpoint.
19 We have no concerns with the overall material 20 condition of the plant that needs to be addressed 21 outside of the baseline reactor oversight process.
22 The licensee has been successful at completing large 23 capital improvement projects that maintain, or improve 24 the material condition of its structure, systems, and 25 components.
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177 1 The license renewal program inspections 2 did not identify any substantial weaknesses in the 3 station's performance in managing the effects of aging 4 at the site. The inspectors will continue to inspect, 5 and assess the licensee's ability to manage the 6 effects of aging through the baseline inspections.
7 Lastly, are there any questions?
8 MEMBER HALNON: This is Greg. Do you have 9 any feel for where they are on the cross cutting 10 aspects? I understand that there's no better than 11 green findings, which doesn't really give you a full 12 story. Were you seeing any things, or anything 13 producing out of that program?
14 MR. BOZGA: I'm not aware of anything, 15 I'll let Tom, and, or Mike speak to that as well.
16 MR. HARTMAN: Hi, I'm Tom Hartman, I'm the 17 senior resident here at Point Beach. I've been here 18 for just about five years now, and we do not have any 19 themes, or patterns, even though they have not 20 triggered any of the cross cutting thresholds, but we 21 do not have any transient themes for any of the areas 22 at this time.
23 MEMBER HALNON: Okay. Because what I was 24 kind of poking at was back to my earlier conversation 25 about the large amount of enhancements. And I 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
178 1 understand that it changes standards over the years, 2 and it's more rigorous, but there was a lot of 3 procedural issues in there. And I was just curious if 4 there were any procedural themes showing up in the 5 cross cutting aspect.
6 So, it sounds like these are truly 7 enhancements that aren't affecting the material 8 condition of the plant at this point.
9 MEMBER BIER: I have another quick 10 question, this is Vicki Bier, and this may be just 11 over interpreting the wording, but it was said that 12 there were no aging management issues specifically 13 related to license renewal. Were there any aging 14 management issues not related to licensing renewal, or 15 am I being overly persnickety?
16 MR. BOZGA: I'm not exactly sure what 17 we're getting at. We're looking for aging management 18 issues, are we talking about service life issues, or?
19 MEMBER BIER: I'm just wondering why that 20 choice of wording, and maybe somebody just threw in 21 some extra words that were not needed.
22 MR. BOZGA: No, the wording I would not 23 say is thrown in there, but sometimes there is 24 confusion, either on the NRC, or even on the industry 25 side with the difference between service life, and an 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
179 1 aging effect. Say a vendor recommends to replace a 2 component every ten years, and the stations decide to 3 go further than that, and the component fails due to 4 some aging degradation mechanism.
5 I would consider that different than a 6 license renewal aging management issue, which may not 7 have to do anything with service life.
8 MEMBER BIER: Okay, thanks for the 9 explanation.
10 MR. BOZGA: You're welcome.
11 MEMBER SUNSERI: so, this is Matt, and 12 I'll ask the question this way. I really do 13 appreciate, and value what the resident inspectors 14 bring to this process. And so from your individual 15 observations at Point Beach, are there things below 16 what I'll call the regulatory threshold that we might 17 be interested in? Because those kind of things can 18 sometimes be a precursor to challenging a regulatory 19 threshold.
20 So, let me give you an example, does the 21 plant have a large amount of leaks, whether they're 22 steam, oil, water, gas? Do they use extensive amount 23 of temporary leak repairs, or are their maintenance 24 practices holding the plant together well? Any kind 25 of insights like that that you can share with us, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
180 1 we recognize that your obligation is to the regulatory 2 threshold.
3 But we don't get to see these plants, and 4 so anything you could kind of offer us would be 5 insightful.
6 MR. HARTMAN: Sure. This is the fourth 7 plant that I've been an inspector at in my tenure with 8 the agency. I actually -- for being as old as it is, 9 in terms of being licensed back in the early 70s, the 10 material condition is actually quite good. The 11 station does a good job of maintaining the situation.
12 There's not a lot of leaks, and material condition 13 issues.
14 There have been in the past, issues with 15 underground cables that the station was very good to 16 address, and we've documented those in our reports.
17 But since I've been here in the last five years, 18 there's not been any significant concerns that I have 19 about any -- even the low level things like 20 continually leaking -- leak management programs, or 21 stuff like that, having to repair, or do any type of 22 ASME repairs to fuel wall leaks, or any of that kind 23 of stuff.
24 I've not seen -- as with every plant, they 25 occur, but nothing that I would call astounding, 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
181 1 especially for a plant that's already 50 years in.
2 MEMBER SUNSERI: All right good. And then 3 from a cultural standpoint, kind of following up with 4 Greg's question about cross cutting issues, how is the 5 staff as far as questioning attitude reporting 6 problems, do you see anything there?
7 MR. HARTMAN: Again, it meets our 8 standards. I have not seen any issues with -- I do 9 talk to the individuals out in the station, do 10 pulsings of people. I've not seen any issue with 11 safety conscious work environment, or safety culture, 12 they're all willing to bring up safety issues. I'm 13 hoping that's in line with your question. I don't 14 have any concern about the cross cutters.
15 Again, every station has challenges, 16 communications, procedure use adherence, there's 17 little things here, and there, but nothing that I 18 would call alarming, or trend worthy.
19 MEMBER SUNSERI: That's good, I did notice 20 in one of the inspection reports there was a 21 commitment that had been missed, but they were 22 identified by the utility that they missed it, and 23 they presented to you that they missed it, the 24 corrective action plan when it was going to be 25 corrected, and when the inspection was going to be 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
182 1 done. So, I mean I thought that was a healthy -- like 2 you said, plants are complex.
3 They have a lot of requirements, and 4 people make mistakes every once in a while. It's how 5 you deal with those mistakes, and how you react to 6 them. So, it seemed to me from that one data point, 7 that they're pretty responsive. But I just wanted to 8 kind of test you in a more generic sense, you had a 9 larger sample than we did. That was good, you 10 answered my question.
11 Members, anything else? Okay then, I 12 guess are you about concluded with your presentation 13 for the staff side?
14 MR. BOZGA: I'm going to turn it over to 15 Bill, so thank you.
16 MEMBER SUNSERI: Thank you. I'm sorry, we 17 may have lost you if you're speaking, I don't hear 18 anything.
19 MR. SMITH: Just one moment, we're 20 reclaiming the business.
21 MEMBER SUNSERI: Got it, thanks.
22 MR. SMITH: Well, actually I can do this 23 without the slides, that's okay. Can everyone hear 24 me?
25 MEMBER SUNSERI: Yes.
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183 1 MR. SMITH: Okay. What I wanted to say 2 for the final slide was thank you. And say in 3 conclusion for the SLRA safety review, the staff finds 4 the requirements of 10 CFR 5429A have been met, and 5 the subsequent license renewal for Point Beach Nuclear 6 Plant Units 1 and 2. And we'll now be happy to answer 7 any additional questions that you might have. Thank 8 you.
