ML19184A158
| ML19184A158 | |
| Person / Time | |
|---|---|
| Issue date: | 05/15/2019 |
| From: | Michael Snodderly Advisory Committee on Reactor Safeguards |
| To: | |
| Snodderly, M, ACRS | |
| References | |
| NRC-0328 | |
| Download: ML19184A158 (337) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards NuScale Subcommittee: Open Session Docket Number: (n/a)
Location: Rockville, Maryland Date: Wednesday, May 15, 2019 Work Order No.: NRC-0328 Pages 1-211 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5 (ACRS) 6 + + + + +
7 NuSCALE SUBCOMMITTEE 8 + + + + +
9 OPEN SESSION 10 + + + + +
11 WEDNESDAY 12 MAY 15, 2019 13 + + + + +
14 ROCKVILLE, MARYLAND 15 + + + + +
16 The Subcommittee met at the Nuclear 17 Regulatory Commission, Two White Flint North, Room 18 T2D10, 11545 Rockville Pike, at 8:30 a.m., Matthew W.
19 Sunseri, Michael L. Corradini, and Vesna B.
20 Dimitrijevic, Co-Chairs, presiding.
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2 1 COMMITTEE MEMBERS:
2 MICHAEL L. CORRADINI, Co-Chair 3 VESNA B. DIMITRIJEVIC, Co-Chair 4 MATTHEW W. SUNSERI, Co-Chair 5 RONALD G. BALLINGER, Member 6 DENNIS BLEY, Member 7 CHARLES H. BROWN, JR. Member 8 JOSE MARCH-LEUBA, Member 9 JOY L. REMPE, Member 10 GORDON R. SKILLMAN, Member 11 12 ACRS CONSULTANT:
13 STEPHEN SCHULTZ 14 15 DESIGNATED FEDERAL OFFICIAL:
16 MIKE SNODDERLY 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 CONTENTS 2 Opening Remarks . . . . . . . . . . . . . . . . . 4 3 Staff Opening Remarks . . . . . . . . . . . . . . 5 4 Chapter 19.0, "Probabilistic Risk Assessment 5 and Severe Accident Evaluation," and 19.1, 6 "Probabilistic Risk Assessment (PRA)," NuScale 7 Design Certification Application (DCA) . . . . . 8 8 Chapter 19.1 PRA Safety Evaluation (SE) with 9 Open Items . . . . . . . . . . . . . . . . . . 161 10 Public Comments . . . . . . . . . . . . . . . . 194 11 Subcommittee Discussion . . . . . . . . . . . . 195 12 Adjourn . . . . . . . . . . . . . . . . . . . . 211 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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4 1 P R O C E E D I N G S 2 (8:30 a.m.)
3 CO-CHAIR SUNSERI: Good morning. The 4 meeting has now come to order. This is a meeting of 5 the Advisory Committee on Reactor Safeguards, NuScale 6 Subcommittee.
7 I am Matt Sunseri, Co-Chairperson for 8 today's Subcommittee meeting, along with Mike 9 Corradini and Vesna Dimitrijevic.
10 Members in attendance today are Dr.
11 Dimitrijevic, Dennis Bley, Jose March-Leuba, Joy 12 Rempe, Matt Sunseri, Michael Corradini, Dick Skillman, 13 Ron Ballinger, and Charlie Brown will be joining us 14 shortly.
15 He is held up in traffic, but will be here 16 soon. We also have our consultant, Steve Schultz, 17 with us today and did I get everybody? Yes. All 18 right.
19 Mike Snodderly is the designated federal 20 official for this meeting. The Subcommittee will 21 continue its review of the staff's evaluation of 22 Chapter 19, Probabilistic Risk Assessment and Severe 23 Accident Evaluation and we will review the staff's 24 evaluation of Chapter 21, Multimodal Design 25 Considerations of the NuScale Design Certification NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 1 Application.
2 Today we have members of the NRC staff and 3 NuScale to brief the Subcommittee. The ACRS was 4 established by statute and is governed by the Federal 5 Advisory Committee Act. That means that the Committee 6 can only speak through its published letter reports.
7 We hold meetings to gather information to 8 support our deliberations. Interested parties who 9 wish to provide comments can contact our office 10 requesting a time after the meeting announcement is 11 published in the Federal Register.
12 That said, we set aside ten minutes for 13 comments from members of the public attending or 14 listening to our meeting. Written comments are also 15 welcome.
16 The ACRS Section of the U.S. NRC public 17 website provides our charter, bylaws, letter reports 18 and full transcripts of full and Subcommittee 19 meetings, including slide presented there.
20 The rules for participation at today's 21 meetings were announced in the Federal Register on May 22 6th, 2019. The meeting was announced as an open and 23 closed meeting.
24 We may close the meeting after open 25 portions to discuss proprietary material and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 presenters can defer questions that should not be 2 answered in the public session to that time.
3 No written statements or requests for 4 making any oral statements to the Subcommittee have 5 been received from the public concerning this meeting.
6 A transcript of the meeting is being kept 7 and will be made available as stated in the Register 8 -- in the Federal Register notice. Therefore, we 9 request that participants in this meeting use the 10 microphones located throughout the meeting room when 11 addressing the Subcommittee.
12 Participants should first identify 13 themselves and speak with sufficient clarity and 14 volume so they can be readily heard. We have a bridge 15 line established for the public to listen in.