9 MEMBER SUNSERI: All right, well thank 10 you, and thank you to staff. So, members, this 11 concludes the two presentations, one from the 12 applicant, and one from the staff on their safety 13 review. Any questions you have, anything from the 14 staff?
15 MEMBER HALNON: I don't see anybody 16 nodding their head.
17 MEMBER SUNSERI: All right, good. Yeah, 18 that's the downside here, I can't see you, you guys 19 can see me, but I can't see you.
20 All right, then, I suppose at this point 21 we can turn to public comments, and see if we have any 22 comments. So, let's open up the phone line. Members 23 of the public listening in, this is the opportunity to 24 make a comment, so just unmute yourself, state your 25 name, and provide your comment.
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184 1 Let's open the lines up now. Okay, and 2 not hearing any comments, I'll give it one more 3 second, or two here. All right, and so that's it for 4 the public comments. I will provide the same courtesy 5 to the members in the room there, if there's anybody 6 in the room that at this point cares to make a 7 comment, please do so. All right, well I guess that 8 wraps up our comment period then.
9 So, members I will now ask Chairman Rempe 10 how you would like to proceed. But at this point in 11 time we are complete with the presentation. If you 12 want to entertain a brief committee discussion on any 13 topics that they would prefer to see in the draft 14 report before I take it out to the subcommittee to 15 prepare that draft report for the full committee 16 meeting in a month, I would appreciate any comments.
17 MEMBER HALNON: Matt, this is Greg. I 18 think a short explanation of the GALL zero, one, two 19 ISG question I had earlier would help make sure that 20 we understand the large number of commitments that 21 were made, and enhancements to the existing programs.
22 That really stuck out to me, that there was a 23 tremendous amount of it.
24 MEMBER SUNSERI: Yeah Greg, so I mean 25 we're kind of spoiled I think a little bit by the last 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
185 1 two of these things that we were doing, which had very 2 few enhancements, and exceptions. We have some other 3 reviews where there's similar numbers, but yeah, we 4 can definitely include that in the discussion.
5 MEMBER HALNON: Yeah, it's this timing 6 thing. I mean this is an older plant license renewal, 7 the initial license happened a long time ago, and it's 8 not a negative at all, I think it actually could be a 9 positive if the --
10 MEMBER SUNSERI: Yeah, right, got it.
11 It's normally not that the plants aren't doing the 12 things, they just aren't committed to doing them. So, 13 like having qualified people do the work, right?
14 MEMBER HALNON: Right, there's no 15 indication that unqualified people were doing it, 16 clearly. So, it kind of cements some of the more 17 stronger standards that we see today than we did in 18 the early 2000s.
19 MEMBER SUNSERI: Perfect, yeah, that's 20 good, I got it.
21 VICE CHAIRMAN KIRCHNER: Matt, this is 22 Walt, what's your take on the AMP with exception 23 regarding the capsules for the pressure vessel 24 radiation?
25 MEMBER SUNSERI: Well, so as it was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
186 1 presented, any exception to the surveillance capsule 2 schedule is an automatic exception. And I think based 3 on our previous topic this day, I imagine they are 4 revising their surveillance schedule to kind of get in 5 sync with those new PWR requests. So, we can explore 6 that a little bit more, but that was my take away on 7 that. So, it's just simply an exception because 8 they're changing their schedule to address the new 9 procedures.
10 MEMBER BALLINGER: Point Beach has one 11 extension, so it's only had one extension.
12 VICE CHAIRMAN KIRCHNER: One extension of 13 what?
14 MEMBER BALLINGER: Extension of the time 15 to take the capsule out.
16 VICE CHAIRMAN KIRCHNER: Okay, so when is 17 their capsule coming --
18 MEMBER BALLINGER: I don't know what the 19 date is --
20 VICE CHAIRMAN KIRCHNER: Because this is 21 an older plant, I think it would be an interesting 22 data point.
23 MEMBER BALLINGER: Yeah. In my head I 24 seem to remember that Point Beach, and Dave Rudland's 25 gone, but it is the plant that they were plotting the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
187 1 data points in his presentation. I should go back, 2 and --
3 CHAIRMAN REMPE: It looks like the 4 licensee has --
5 VICE CHAIRMAN KIRCHNER: Yeah, Steve Hale, 6 could you answer that?
7 MEMBER SUNSERI: Steve Hale is here, 8 you're muted Steve, we can't hear you.
9 MR. HALE: We had one capsule we were 10 going to remove at a specific timeframe to account for 11 60 years, and we essentially just delayed removal of 12 that capsule so we could get a capsule that had 80 13 years of exposure. So, we were going to remove one, 14 but in order to ensure that we captured 80 years of 15 fluence, we just delayed the time we would remove it.
16 But the way the GALL AMP is written for the reactor 17 vessel surveillance program, if you have to make a 18 change to your existing capsule removal schedule, that 19 is an automatic exception.
20 But as part of the application, we 21 submitted all the requirements that you need to follow 22 to change your surveillance capsule removal. We have 23 the specific timeframe, I believe.
24 MR. FRANZONE: Yeah, it's 51 effective 25 full power years, which right now is estimated to be 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
188 1 in the spring of 2035.
2 MR. HALE: That was strictly just a delay 3 in the removal of the final capsule to capture 80 4 years of fluence.
5 MR. FRANZONE: And that's for Unit 2, 6 that's the capsule in Unit 2.
7 MEMBER BALLINGER: If I recall, based on 8 the EPRI presentation this morning, 2035, by that time 9 they will have a lot more data from other -- from 10 their program if I recall.
11 MR. FRANZONE: Correct, it's a living 12 program, so as we move forward, we'll have a lot more 13 knowledge, so things may change again.
14 MEMBER SUNSERI: Right, no doubt. So, 15 yeah, thanks for that explanation. Walt, does that 16 address your point?
17 VICE CHAIRMAN KIRCHNER: Yeah, thank you.
18 MR. MAHER: This is Bill Maher with 19 NextEra Point Beach, we did want to address some of 20 the exception questions --
21 MR. FRANZONE: Yeah, the enhancement 22 questions.
23 MR. MAHER: The enhancement portion, so 24 Steve has some discussion on that.
25 MR. FRANZONE: Yeah. For our purposes, we 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
189 1 don't write the actual implementation procedures until 2 after we obtain approval -- or we actually don't 3 implement them, our implementation procedures are not 4 approved until actually we get the license. And so 5 although we have the enhancements, I think other 6 utilities kind of do it, they actually write their 7 procedures, get the enhancements in place, so they 8 actually have something to show the reviewers.
9 We don't do it that way, we list the 10 enhancements in our basis documents, which support the 11 application. And I think that may be one reason for 12 say the other sites, like North Anna, or Surry where 13 they actually have gone through, and actually have 14 written all those, and have them place, where we don't 15 do that. So, it may look like we have. There may be 16 a big difference, but it may not really be a 17 difference, it may be just a tiny difference in that.
18 So, I just wanted to throw that out there.
19 But we do like to give clear directions, and I think 20 we're hard, we get to pass this on, people have to 21 implement this maybe ten years from now. And so we 22 want to make sure it's very clear to the future out 23 there, the people who are actually implementing the 24 feature. And so I think for us, for our benefit, 25 we're very clear, and we're kind of -- I'll say maybe 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
190 1 overly conservative in our directions to make sure we 2 capture it for future use.
3 MR. HALE: Yeah, one thing I wanted to add 4 Steve also, is our threshold for identifying 5 enhancements is a lot lower than what it was for 6 original license renewal. So, we've taken all the way 7 down to specific procedural enhancements in some 8 cases. And having it there as a commitment, it's out 9 in front of everybody, and it's very difficult to step 10 away from that. You have to follow a formal 11 regulatory process to change that commitment.
12 MEMBER SUNSERI: Yeah, I think as it was 13 mentioned here, we're not saying that we see this as 14 an issue, or anything. But because people read our 15 reports, and pay attention to them, and we usually 16 replicate this table, or the statistics in our 17 reports, it would be -- I think prudent for us to at 18 least have a statement about why this is occurring 19 here.
20 So, that if anybody looks, doesn't just 21 take our report on face value, and try to make 22 something negative because of a previous report.
23 That's all I think we're trying to do.
24 MEMBER HALNON: Yeah, let me just ask one 25 question to clarify. So, in the fire protection AMP, 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
191 1 there is a commitment to make sure that semiannual dry 2 chemical inspections -- dry chemical and fire 3 extinguisher inspections are completed. So, from a 4 face value just looking at this, you could make the 5 conclusion that right now you don't do semiannual dry 6 chemica, and fire extinguisher inspections.
7 I have a hard time believing that's the 8 case, because that's required by the -- that was what 9 got my interest up. So, I'm assuming, and you can 10 correct me if I'm wrong, that if you see a deficiency 11 based on this review, you would fix the deficiency, 12 not wait ten years to do it, is that correct?
13 MR. FRANZONE: So, probably the rev zero 14 program was probably different than what's in the GALL 15 SLR. I think that's where you see the difference.
16 The level of detail now that you see with GALL SLR is 17 much more, we found that at Turkey Point and Point 18 Beach, that the level of detail required by GALL SLR 19 is much higher. And so you'll see in the fire 20 protection program, and I think even for St. Lucie, 21 we've included a lot more detail in there to capture 22 all those requirements, and enhancements.
23 Whereas before the program wasn't 24 required, we may be doing those under the fire 25 protection program, but not under the license renewal 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
192 1 fire protection program.
2 MR. HALE: And I'd like to also mention 3 that the GALL fire protection guidance actually goes 4 beyond what's required currently under NFPA. We have 5 requirements in GALL that requires you to do things 6 beyond say what you're currently doing under the fire 7 protection program. And with regards to ongoing 8 inspections of dry paint systems, and things of that 9 sort that are currently not part of the NFPA, for 10 example 25 program that you're doing.
11 So, when you line yourself up, you take 12 your existing program, and you line yourself up to 13 GALL, you'll identify those areas, and in some cases 14 we say hey, we need to strengthen that requirement, 15 because we don't think it's as tight as it needs to 16 be. And that's why you see those in those 17 enhancements.
18 MEMBER HALNON: So, you wait ten years to 19 tighten it up?
20 MR. HALE: No, that is only the 21 requirements for implementation. As we're going 22 through at Turkey Point right now, once we receive the 23 license, we start looking at the commitments under the 24 original license renewal, as well as subsequent 25 license renewal. In some cases it makes a lot of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
193 1 sense to pull that up, and do it early.
2 MEMBER HALNON: Is that going to be -- I 3 was going to say is that part of the implementation 4 plan for license renewal, to do that, or is that --
5 MR. HALE: Yes.
6 MEMBER HALNON: So, you're not relying on 7 the good will of your program managers to say hey, 8 this sounds like a good idea.
9 MR. HALE: No, and in fact, as you see 10 with subsequent license renewal, and I'm speaking from 11 experience here, at say later sites for original 12 license renewal, there are a lot of inspections that 13 are required pre SPEO that weren't there for original 14 license renewal. And it's quite an activity, and it's 15 quite a detailed implementation plan if you look at 16 everything that needs to be accomplished, say between 17 now, and once you enter the SPEO just from a component 18 inspection standpoint.
19 MR. MAHER: This is Bill Maher again. So, 20 that is why for subsequent license renewal we have a 21 dedicated individual that is purposefully meant to 22 make sure that those AMPs, and the procedures are put 23 in place, and the commitment for pre SPEO inspections, 24 and programs are all put in place, that's why that 25 person is dedicated there.
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
194 1 MEMBER HALNON: Okay, and again I'm going 2 to put words in your mouth. When you went through the 3 initial analysis of the AMPs, if you found a 4 deficiency, I assume that it was fixed. I mean, 5 because anytime you put these programs under a 6 microscope like you do when you are developing the 7 license application, you're going to find stuff.
8 MR. FRANZONE: Right, and so what we did, 9 we identified to the site, and we put it in the 10 plant's corrective action program. And you're right, 11 we did find some deficiencies, and we'd enter it into 12 the plant's corrective action program, and then it 13 would just go through its normal process.
14 MEMBER HALNON: Okay, that makes sense, 15 good, thank you.
16 MEMBER SUNSERI: Members, any other 17 feedback? All right then, we will conclude this 18 discussion, and I will just remind everyone that the 19 next time we take up this topic will be at our May 20 full committee meeting, which will be May 4th. This 21 is likely to be the first item on the agenda. So, if 22 you want to see the -- observe the letter writing 23 report, and be able to make any factual corrections if 24 we get something wrong during that process, you're 25 allowed to bring up factual corrections.