16 To minimize disturbance, the public line 17 will be kept in a listen-in only mode. To avoid 18 disturbances, I request that the attendees now put any 19 electronic devices, like, cell phones in the off or 20 noise-free mode.
21 So before we begin today's topics, I'd 22 like to pick up on a continuation from yesterday. We 23 had a discussion at the end of the -- yesterday about 24 acquiring information regarding reliability of some 25 components.
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7 1 And I'm going to ask Mike Snodderly to 2 pick this up and walk us through that issue and where 3 we are as far as getting the information we need.
4 Mike?
5 MR. SNODDERLY: Thank you. Thank you.
6 Yesterday there was --
7 (Simultaneous speaking.)
8 CO-CHAIR SUNSERI: Is your mic on?
9 (Simultaneous speaking.)
10 MR. SNODDERLY: Hopefully, well --
11 PARTICIPANT: Yes. You have to talk 12 straight into it, Mike, and that'll do it.
13 MR. SNODDERLY: Let's use this one. So 14 there was some discussion about the availability of 15 the NuScale report Passive Safety System Reliability 16 PRA Report.
17 This report was audited as part of the 18 first audit and it is not on the docket. But there 19 were -- there are four slides that were a part of the 20 October 18th. So yes.
21 We're going to get those four slides to 22 Dr. Bley and to Dr. Dimitrijevic. Also, because the 23 Committee does not have access to that report at this 24 time, we'd like to request that at the June full 25 Committee that the staff be prepared to provide NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 further details from their notes on what -- on their 2 findings or what they looked at that gave them 3 confidence in the reliability numbers used in that 4 report.
5 Also, the -- so just wanted to make sure 6 that that was the Committee's understanding of the 7 availability of that report and us -- and going 8 forward. Okay. Thank you.
9 CO-CHAIR SUNSERI: Michelle, do you have 10 any comments you want to make before we get started?
11 MICHELLE: I'm very good, thank you.
12 CO-CHAIR SUNSERI: All right. Now I turn 13 to Sarah Bristol for the start of the presentation 14 today. Sarah?
15 MS. NORRIS: Good morning, everyone.
16 Welcome back. This is actually Rebecca Norris with 17 NuScale. So today we're going to be covering three 18 sections of Chapter 19.
19 Starting with Chapter 19.2 or the break 20 with 21 and then 19.3 and 19.5. We have added one 21 person on our presentation team. So I'd like everyone 22 to introduce themselves again for the Members.
23 MR. RAD: All right. This is Zac Rad, 24 Director of Regulatory Affairs, NuScale Power.
25 MR. MULLIN: Hi. This is Etienne Mullin.
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9 1 I'm a PRA analyst.
2 MS. BRISTOL: Sarah Bristol, supervisor of 3 the PRA group.
4 MR. GALYEAN: Bill Galyean, NuScale PRA.
5 MS. NORRIS: So first, we're going to be 6 covering 19.2, Severe Accident Evaluation and with 7 that, I'll let Etienne take it away.
8 MR. MULLIN: All right. So today we're 9 going to talk about severe accidents briefly on topic.
10 I did not bring as many slides as I did in October and 11 they're pretty high level. So I hope this spurs some 12 discussion.
13 I'm sure you all won't hesitate to speak 14 up. So in Chapter 19.2, we evaluated severe accidents 15 for the potential to challenge containment integrity 16 and thereby result in a large release per our 17 definition.
18 As we discussed yesterday, we define a 19 large release as a source term that results in a dose 20 of 200 rem at the site boundary for a stationary 21 individual, standing there for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, in the worst 22 possible location.
23 We used the MACCS code, that's MELCOR 24 Accident Consequence Code System, to evaluate the 25 outside consequence for our worst case intact CNV NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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10 1 release and that's assuming a certain leak rate from 2 the containment.
3 And determined from that calculation that 4 there's no way that an intact containment core damage 5 event could possibly result in a large release with 6 orders or magnitude of margin.
7 MEMBER BLEY: How did you pick the leak 8 rate that you're using?
9 MR. MULLIN: We have a -- there's a 10 technical specification we create that has to be met 11 at .25.
12 MEMBER BLEY: Okay.
13 MR. MULLIN: Yes.
14 MEMBER BLEY: So we meet the standard?
15 MR. MULLIN: Yes. And from that result, 16 we decided, as a simplification, to take all of our 17 intact CNV cases, assign them a not-large release and 18 the Level 2 event trees and all of our bypass 19 accidents as a simplification or a large release.
20 Now that's not to say that we evaluated 21 all of our bypass accidents and calculated doses and 22 determined them to be large releases. It's just a 23 simplification. So, therefore, in our severe accident 24 evaluation --
25 MEMBER BLEY: I'm sorry. I'm just --
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11 1 MR. MULLIN: Go ahead.
2 MEMBER BLEY: -- hanging on that. I 3 remembered some other places where I've heard things 4 like this. So the tech spec leak rate means you can 5 have up to that leak rate and keep operating the 6 facility. So you assume there's no -- essentially 7 assume there's no damage when this falls over?
8 MR. MULLIN: Yes.
9 MEMBER BLEY: It's leaking at whatever 10 leak rate it's --
11 (Simultaneous speaking.)
12 MR. MULLIN: We're not talking about 13 module drop.
14 MEMBER BLEY: Okay.
15 MR. MULLIN: This is it internalize --
16 MEMBER BLEY: But this is -- okay. And 17 you've -- you check this every refueling or something?
18 MR. MULLIN: I believe so. Yes.
19 MEMBER BLEY: Okay.
20 MR. MULLIN: I think each individual 21 penetration is tested and then they sum them together 22 to get leak rate.
23 MEMBER BLEY: Thanks.
24 MR. MULLIN: So we've evaluated the 25 phenomenological challenges to containment integrity NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 that have been evaluated in the industry for years 2 now.
3 By virtue of using the same fuel, same 4 coolant, similar materials, you know, we were unable 5 to identify any new phenomenological challenges 6 introduced by the NuScale design.
7 So the challenges that we looked at are 8 steam explosion, in-vessel, ex-vessel, hydrogen 9 combustion, high pressure melt ejection and direct 10 containment heating. That's another ex-vessel 11 challenge --
12 MEMBER BLEY: I'm sorry. I'm still 13 hanging on that. So even if you have a steam 14 explosion in-vessel, you assume there's no damage to 15 the containment from that?
16 MR. MULLIN: Well we evaluate the 17 potential for a challenge to containment integrity 18 from a steam explosion.
19 MEMBER BLEY: Okay. So you have some 20 chance that you do more damage than be allowed?
21 Correct?
22 MR. MULLIN: Well --
23 MEMBER BLEY: Are there other accidents 24 for which that's true? That you look for 25 consequential containment damage?
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13 1 MR. MULLIN: Yes. I mean, we looked at 2 all of these challenges from steam explosion to 3 hydrogen combustion to direct containment heating, 4 melt through.
5 MEMBER BLEY: And what approach did you 6 use to calculate the probability -- well the chance 7 that you break the containment or break it more than 8 it was leaking?
9 MR. MULLIN: Well we evaluated all these 10 phenomena. Determined it's --
11 MEMBER BLEY: So it's a judgment, 12 engineering judgment?
13 MR. MULLIN: Well calculations are 14 involved.
15 MR. GALYEAN: Dennis, you're getting ahead 16 of him.
17 MR. MULLIN: Yes.
18 MR. GALYEAN: Okay. At this point --
19 MR. MULLIN: Okay.
20 MR. GALYEAN: -- he's just saying that if 21 for a nominally intact containment, okay, that --
22 MEMBER BLEY: So we're going to get to 23 other accidents --
24 MR. GAYLEAN: There's no way --
25 (Simultaneous speaking.)
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14 1 MEMBER BLEY: -- and talk about --
2 MR. GALYEAN: Yes.
3 MR. MULLIN: We're going to start talking 4 about --
5 MEMBER BLEY: When you get there, I'm 6 interested in how you -- what kind of calculations you 7 use to come up with a whole size, if you will, and a 8 probability --
9 (Simultaneous speaking.)
10 MR. GALYEAN: And again, we'll get to 11 that.
12 MEMBER BLEY: Okay.
13 MR. GALYEAN: We'll get to that.
14 MEMBER BLEY: I'm happy to wait.
15 MR. GALYEAN: Okay.
16 MEMBER BLEY: Don't let me forget.
17 (Laughter.)
18 DR. SCHULTZ: One question. On the 19 simplification of the bypass accidents counted as 20 large releases. Is that something you expect to get 21 back to later or is it a simplification that is not to 22 be addressed again?
23 MR. MULLIN: You can go ahead.
24 MR. GALYEAN: Yes. It's a simplification, 25 because we can easily meet the NRC acceptance criteria NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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15 1 with that simplification. Okay? And so there's no 2 motivation to do detailed calculations when we don't 3 have to, basically.
4 If the time comes and we need to recapture 5 some margin, so to speak, we certainly have the option 6 to revisit this issue.
7 But at this point in time, we're simply 8 making the assumption that if there's a bypass or some 9 mechanism that would fail the containment integrity, 10 okay, we're just at this point in time just 11 classifying that as a large release just as a modeling 12 simplification.
13 There is one exception, which Etienne will 14 get to with regard to a dropped module case, but just 15 to avoid having to do all the calculations for all the 16 different action scenarios that we do have in the PRA, 17 we're simply making the simplification that if it's a 18 bypass or a failure of containment isolation, okay --
19 DR. SCHULTZ: Uh-huh.
20 MR. GALYEAN: -- that it automatically 21 goes to a large release. Okay? If the containment is 22 nominally intact, okay, it's not a large release and 23 that's as far as, you know, that's the distinction 24 we're making at this point in time.
25 DR. SCHULTZ: All right. Thank you.
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16 1 MR. MULLIN: And to add to that, when it 2 comes to actually evaluating offsite consequences, 3 yet, for the environmental report for emergency 4 planning zone purposes, we are evaluating the actual 5 offset consequences of the accident. We're not just 6 binning them as large release versus not large 7 release.
8 DR. SCHULTZ: Okay. That gets more to my 9 point. Thank you.
10 MEMBER BLEY: You'll explain that in more 11 detail later? Or if not, please do a little bit now.
12 MR. MULLIN: It might come up in 13 conversation. I don't have the slide --
14 MEMBER BLEY: Okay.
15 (Simultaneous speaking.)
16 MEMBER BLEY: It's up in conversation now.
17 MR. MULLIN: Yes.
18 MEMBER BLEY: Okay. You say you -- I'm 19 not sure what your statement meant. For every 20 accident, you consider a release fraction or 21 something? What are you doing?