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
195 1 So, that will begin at 8:30 Eastern 2 Daylight Time on May 4th. Having no other business on 3 this topic, I return back to Chair Rempe.
4 CHAIRMAN REMPE: Thank you, Matt. And I'd 5 also like to express my appreciation for the staff to 6 include the folks in the region, as well as the folks 7 in headquarters. I appreciated the discussions that 8 we had with them. At this time I'd like us to go off 9 the record, and the court reporter won't be required 10 to come back until 1:00 p.m. tomorrow, when we take up 11 another VWRX300 topical report.
12 And why don't we take a 19=minute break, 13 and come back at 5:00 p.m., and we'll continue letter 14 writing on the containment evaluation method topical 15 report for VWRX300.
16 (Whereupon, the above-entitled matter went 17 off the record at 4:41 p.m.)
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
Reactor Pressure Vessel Embrittlement Monitoring and Prediction in Long-Term Operation 694th ACRS Full Committee April 6, 2022
Meeting Purpose
- Discussion of Issues
- Regulatory Guide 1.99 Rev 2 (RG 1.99) and 10 CFR 50.61 embrittlement trend curve
- Appendix H surveillance testing
- Discussion of RPV Embrittlement rulemaking plan 2
Background
Monitoring and Prediction of Embrittlement
- Embrittlement Trend Curve (ETC) provides estimates of change in fracture toughness (T or RTNDT) as a function of fluence
- Surveillance capsule testing provides monitoring to ensure ETC predicts plant specific behavior properly
- Together they are used to determine pressure-temperature (PT) limits for normal operation 40yr 60yr 80yr Heatup & Cooldown Embrittlement (T)
ART 40yr 60yr 80yr Max Pressure per ETC Data 10 CFR 50 App. G RTNDT(u)
Operating Time (years)/Fluence Coolant Temperature ART = Adjusted Reference Temperature 3
Ideal Scenario
- ETC provides conservative predictions of embrittlement
- Surveillance data covers all operating periods Potential Uncertainty Sources IF ETC under-predicts measurements IF Limited Surveillance Data is Available 40yr 60yr 80yr 40yr 60yr 80yr
?
Embrittlement (T)
Current
?
Embrittlement (T)
?
Future Data?
Operating Time (years)/Fluence Operating Time (years)/Fluence 4
Current Perspective of Potential Issue
- High confidence that currently operating plants remain safe
- Recent licensing actions remain valid
- Insufficient embrittlement monitoring and under predictions of reactor vessel embrittlement will eventually (after about 10 years for PTS; after about 23 years for P-T limits and upper shelf energy) impact the staffs confidence in the integrity of the reactor pressure vessel in long-term operation, i.e., both safety margins and performance monitoring may be impacted
- Further work is needed to determine which plants are impacted by this potential issue 5
Embrittlement Trend Curve
- May 1988, NRC published RG 1.99, which contained an improved embrittlement trend curve (ETC)
- Fit based on 177 datapoints
- June 1991, NRC updated 10 CFR 50.61 to include the ETC from RG 1.99
- Addressed lower than measured predictions (up to 60°F) of embrittlement in some vessels
- This ETC was re-evaluated for continued adequacy in 2014 (ML13346A003) and in more detail in 2019 (ML19203A089) 6
Issue - ETC
+180°F Deviates Statistically from mean significant
-180°F DT41J = T41J is a measurement of embrittlement representing the shift in transition temperature from brittle to ductile fracture at an impact toughness of 41J 7
Surveillance Capsule Delays
- Appendix H to 10 CFR Part 50 requires periodic monitoring of changes in fracture toughness caused by neutron embrittlement
- ASTM standard (E185-82) allows final capsule fluence to be 2X RPV design fluence - plants change (intended 40-year) design fluence to current license length (e.g., 60 or 80 years)
- ASTM standard (for 40 years) permits holding last capsule without testing
- Commission finding (Perry decision, NRC Administrative Letter 97-04) that staff review of requests to change capsule withdrawal schedules is limited to verification of conformance with the ASTM standard (i.e., not based on technical or safety considerations)
- Capsule withdrawal and testing repeatedly delayed in some cases to achieve higher fluence 8
Issue - Appendix H Performance Monitoring Many licensees have delayed capsules (time and/or fluence),
some recent examples:
Plant Capsule # of times
- delayed Turkey Point 5 4 Robinson 5 2 Surry U1 5 2 Surry U2 5 2 North Anna U1 4 2 North Anna U2 4 2 St. Lucie U2 4 1 Point Beach 5 1 Capsule withdrawal schedule changes include Not all plants have delayed delays in both time and/or fluence withdrawal of capsules 9
Potential Impact of Issue 400 Follows Fit with potential additional data underprediction 350 from Slide 7 300 No correction 75 F 150 F 250 Embrittlement DRTNDT,F 200 150 Data Potential additional data 100 RG1.99 Fit to original data Fit through Plant data 50 Updated fit with additional data 0
1.0E+17 2.0E+19 4.0E+19 6.0E+19 8.0E+19 1.0E+20 1.2E+20 1.4E+20 Fluence, n/cm^2 10
Recent Plant Surveillance Data 60 F 100 F 60 yr. 80 yr.
Fluence Fluence (est) 11
Safety Case
- Risk of Failure
- Conditional probability of failure during normal operation may increase several orders of magnitude (e.g., 3 orders of magnitude at 50F), but expected transient frequency makes overall risk low.
- Some plants may exceed Pressurized Thermal Shock screening limit in 10 CFR 50.61, but analyses suggest risk is low
- Large uncertainty exists - many plant specific details unknown
- Safety Margins
- Inaccurate embrittlement prediction and increasing uncertainty due to lack of surveillance decreases safety margins to failure - plant specific
- Performance Monitoring
- Delaying capsules does not provide adequate performance monitoring to ensure embrittlement trends are reasonable 12
Risk of Failure ESD represents the underprediction of RTNDT Large Uncertainties:
- Unknown frequency of transient
- Actual plant fluence variations
- Are these analyses bounding?
Unknown plant-specific considerations
- How much protection do administrative and other operational limits provide against violating the PT limit?