22 MR. GALYEAN: Yes. We're straying out of 23 the Chapter 19.2 here.
24 MEMBER BLEY: Not -- where does it show 25 up?
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17 1 MR. GALYEAN: Etienne talked about the 2 environmental report. Okay? That's a separate 3 application --
4 MEMBER BLEY: Yes. It is.
5 MR. GALYEAN: -- it's a separate --
6 (Simultaneous speaking.)
7 MEMBER BLEY: We don't review that.
8 MR. GALYEAN: Okay. And in the 9 environmental report, we do what you call a SAMDA 10 analysis report, you know.
11 MEMBER BLEY: Right.
12 MR. GALYEAN: And there, we do actually do 13 detailed dose calculations in order to support the 14 dose benefit -- the cost benefit analysis as part of 15 the SAMDA analysis.
16 But that's not -- we're not going to talk 17 about here.
18 MEMBER BLEY: So until --
19 MR. GALYEAN: It's not a Chapter 19 --
20 (Simultaneous speaking.)
21 MEMBER BLEY: So the person who buys one 22 of these from you gets the fuel load. You won't have 23 a Level 3 PRA.
24 MR. GALYEAN: Right. Yes. Right now, 25 again, we did the calculation for the environmental NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 1 report.
2 MEMBER BLEY: Okay.
3 MR. GALYEAN: That's not what we're going 4 to talk about here.
5 MEMBER BLEY: Okay.
6 DR. SCHULTZ: But the second piece, then, 7 Etienne mentioned was the offsite release, the site 8 boundary determination and so forth and that does tie 9 into an open item that the staff has related to the 10 general offsite release calculation. Isn't that 11 right?
12 PARTICIPANT: Is it?
13 MS. BRISTOL: Just probably the --
14 MR. GALYEAN: I'm sorry. Can you repeat 15 the question?
16 DR. SCHULTZ: The general evaluation of 17 offsite releases and site boundary calculations or 18 determination has to do with the source term -- the 19 general source term documentation that is under review 20 by the staff.
21 MS. BRISTOL: That's correct.
22 DR. SCHULTZ: And that piece is an open 23 item that the staff has tied to 19.2.
24 MS. BRISTOL: That's correct.
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19 1 get to that later being a different meeting, but it 2 does tie in here. So we need to pay attention to 3 that.
4 MEMBER BLEY: That's set somewhere in the 5 summer.
6 DR. SCHULTZ: I'm sorry?
7 MEMBER BLEY: That's sometime in the 8 summer we're supposed to --
9 CO-CHAIR CORRADINI: Well I think what 10 Steve said is accurate. Staff has received the 11 topical report from NuScale I think revision three --
12 MEMBER BLEY: Three.
13 CO-CHAIR CORRADINI: -- or four I think.
14 Yes. Three. And staff is still early in the game of 15 reviewing it. We probably shan't see it until fall.
16 MEMBER BLEY: Fall. Okay.
17 MR. RAD: So to be clear, the accident 18 short term topical report is within the scope of the 19 DCA. Emergency planning zone determination is not 20 within the scope of the DCA.
21 MEMBER BLEY: Right.
22 MR. RAD: There is a topical report under 23 review, but that is for a COL application stage.
24 CO-CHAIR CORRADINI: But we're going to --
25 but to answer Dennis' question, we're going to get the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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20 1 chance to look at that topical report once staff has 2 done its review and issued its questions to the 3 applicant. But the estimate is fall.
4 PARTICIPANT: Fall.
5 CO-CHAIR CORRADINI: For the source term.
6 PARTICIPANT: Yes.
7 CO-CHAIR CORRADINI: And by then, we'll 8 have our full compliment of folks.
9 MR. MULLIN: All right. I think we can 10 move on to the next slide. So to kind of support all 11 of our severe accident evaluations, we use the severe 12 accident code MELCOR.
13 And what MELCOR kind of does for us is it 14 provides the best estimate evaluation of a severe 15 accident progression.
16 And it also kind of gives us an evaluation 17 of the severity of these severe accident challenges to 18 containment.
19 These challenges that I listed off -- and 20 MELCOR can evaluate hydrogen combustion, obviously, 21 over pressurization, in-vessel retention, things like 22 that, at least to some degree.
23 But on top of that, of course, we perform 24 standalone conservative evaluations for all these 25 severe accident challenges to evaluate their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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21 1 credibility or their possibility.
2 And we use MELCOR to kind of support those 3 evaluations. It provides a physical basis for timing 4 of these events.
5 It'll give us an idea of how much 6 relocated material we might expect, what kind of 7 materials will relocate, what pressures are as a 8 function of time, temperatures throughout the system, 9 how much hydrogen is produced is an important one.
10 And these conservative severe accident 11 challenge evaluations all state -- they don't just use 12 kind of our best estimate result from one MELCOR run.
13 We use kind of limiting values from 14 database simulations and these simulations themselves 15 involve a number of bounding conservative 16 simplifications because we don't want to run at 17 beginning of cycle, middle of cycle, end of cycle.
18 We've always just run everything at end of 19 cycle to get the maximum decay heat load. We tend not 20 to credit DHRS.
21 Even though if we did, these events would 22 be -- well if DHRS weren't to work, the probability of 23 these sequences we're evaluating would be below 10 to 24 the negative 15, 10 to the negative 16.
25 But we assume they don't work and as a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 1 result, their accidents are a lot faster than they 2 really would be.
3 On top of that, our kind of conservative 4 severe action challenge evaluations, we consider 5 parameters to bound even what we see in MELCOR 6 entirely.