RG 1.99 Revision 2 Update FAVOR Scoping Study, May 6, 2021, TLR RES/DE/CIB-2020-09, Rev. 1, ML21126A326 13
Pressurized Thermal Shock Considerations
- 10 CFR 50.61 uses ETC from RG 1.99
- RTPTS from 10 CFR 50.61 might be impacted
- Limits of 270 °F for plates, forgings, and axial weld materials, and 300 °F for circumferential weld materials
- However, through-wall crack frequency calculated with corrected embrittlement less than 1x10-6 for all cases investigated 14
Safety Margins
- Uncertainties in risk calculations are high and increasing with time
- Even though the risk appears low, resolving these issues will help maintain the fundamental safety principles that are the basis of plant design and operation
- Safety margins, as provided by regulations and current license bases, provide reasonable assurance against brittle fracture 15
Safety Margins Illustration Structural limit Reduced Margin Adequate PT-curve using Margin RG 1.99 Pressure, psi Accurate Operating PT-curve Margin Margin Uncertainty Temp, F Uncertainties increasing due to lack of surveillance, but margin is less due to embrittlement underprediction 16
Analysis Summary
- With the current state of knowledge, a generalized analysis suggests the overall risk of brittle fracture is low
- The uncertainty in these results is high and increases with time
- Plant specific details not considered
- Under certain conditions, safety margins are impacted and are decreasing as uncertainty increases
- Delaying capsules at high fluence represents a lack of sufficient performance monitoring
- Issues are plants with fluences > 6x1019 n/cm2 17
Who is Impacted?
- Embrittlement Underprediction Percentage of Fleet Surpassing Fluence Levels Percentage of PWRs Surpassing Fluence Levels Year\Fluence 6 x 1019 n/cm2 8 x 1019 n/cm2 6 x 1019 n/cm2 8 x 1019 n/cm2 60 years 6% 0% 9% 0%
80 years 22% 10% 34% 15%
- Plant specific details (e.g., limiting material, etc.) may contribute to which plants are impacted
- More work is needed to determine which plants are impacted
- Lack of Surveillance Data
- Any plant renewing license that chooses to delay last capsule 18
Staff Alternatives
- Alternative 1 - Status Quo: Make no changes to Appendix H to 10 CFR Part 50, 10 CFR 50.61, or RG 1.99. Handle issues through plant-specific action and generic communications.
- Alternative 2 - Focused Solution: Revise Appendix H to 10 CFR Part 50 to include additional surveillance testing requirements for long-term operation, revise fluence function fit for only impacted RPV materials.
- Alternative 3 - Comprehensive Solution: Revise Appendix H to 10 CFR Part 50 to include additional surveillance testing requirements for long-term operation, update the applicable regulations (e.g., 10 CFR 50.61) to require all licensees to use an NRC-approved ETC that properly accounts for radiation effects, update RG 1.99 to contain an ETC with one that appropriately accounts for radiation effects, and update implementing guidance.
19
Backfit/Forward Fit Considerations
- Alternative 1 (Plant-specific actions) -
backfitting or forward fitting considered for each plant-specific case
- Alternatives 2 and 3 (Rulemaking)
- Modification of App H will require some plants to test capsule during SLR period
- Modification of embrittlement trend curve may require some plants to modify P-T limits and/or PTS evaluation 20
Staff Recommendation
- Alternative #2 - Focused Solution
- Address issues in a focused and risk-informed manner
- Target those plants with materials that are impacted by the underprediction issue
- Modify current surveillance testing requirements to ensure periodic performance monitoring
- Details of implementation to be worked out during regulatory basis effort 21
Rulemaking Schedule
- Alternative #2 Schedule
- Deliver regulatory basis to the Commission14 months after receipt of the Commission staff requirements memorandum (SRM) approving rulemaking.
- Deliver proposed rule to the Commission15 months after the regulatory basis is issued for public comment.
- Deliver final rule to the Commission15 months after the proposed rule is issued for public comment.
22
Summary
- High confidence that currently operating plants remain safe, and recent licensing actions remain valid
- Issue will eventually (after about 10 years for PTS and 23 years for P-T limits) impact the staff confidence in the integrity of the reactor pressure vessel in long-term operation, i.e., both safety margins and performance monitoring may be impacted
- Further work is needed to determine which plants are impacted by this issue
- Proactively ensure continued reasonable assurance though a risk-informed, performance-based solution
- Staff delivered rulemaking plan to Commission - desires focused solution to only those conditions adversely impacted by this issue 23
Thank You 24
This presentation offers the professional opinions of EPRI staff only and does not represent an Industry position of the U.S. utilities.
RPV Embrittlement Monitoring and Prediction in Long-Term Operation April 2022 EPRI MRP Technical Brief Elliot J. Long Principal Technical Leader ACRS Meeting April 6, 2022 www.epri.com © 2022 Electric Power Research Institute, Inc. All rights reserved.
Presentation Outline Future Sources of High Fluence Capsule Data
- PWR Coordinated Reactor Vessel Surveillance Program (CRVSP)
MRP-326, Revision 1
- PWR Supplemental Surveillance Program (PSSP)
MRP-412 Review of Prior EPRI MRP Conclusions from the November 2019 ACRS Meeting
- Potential to impact plant Pressure-Temperature (P-T) limit curves 2 © 2022 Electric Power Research Institute, Inc. All rights reserved.
Update to the CRVSP, MRP-326, Revision 1 3 © 2022 Electric Power Research Institute, Inc. All rights reserved.
Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)
Materials Issue Being Addressed:
- Optimize the U.S. PWR surveillance capsule withdrawal schedules to increase the amount of high-fluence (f > 3.0 x 1019 n/cm2) surveillance data which can be used to inform development of embrittlement trend correlations (ETCs) applicable for RPV operation to high fluence (60+
years)
Objectives of the Project
- Revision 0 (2011): Review the reactor vessel surveillance programs (RVSPs) of the operating U.S. PWR fleet and recommend changes to selected RVSP withdrawal schedules in order to increase the amount of high fluence surveillance data by 2025
- Revision 1 (2021): Review of how we did, what has occurred, whats left to do, and when it is most likely to happen across the US fleet Updates to the evaluation include
- Evaluated capsules withdrawn since 2011
- Future capsule pull schedules
- Capsule fluence values
- Analysis of closed (or to be closed) plants 4 © 2022 Electric Power Research Institute, Inc. All rights reserved.
Update to the CRVSP, MRP-326, Revision 1 Current high fluence capsule withdrawal results
- 16 out of 30 CRVSP Capsules are tested or planned to be tested
- There are 14 remaining CRVSP Capsules Half of these are not planned to be tested (i.e., due to plant shutdown) or will be delayed beyond 2025 Summary of available high fluence data
- 48 U.S. capsules have been tested at f > 3.0 x 1019 n/cm2 4 of these are f > 8.0 x 1019 n/cm2
- By 2025, the remaining 7 planned CRVSP capsules will be tested at f > 3.0 x 1019 n/cm2 2 of these are predicted to be f > 8.0 x 1019 n/cm2 5 © 2022 Electric Power Research Institute, Inc. All rights reserved.
PWR Supplemental Surveillance Program (PSSP) 6 © 2022 Electric Power Research Institute, Inc. All rights reserved.
PWR Supplemental Surveillance Program (PSSP)
Materials Issue Being Addressed:
- Additional high-fluence (f > 5.0 x 1019 n/cm2) surveillance data is needed to inform development of embrittlement trend correlations (ETCs) applicable for RPV operation to high fluence (60+ years)
Objectives of the Project:
- Fill projected gaps in the tested surveillance capsule database
- Inform future ETCs using actual RPV surveillance materials from commercial PWRs (not test reactor data)
Project Goal: Irradiate two supplemental surveillance capsules for ~10 total years before withdrawal, testing, evaluation and publication of capsule test reports
- These two surveillance capsules have 288 Charpy Specimens from 27 unique plates, forgings and welds
- The data generated from these capsules will ultimately yield 24 new transition temperature shift results and 3 additional upper shelf energy results Fluence levels of the to be evaluated specimens will range from ~4.5E+19 up to ~1.2E+20 (n/cm2) 7 © 2022 Electric Power Research Institute, Inc. All rights reserved.
PWR Supplemental Surveillance Program Project History
- Program designed and fabricated 2 supplemental surveillance capsules containing previously-irradiated, reconstituted PWR materials
- EPRI MRP sponsored the fabrication of these 2 surveillance capsules:
ALA-P; 14 materials (Host: Farley 1), inserted October 2016 CQL-P; 13 materials (Host: Shearon Harris),
inserted April 2018
- MRP-412 (PSSP Capsule Fabrication report) was published in 2016 8 © 2022 Electric Power Research Institute, Inc. All rights reserved.
PWR Supplemental Surveillance Program Current Project Status and Timeline
- Farley 1 Capsule P to be withdrawn in Spring 2027; Shearon Harris Capsule P in Fall 2028 per MRP-326,R1
- Testing of surveillance capsules and data evaluation in 2028-2030
- Anticipated Project Deliverable Date:
Capsule report within ~18 months of each capsules withdrawal date (2 reports total)
- Data evaluation and impact on future ETCs in 2030-2032 PSSP Capsule seated in its holder 9 © 2022 Electric Power Research Institute, Inc. All rights reserved.
Prior EPRI Conclusions on the Potential Revision of RG1.99R2 10 © 2022 Electric Power Research Institute, Inc. All rights reserved.
Conclusions from the EPRI Presentation to the ACRS in November 2019 EPRI MRP previously presented on the potential revision of RG1.99R2 to the ACRS in Nov. 2019 The conclusions from that meeting (shown at right) have not changed If a future revision to RG.199R2 is implemented, ASTM E900-15 remains the preferred ETC model (today)
It is understood that for fluence values below 6E+19 n/cm2, RG1.99R2 remains adequate for predicting RPV embrittlement The next slide details when certain plant designs will see that fluence level at 1/4T 11 © 2022 Electric Power Research Institute, Inc. All rights reserved.
What surface fluence = 6E+19 @ 1/4T and 3/4T?
The chart at right was developed using Design RPV Maker Vessel T (in)
Surface Fluence Needed to Reach 6E+19 @
1/4T (E+19) 3/4T (E+19) the current fluence attenuation formula WEC 2-Loop B&W/CE 6.5 8.86 19.3 WEC 3-Loop/Smaller CE CE 7.875 9.62 24.8 from RG1.99R2* B&W NSSS B&W 8.44 9.96 27.4 WEC 4-Loop RDM 8.45 9.96 27.5 It is understood that the 3-Loop WEC 4-Loop WEC 4-Loop/Larger CE B&W CE 8.5 8.625 9.99 10.1 27.7 28.3 Westinghouse design has the highest CE - Special CE - Sys80 CE CE 8.79 11.2 10.2 11.7 29.2 45 surface fluence at end-of-life SLR plants most limiting surface fluence values are also summarized herein Only Plant A will hit a 1/4T fluence of Plant Design EFPY Surface Fluence (E+19) Potentially Impacted?
A WEC 3-Loop 72 10.8 at ~65 EFPY 6E+19 n/cm2 for 80-years B WEC 3-Loop 68 7.26 No C Smaller CE 72 6.56 No
- This is currently predicted to occur well into D E
WEC 3-Loop WEC 2-Loop 72 72 7.34 7.80 No No the SLR operating period F B&W 72 2.02 No
- Note that BWR Units have an NRC approved Integrated Surveillance Program (ISP) for up to 60 years for the U.S. BWR Fleet (BWRVIP-86, Rev. 1-A). The implementation plan for Subsequent License Renewal (SLR) has also been accepted by NRC in report BWRVIP-321-A.
The highest 1/4T SLR fluence value is less than 1 x 1019 n/cm2 (E > 1.0 MeV) for the U.S. BWR fleet.
12 © 2022 Electric Power Research Institute, Inc. All rights reserved.
TogetherShaping the Future of Energy' 13 © 2022 Electric Power Research Institute, Inc. All rights reserved.
NRC - CNSC Joint Report The Canadian Nuclear Safety Commission (CNSC) and the NRC have issued a joint report on theBWRX-300 Containment Evaluation Method as part of their ongoing cooperation aimed at enhancing technical reviews of advanced reactor and small modular reactor technologies.
The report is based on the joint review of the licensing topical report GEH submitted for NRC review approval.
Both regulators evaluated how the information would be used in each countrys regulatory infrastructure and confirmed items that could be documented for mutual understanding.
Nothing in the joint report fetters the powers, duties or discretion of CNSC or NRC designated officers, CNSC or NRC inspectors or the respective Commissions regarding making regulatory decisions or taking regulatory action.
April 6, 2022
ENCLOSURE 1 M220049 ACRS Full Committee Presentation Slides for NEDC-33922P, BWRX-300 Containment Evaluation Method Licensing Topical Report Non-Proprietary Information
ACRS Full Committee Presentation GE-Hitachi (GEH)
Licensing Topical Report (LTR) NEDC-33922P BWRX-300 Containment Evaluation Method (Open Session)
April 6, 2022
LTR Purpose and Scope
- GEH is seeking NRC approval for application of an analysis method to be used for evaluating the BWRX-300 dry containment thermal hydraulic performance.