7 For example, for in-vessel retention, 8 which we'll discuss shortly, we assume the full core 9 relocates as soon as possible, essentially. We assume 10 the whole core is molten.
11 It's pure UO2. There's no fill-in 12 materials, no steel, no zirconium to dilute the heat 13 flux. We don't credit water in the lower plenum at 14 the time of relocation for in-vessel retention, for 15 example.
16 So going to the next slide, talk about how 17 we developed the model. I think yesterday we talked 18 about NRELAP5. We used that code to evaluate our 19 success criteria for the Level 1.
20 It's a more rigorous thermal-hydraulic 21 code than MELCOR. It's validated to test facilities 22 for the NuScale design.
23 Therefore, part of our goal in MELCOR 24 model development is to match the NRELAP5 25 characteristics up until core damage, at which point NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 1 NRELAP5 can no longer calculate things.
2 So because of that, we kind of mimic the 3 NRELAP5 model. We've matched all the elevations, the 4 volumes, the flow areas, all of these kind of inputs.
5 And then we do a pretty rigorous 6 benchmarking of a number of accident sequences up to 7 the point of core damage to show the excellent 8 agreement for the most part.
9 And then, we move into severe accident 10 space where, of course, we enter a world of 11 uncertainty, but we ensure that we're doing the best 12 we can by modeling appropriate and accurate design 13 characteristics of the NPM and that's -- you get an 14 accurate decay power curve, and all the masses in the 15 core in the right places and locations.
16 We have the radio and inflight inventory 17 correct from origin calculations and we incorporate 18 modeling best practices that have been around for a 19 while.
20 We talked to MELCOR co-development staff.
21 We have a good relationship with Sandia National 22 Laboratories. We work with subject matter experts in 23 the industry.
24 We followed rigorously guidance from 25 state-of-the-art reactor consequence analyses reports.
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24 1 And, of course, we use the guides and manuals 2 available to us.
3 DR. SCHULTZ: Etienne --
4 PARTICIPANT: I'm a --
5 DR. SCHULTZ: -- does this mean that your 6 modeling, your nodalization that you've developed for 7 your MELCOR model is the same as it is for NRELAP?
8 MR. MULLIN: I mean, we might have some 9 nodes that are bigger. For example, I mean, MELCOR 10 and RELAP function quite differently in terms of how 11 they handle thermal-hydraulics such that you want 12 different node sizes, essentially.
13 You can't just copy them one for one.
14 Mainly, our MELCOR would run really slowly if we did 15 that.
16 DR. SCHULTZ: And the thermal-hydraulic 17 performance -- as you said that you got excellent 18 agreement between your previous evaluations using 19 NRELAP and MELCOR --
20 MR. MULLIN: Yes.
21 DR. SCHULTZ: -- convincing yourself that 22 the thermal-hydraulics, which are different in this 23 design than for what MELCOR has been applied to 24 previously. That the modeling is satisfactory for 25 MELCOR, particularly.
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25 1 MR. MULLIN: Well, I mean, from my point 2 of view, the NRELAP5 code for the new CL design, the 3 NRELAP5 model we have built for it, that came from our 4 safety analysis group that does Chapter 15 analysis, 5 NQA-1 safety related.
6 It's validated to, you know, we have a 7 test facility it's validated to. You know, I take 8 that as the best I can get as an indication of how our 9 plant's going to perform.
10 So if I can match that -- those 11 simulations, I think that's the best I can do.
12 DR. SCHULTZ: Okay. And in matching that 13 with MELCOR, your nodalization is the same as or 14 similar to -- the models that you've developed for 15 MELCOR are similar to it, what you've done, what has 16 been done for -- under the QA program for --
17 MR. MULLIN: Yes. For the most part.
18 RELAP?
19 DR. SCHULTZ: -- for the most part?
20 MR. MULLIN: I might combine two RELAP 21 nodes into one MELCOR node, for example.
22 DR. SCHULTZ: Are there sensitivity --
23 nodalization sensitivity studies that you've done?
24 The staff has done one we're going to talk about.
25 That is to say --
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26 1 MR. MULLIN: Okay.
2 DR. SCHULTZ: -- they've got a different 3 nodalization with the MELCOR work that they've done.
4 We'll talk about that later, but have you done any 5 nodalization studies, given that you're working with 6 a different sized machine than what MELCOR has been 7 applied to before?
8 MR. MULLIN: I mean, yes. We've done that 9 over time. I mean, I've built this model in a number 10 of iterations and I've definitely played with 11 nodalization size in various areas. But that's not 12 necessarily been a formalized process.
13 DR. SCHULTZ: Okay.
14 MR. BLEY: I'm a little concerned on two 15 points. And I haven't thought of a way out of the 16 things I'm concerned about.
17 One is given the set of assumptions you 18 made about what you don't consider, this is not in the 19 spirit of a PRA model, except you'd want to look at 20 cases where those things actually were failed and 21 couldn't work.
22 The reason -- I'm guessing. I don't know.
23 The reason -- if you included those, the results are 24 so low in frequency have to do a lot with the low 25 frequencies we calculate out of the Level 1 PRA of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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27 1 getting into situations where you'd have this problem.
2 I wonder, don't have a clue, if you did 3 your calculations just conditional on getting into a 4 core damage situation or a challenge situation, your 5 results would look lower or kind of similar to 6 existing reactors.
7 That's one I worry about. The thing about 8 all these conservatisms in a PRA, you almost want to 9 have a stamp on results saying that.