- The LTR scope includes
- Method description and qualification
- Sensitivity studies
- Demonstration cases
- The analysis method used for the BWRX-300 containment thermal hydraulics performance demonstrates that the containment design complies with 10 CFR 50, Appendix A, General Design Criteria 2, 4, 16, 38, 41, 50, and 51.
- Details of the containment performance acceptance criteria are listed in Section 4.0 of NRC-approved LTR NEDC-33911P, BWRX-300 Containment Performance
- GEH is not requesting NRC approval for exemptions from any regulatory requirements.
Copyright 2022 GE Hitachi Nuclear Energy Americas, LLC - All Rights Reserved 2
BWRX-300 Containment Design Features The following containment design features are relevant to the purposes of the LTR:
- Dry enclosure, near atmospheric pressure during normal operation
- Inerted with nitrogen during normal operation
- Design pressure and temperature are within the experience base of conventional BWRs
- No subcompartments containing large bore high energy lines
- The subcompartments have sufficiently large openings such that the boundaries of the subcompartments do not experience large pressure differentials resulting from pipe breaks outside the subcompartments Copyright 2022 GE Hitachi Nuclear Energy Americas, LLC - All Rights Reserved 3
Limiting Pipe Breaks
- The limiting large breaks are:
- Main steam pipe
- Feedwater pipe
- All design basis large breaks are rapidly isolated at the RPV nozzle.
- The limiting small breaks are unisolated instrument line breaks, either in the steam or liquid space.
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Overview of the Evaluation Model TRACG used to evaluate the mass and energy release
- Applies the ESBWR TRACG-LOCA method
- Performed both base and conservative demonstration cases:
- Main Steam and Feedwater Large Breaks
- Small Steam and Liquid Pipe Break Cases GOTHIC used to evaluate containment response
- New containment model developed for BWRX-300
- Code has been benchmarked to separate effect and integral tests. Benchmarking to the test data of a similar size containment is included in the LTR.
- Performed both base and conservative cases
- Same approach that was taken for the ESBWR containment method Copyright 2022 GE Hitachi Nuclear Energy Americas, LLC - All Rights Reserved 5
Conclusion In summary
- GEH is seeking NRC approval for application of an analysis method to be used for evaluating the BWRX-300 dry containment thermal hydraulic performance.
- GEH is not requesting NRC approval for exemptions from any regulatory requirements.
- TRACG utilizes the applicable parts of the TRACG Application for ESBWR approved LTR, which is incorporated in the approved ESBWR Design Certification
- Utilizes GOTHIC, a standard code used for evaluating thermal-hydraulic containment response in the nuclear industry
- Individual key inputs, assumptions and modeling parameters conservatively biased simultaneously in the conservative cases (same approach taken for the ESBWR containment method)
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Questions or Comments Regulatory Review of GEH Topical Report BWRX-300 Containment Evaluation Method NEDC-33922P, Revision 2 NRC Staff Presentation BWRX-300 Small Modular Reactor ACRS Full Committee Meeting April 6, 2022 1
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NRC Staff Review Team
- NRR Nuclear Systems Performance Branch (SNSB)
- NRR New Reactor Licensing Branch (NRLB)
- NRR Nuclear Methods, Systems & New Reactors Branch (SNRB)
- NRR Containment and Plant Systems Branch (SCPB)
- Chang Li
- RES Code and Reactor Analysis Branch (CRAB)
- Peter Lien, Joe Staudenmeier, Andrew Ireland
- RES Fuel & Source Term Code Development Branch (FSTCB)
- Shawn Campbell 2
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Presentation Outline
- Overview of BWRX-300 Containment Evaluation Method (CEM) LTR NEDC-33922P, Revision 2
- BWRX-300 acceptance criteria for containment response
- BWRX-300 containment design background
- BWRX-300 containment evaluation method demonstration analyses
- TRACG mass and energy release calculation methodology review
- GOTHIC containment response calculation methodology review
- Confirmatory analysis approach and results
- Resulting four limitations and conditions
- Conclusions 3
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NRC Staff Review of the LTR
- The purpose of GEH LTR NEDC-33922P, Revision 2, is to obtain NRC staff approval of the BWRX-300 containment evaluation method (CEM) for peak containment pressure and temperature analysis
- The approved methodology will be used to design the BWRX-300 containment, and support a license application for a CP and OL under 10 CFR 50, or a DCA and COL under 10 CFR 52
- BWRX-300 LTR Acceptance Criteria Accident pressure and temperature are less than design pressure and temperature with appropriate margin Containment pressure is reduced to less than 50% of the peak accident pressure for the most limiting LOCA within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Containment pressure responses after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for LOCAs that do not produce the peak accident pressure are maintained below 50% of the peak pressure for the most limiting LOCA Containment atmosphere remains sufficiently mixed such that deflagration or detonation does not occur inside containment 4
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BWRX-300 Containment Design Background
- BWRX-300 has a nitrogen-inerted, dry containment
- No suppression pool inside the containment
- Passive Containment Cooling System (PCCS)
Long-term containment SBLOCA pressure reduction/mitigation Demonstration analyses with specific PCCS units described in the LTR
- Reactor cavity pool for containment heat removal
- Containment dome interfacing with the reactor cavity pool 5
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BWRX-300 Containment Evaluation Method Demonstration Analyses
- Containment analysis method for BWRX-300 thermal-hydraulic performance is used to demonstrate that the containment design satisfies the acceptance criteria for:
Large-Break Loss-of-Coolant Accident (LBLOCA)
Small-Break Loss-of-Coolant Accident (SBLOCA)
- Analyzed containment DBAs include liquid and steam breaks
- Acceptance criteria were satisfied for the LTR demonstration cases
- Applicant used TRACG code to calculate the mass and energy release, and GOTHIC code to calculate the containment response
- NRC Staff used TRACE and MELCOR to develop models to perform confirmatory analyses 6
TRACG Code - Overview for BWRX-300
- Overview of TRACG code Latest TRACG versions used in analysis, no significant changes since ESBWR RPV model and internal components scaled from ESBWR De-coupled method assumes containment remains at atmospheric pressure
- Past TRACG approval and relevance to BWRX-300 ESBWR qualification extended to BWRX-300, such that ESBWR PIRT and model biases applied for RPV and internals BWR/2-6 methods evoked since some events result in core uncovery ICs safety function changed and modeled in considerably more detail Modeling deemed adequate for M&E release calculations (w/ L&Cs applied) 7 We make SAFE use of nuclear technology POSSIBLE.