10 Because if somebody finds something that 11 goes the other way, they're going to pretend like 12 everything's modeled perfectly and now one thing's 13 worse in calculating high numbers.
14 So that worries me. The other thing is 15 close to what Steve was talking about and I don't know 16 MELCOR at all.
17 I've never worked with it. I kind of know 18 what sort of things it's supposed to do. If there's 19 any reason why the assumptions built into MELCOR are 20 not appropriate to this machine, it could change your 21 results and I'm -- have you really studied that?
22 Is there anything built into the 23 assumptions that are inside of MELCOR that you might 24 not meet?
25 MR. GALYEAN: Dennis, by the way Bill --
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28 1 this is Bill Galyean speaking.
2 MEMBER BLEY: For the front row, you don't 3 have to announce yourself, because your name tag is --
4 MR. GALYEAN: I want to clarify something 5 you said early on about things we didn't consider. We 6 considered everything pretty much.
7 I mean, we didn't just dismiss those 8 severe accident challenges that Etienne talked about 9 at the beginning. We analyzed it.
10 MEMBER BLEY: Uh-huh.
11 MR. GALYEAN: Okay? And we found that 12 they were simply not credible failure mechanisms for 13 our design. Okay? But we do have analyses to support 14 those conclusions.
15 Also, with regard to, you know, MELCOR, we 16 worked very closely with Sandia National Laboratories.
17 MEMBER BLEY: Uh-huh.
18 MR. GALYEAN: Okay? To modify MELCOR, to 19 accommodate the NuScale design. We paid -- we ended 20 up paying Sandia over a million dollars, okay, for 21 them to modify MELCOR to accommodate our design.
22 And, again, we worked closely with them.
23 They added some models --
24 MEMBER BLEY: Yes.
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29 1 the second lower head. They added some models to 2 accommodate our decay heat removal system better.
3 So there has been a -- there's been a lot 4 of time, effort and money spent, okay, to ensure that 5 MELCOR is, in fact, a high fidelity representation of 6 the NuScale design.
7 MEMBER BLEY: Yes. Etienne hinted at it 8 or just barely mentioned it. How much benchmarking 9 against your experiments was done?
10 MR. GALYEAN: You mean the test facility?
11 MEMBER BLEY: Yes.
12 MR. GALYEAN: Well, again, the 13 benchmarking was against the NRELAP5 model and 14 results. Okay?
15 So as Etienne said, NRELAP5 is an NQA-1 16 code developed by our safety analysis folks. And it 17 is -- it has been benchmarked against the test 18 facility, okay, very closely and --
19 (Simultaneous speaking.)
20 MEMBER BLEY: The one code's --
21 MR. GALYEAN: That's right.
22 MEMBER BLEY: -- benchmarked against tests 23 and the other one matches. Yes.
24 MR. GALYEAN: That's exactly right.
25 MEMBER BLEY: Okay.
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30 1 MR. GALYEAN: And MELCOR was benchmark 2 against the NRELAP5 code.
3 MEMBER BLEY: Thank you.
4 DR. SCHULTZ: Of the MELCOR experts that 5 you worked with in the development and application to 6 the NuScale design, have they been involved in looking 7 at your current results?
8 The results that you've generated in the 9 MELCOR analyses? Have they still -- have they been 10 involved throughout the process to the end point here?
11 MR. GALYEAN: I mean, we've hired 12 contractors, consultants that have worked with us for 13 quite a while and have been with us as we develop 14 these results.
15 DR. SCHULTZ: Okay.
16 MR. GALYEAN: Yes.
17 DR. SCHULTZ: Doing the work. So it's 18 continued so that the results that you have achieved 19 with the models that Sandia worked on, have gotten 20 some review?
21 The way you described it, this is not a QA 22 program that -- this is not under the QA program. So 23 you'd want to have expertise that are -- have done the 24 development also reviewing the results that have been 25 achieved with the models?
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31 1 MR. GALYEAN: Yes. And I'm sure we'll 2 hear about it later. The NRC has performed an 3 independent assessment.
4 DR. SCHULTZ: Okay. As they should.
5 MR. GALYEAN: As they should.
6 DR. SCHULTZ: Thank you. We'll get into 7 that, too. Thank you.
8 MR. MULLIN: So let's move on and talk 9 about in-vessel retention and I expect we'll talk 10 about this a fair amount today. This is kind of key 11 to our severe accident management strategy, I guess.
12 So for the purpose of the rest of this 13 discussion on severe accidents, we're really only 14 going to talk about events where the containment 15 successfully isolated.
16 We're not talking about bypass events. So 17 for an accident -- for a severe accident, a core 18 damage event in which you successfully isolate the 19 containment, all of the water that leave your RPV, 20 which is necessary to achieve core damage, is going to 21 be in the containment, submerging the RPV.
22 And we performed conservative analysis to 23 demonstrate that upon a core damage event and a 24 relocation of that core into the lower plenum of the 25 RPV, it will remain there.
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32 1 And this is based off essentially an 2 evaluation of the heat flux imposed, the maximum heat 3 flux that can be imposed by the core debris on the 4 lower head compared to the maximum or the critical 5 heat flux or maximum amount of heat that you can 6 remove without -- via boiling on the outside of the 7 lower head.
8 And essentially, we maximize the heat 9 generation rate. We used a bunch of conservative 10 assumptions. We have the full core. The quickest 11 possible relocation times. We have the highest decay 12 heat, end of cycle.
13 We don't create any upward radiation heat 14 losses. We kind of assumed the worst possible 15 convection situation inside there --
16 MEMBER BLEY: You're looking at --
17 MR. MULLIN: -- the core debris.
18 MEMBER BLEY: -- heat transfer and you're 19 looking at overheating the bottom head. What about 20 when you overheat that whole lower area and it 21 expands?
22 You look at the stresses on the vessel 23 from that consideration as well?
24 MR. MULLIN: The expansion of the lower 25 head --
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33 1 MEMBER BLEY: Yes.
2 MR. MULLIN: -- and the strains on that?
3 We did do a --
4 MEMBER BLEY: Connection to the cylinder, 5 you know, blowing up.
6 MR. MULLIN: We did a structural 7 evaluation as well.
8 MEMBER BLEY: You did?
9 MR. MULLIN: Yes. But we focus mostly on 10 this critical heat flux evaluation, which has been 11 done in the past, especially with AP1000 evaluation of 12 in-vessel retention.
13 MEMBER BLEY: Just --
14 MR. MULLIN: Go ahead.
15 MEMBER BLEY: I assume all of those 16 calculations are in engineering reports --
17 (Simultaneous speaking.)
18 MR. MULLIN: Absolutely. Yes.
19 MEMBER BLEY: -- you're don't see.
20 MR. MULLIN: So we've kind of maximized 21 the amount of heat that you can put into this lower 22 head from the core.
23 And then, we applied a critical heat flux 24 criterion based on experiments that have been done in 25 the past on hemispheres that do not credit the kind of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 the high pressure that you're going to have in the 2 containment and the large subcooling that you're going 3 to have in the containment.
4 So this is one thing that's pretty unique 5 about the NuScale design is that when you have a 6 severe accident, you oxidize all your cladding and a 7 lot of steel.
8 You're going to generate a bunch of 9 hydrogen. It's going to go into the containment and 10 you're going to get up to one or two -- 250 psi in the 11 containment during severe accident.
12 That's well below our maximum design 13 pressure of 1,000. I don't know if that's 14 proprietary. And as a result, you're going to get a 15 lot of subcooling in the pool in the NCV.
16 You're going to have about 100 degrees 17 Kelvin subcooling. So you can actually remove a lot 18 of heat from the lower head without even boiling 19 water.
20 And when you do start boiling water, you 21 can really boil a lot before you reach the critical 22 heat flux.
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35 1 with substantial margin.
2 And, therefore, we concluded that failure 3 of the lower head is unlikely. And, therefore, the 4 accident is essentially stabilized in the RPV.
5 CO-CHAIR SUNSERI: Unlikely from the 6 standpoint of the analysis is probability zero?
7 MR. MULLIN: We said it does not occur.
8 Yes.
9 CO-CHAIR SUNSERI: And these assumptions 10 about the heat that you're talking about, I mean, when 11 you consider that in light of industry standards, this 12 is an extremely conservative approach. Is that a fair 13 statement?
14 MR. MULLIN: Yes. Yes. And I'd like to 15 kind of talk about why we thought we could do that.
16 You know, our core is a lot smaller than a traditional 17 core.
18 It's, you know, five percent the size and 19 so that takes advantage of, you know, this cubed 20 square flaw. So for the amount of heat generation you 21 have for a smaller core, you're going to have more 22 surface area.
23 Furthermore, our power density is less 24 than half of a traditional core and then, we have 25 about five or six times as much primary coolant per NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 unit power.
2 So it slows down our accidents and we have 3 all this hydrogen in containment that really 4 pressurizes that water and subcools it.
5 And for all these reasons, we have, you 6 know, it's our opinion that we have quite a lot of 7 margin to CHF and we can conclude that it doesn't 8 occur.
9 So given that we've stabilized the 10 accident in the RPV, the remaining severe accident 11 challenges, by our analyses, by our evaluation, do not 12 challenge the containment.
13 Mainly, hydrogen combustion really cannot 14 occur, because of limited oxygen that we've evacuated 15 the containment prior to this even starting to occur.
16 The containment is just going to be pure 17 hydrogen. So it's hydrogen inerted. Fuel core 18 interactions or steam explosions in the vessel are 19 unlikely.
20 And if they do occur, are not really 21 energetic enough to induce this supposed alpha mode 22 failure of the containment.
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37 1 penetrates it.
2 It's already kind of been dismissed as 3 resolved from the risk perspective by the industry in 4 the '90s, but it still kind of lingers in discussions.
5 The reasons why it's even less of an issue 6 for our design is we have a small core, lower debris 7 temperatures.
8 We have a small drop height. So you can't 9 premix as well. It's a shallow pool and relatively 10 high system pressure, which seems to stabilize the 11 steam explosion phenomena.
12 Furthermore, the big risk in conventional 13 containments is containment over pressure, gradual 14 over pressure. That just can't occur for our design.
15 We have a high-pressure steel CNV.
16 It's designed for a very limiting LOCA 17 blowdown, which greatly exceeds the maximum pressures 18 we'll see in a severe accident. The CNV is submerged 19 at all times.
20 In the ultimate heat sink, it provides 21 effective pressure suppression and there's no concrete 22 in the containment to produce more non-condensable 23 gases from a core concrete interaction.
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38 1 accident, if the RPV were to fail, if we were to melt 2 through the RPV and corium were to come into the pool 3 in the containment vessel, these various ex-vessel 4 phenomena that are evaluated in conventional designs, 5 are really not credible for our design either.