TRACG Code - Mass and Energy Release Calculation Methodology
No suppression pool ICs are the primary decay heat removal path
- RPV isolation valves limits break flow and M&E release for large piping but small breaks are un-isolated and continue blowdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- One ICS train inoperative (due to limiting single failure)
- Conservative inputs for initial power level, power history, scram time, choke flow model, atmospheric pressure break boundary condition and bounding operating conditions 8
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TRACG Code - Mass and Energy Release Calculation Methodology Significant Issues and Resolution RAI - Radiolytic gas accumulation and removal in the ICs L&C 1: total volumetric fraction of radiolytic gases in the IC lower drum limited to a sufficiently low level such that condensation heat transfer in the ICs is not adversely affected and the hydrogen deflagration margin is maintained RAI - ICs return line steam trap L&C 2: IC return line layout must include a loop seal, or water trap, that prevents reverse flow from RPV back into the IC return line 9
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GOTHIC Code Overview for BWRX-300
- Overview of GOTHIC code An established industry code widely used for containment response analysis 10 CFR Part 50, Appendix B compliant code Latest GOTHIC version 8.3 used in the BWRX-300 analysis
- Past GOTHIC approval and relevance to BWRX-300 GOTHIC previously approved for containment response analysis BWRX-300 containment PIRT consistent with GOTHIC functionalities BWRX-300 relevant GOTHIC benchmarking against CVTR test data reviewed GOTHIC is qualified for the thermal stratification and 3D effects 10 We make SAFE use of nuclear technology POSSIBLE.
BWRX-300 GOTHIC Containment Response Methodology
- Based on Reg Guide 1.203, "Transient and Accident Analysis Methods
- Decoupled M&E release from the TRACG RPV model with no backpressure as a containment BC for the stand-alone GOTHIC containment model
- 4-component GOTHIC model Containment (nodalized)
Dome (nodalized)
PCCS (nodalized)
Reactor Cavity Pool (lumped)
- Conservative Diffusion Layer Model (DLM) used for condensation
- Thermal stratification inside the containment 11 We make SAFE use of nuclear technology POSSIBLE.
Staff Review of the BWRX-300 GOTHIC Containment Response Methodology
- Physical phenomena (GOTHIC PIRT)
- GOTHIC input model (Nominal inputs, assumptions, and correlations)
- Key modeling uncertainties and conservative biases -- Overall GOTHIC model conservatism
- Nodalization sensitivity studies for the containment and PCCS
- GOTHIC benchmarking of the test data
- BWRX-300 containment sensitivity analyses for large/small breaks
- PCCS capacity to mitigate the containment pressure in the long-term
- Containment mixing for combustible gases 12 We make SAFE use of nuclear technology POSSIBLE.
Significant Containment-specific Issues & Resolutions
- Break location & and break flow orientation sensitivities Limiting PCP LBLOCA location and orientation modified A liquid, and not steam, SBLOCA is limiting
- Sensitivity to containment nodalization Potential for reverse flow and non-condensable gas return to RPV L&C #3 -No break flow reversal or reversal not safety significant
- Containment heat transfer modeling Differences b/w confirmatory & GOTHIC Dome & PCCS heat transfers Justification for the condensation/natural convection heat transfer correlations An error identified & corrected in GOTHIC PCCS condensation modeling
- PCCS modeling and nodalization sensitivity study Confirmatory sensitivity study performed L&C #4 - Applicability to the final PCCS design for licensing basis 13 We make SAFE use of nuclear technology POSSIBLE.
Confirmatory Analysis Approach and Results Containment Reactor, IC, MS, FW GOTHIC M & E (standalone)
TRACG TRACE-CONTAN TRACE-VESSEL Atmosphere MELCOR Reactor, IC, MS, FW M & E (decoupled) Containment TRACE TRACE-CONTAN M & E (coupled)
- The proposed methodology using TRACG/GOTHIC codes is conservative
- Containment response is sensitive to nodalization, break location/orientation
- The accumulation of radiolytic gases during LOCA is possible (L&C#1)
- The ICs return line water trap is needed (L&C#2)
- Flow reversal from containment to RPV is possible (L&C#3) with insufficient 14 PCCS heat removal capacity (L&C#4) 14 We make SAFE use of nuclear technology POSSIBLE.
RPV-specific Limitations and Conditions
- L&C #1 The use of this CEM is limited to a BWRX-300 design that limits the total volumetric fraction of radiolytic gases in the IC lower drum to a sufficiently low level throughout a 72-hour period following the event such that condensation heat transfer in the ICs is not adversely affected and the hydrogen deflagration margin is maintained.
- L&C #2 The use of this CEM is limited to a BWRX-300 design that a proper isolation condenser return line layout is chosen, such as a loop seal or a water trap, to prevent reverse flow from RPV into the IC return line throughout a 72-hour period following the event or where an applicant or licensee referencing this report demonstrates that the TRACG code is capable of conservatively modeling the overall ICs heat removal capacity when reverse flow occurs in the IC discharge lines.
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Containment-specific Limitations and Conditions
- L&C #3.
The use of this CEM is limited to a BWRX-300 design in which the PCCS is sized sufficiently large such that a reverse flow from containment back to RPV does not occur during the first 72-hours into the event. The applicant or licensee referencing this report needs to demonstrate that no reverse flow could occur, or any reverse flow that occurs under the most bounding flow reversal conditions resulting in the degradation of IC heat transfer is not safety-significant with respect to the acceptance criteria for the BWRX-300 CEM.
- L&C #4.
The use of this CEM was demonstrated for a BWRX-300 design with the PCCS described in this LTR. For any alternate PCCS design configuration and placement, the applicability of this method and the PCCS modeling approach must be reviewed and found to be acceptable by the NRC for BWRX-300 licensing-basis analyses.
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Conclusions
- The proposed BWRX-300 analytical approach and TRACG/GOTHIC modeling described in the LTR for M&E release and containment response are acceptable, with the appropriate conservative biases and modeling inputs to address the model uncertainties.
- With the four NRC L&Cs specified in the staff SER Section 7.0, the CEM presented in GEH LTR NEDC-33922P, Revision 2, is acceptable for BWRX 300 peak containment pressure and temperature analysis of the containment DBAs.
- The NRC staff will evaluate the regulatory compliance of the final BWRX-300 containment design using the CEM during the future licensing activities, in accordance with 10 CFR Part 50 or 10 CFR Part 52, as applicable.
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