6 High pressure melt ejection, which 7 requires a large pressure differential between the 8 reactor vessel and the containment vessel can't occur 9 because we don't have that pressure differential.
10 Even if the break in the containment is 11 very small and restrictive, just having the water 12 touching the RPV will cool you down and you'll 13 equalize pressures very quickly.
14 Energetic ex-vessel steam explosions are 15 not likely for similar reasons as in-vessel FCI.
16 MEMBER BLEY: Can you explain that one a 17 bit? I'm just remembering Hans Fauske's experiments 18 and seeing him poor stuff together and blow up. So why 19 it is the ex-vessel steam explosion can occur?
20 MR. MULLIN: Well it's a difficult problem 21 not to occur, really in any situation. Maybe assume 22 a hole in the RPV or, you know, corium starting to 23 come out.
24 It doesn't have enough distance to drop, 25 to pre-mix into little droplets and then trigger and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 1 then a shockwave comes up as the steam explosion.
2 MEMBER BLEY: So it's kind of more of a 3 surface --
4 MR. MULLIN: It kind of --
5 MEMBER BLEY: -- heat transfer phenomena 6 in that physical explosive?
7 MR. MULLIN: Well it's, I mean, a steam 8 explosion occurs as you introduce molten -- very high 9 temperature of molten material into -- you pour it 10 kind of violently into a pool. It hits the pool.
11 It distributes into little droplets and 12 then, you need a triggering event, which has been hard 13 to identify in a lot of experiments and they usually 14 apply a triggering event that applies a little 15 shockwave.
16 And then, it collapses all the droplets, 17 which have a little vapor blanket around them. It 18 collapses the vapor blanket.
19 The droplets solidify and they fragment 20 and they transfer a lot of heat very rapidly to the 21 pool.
22 And that's a propagating shockwave that 23 sends a slug of water up or just provides an impulse 24 on the sides of the vessel. And we don't really have 25 the geometry for that.
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40 1 MEMBER BLEY: Okay. Have to think about 2 that a little more, but the size of the pool, yes, you 3 won't get that concentrated slug --
4 MR. MULLIN: Yes.
5 MEMBER BLEY: -- you see in the other 6 case. Bill looks like he wants to say something.
7 MR. GALYEAN: Yes. Again, we're talking 8 -- Etienne talks about dropping it into the pool, into 9 the containment. Right?
10 I mean, and the volume in the containment 11 that is between the bottom of the reactor pressure 12 vessel lower head and the bottom of the containment 13 lower head, is very limited.
14 Okay? And so there's just not a lot of 15 volume there to accommodate or to allow this, you 16 know, rapid energy transfer that Etienne's talking 17 about, that would spawn an energetic FCI.
18 MEMBER BROWN: What FCI?
19 MR. MULLIN: Fuel coolant interaction or 20 steam explosion. So, you know, further, if you 21 relocate all your core debris into the containment 22 lower head, mind you, you've got water on top of it 23 now and it's poured into water and it's probably kind 24 of plugged up the hold a little bit.
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41 1 debris in the CNV lower head. For the same reasons 2 that we don't think you melt through the RPV lower 3 head, you won't melt through the CNV lower head.
4 However, even if that were to fail, you 5 would need to -- the release would need to pass 6 through 60 feet of water and, therefore, it's a small 7 release.
8 As we've -- to demonstrate our -- as we 9 discussed in the context of drop module, even if you 10 have a release at the bottom of the pool, it cannot be 11 a large release because of the pool scrubbing effect.
12 And further, even if this upper CNV were 13 to fail from a steam explosion type event --
14 MEMBER BROWN: Yes. If it goes through the 15 containment vessel, well that just disburses it to the 16 entire pool. Doesn't it?
17 MR. MULLIN: Yes.
18 MEMBER BROWN: So, I mean, you effectively 19 contaminate 12 modules if that's what you've done.
20 Don't --
21 MR. MULLIN: You contaminate the pool, 22 certainly.
23 MEMBER BROWN: Okay. And there's no 24 chance that based on the volume of water that you 25 could have a severe steam energetic reaction where you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 1 could cross it up through the top or anything like 2 that?
3 I mean, I'm not a thermal-hydraulic guy.
4 I'm just a plain old engineer. So just --
5 (Laughter.)
6 MR. MULLIN: No. I don't --
7 MEMBER BROWN: There's a heck of a lot of 8 energy in that hot stuff coming out through whatever 9 is does and initially it's fairly confined as it goes 10 through from the reactor to the pressure vessel down 11 into the containment.
12 Then, it leaks through that. A little bit 13 of it then all of a sudden hits it and it just goes.
14 Just seems like that very high temperature fuel and 15 has anybody modeled that to see what that reaction 16 would be?
17 MR. GALYEAN: I understand that we are 18 being, I want to say wildly speculative here. In the 19 vast majority of core damage accident scenarios that 20 we have modeled using MELCOR and, I mean, the vast 21 majorities, we do not even see fuel temperatures 22 reaching melting temperatures.
23 Okay? We see core relocation by virtue of 24 failure of the support structures, but we almost never 25 see temperatures reaching melting temperatures --
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43 1 (Simultaneous speaking.)
2 MEMBER BROWN: Where you could melt the 3 RPV itself?
4 MR. GALYEAN: Exactly. Exactly. I mean, 5 we have seen in very limited extreme sensitivity 6 studies, okay, if we tailor the assumptions and the 7 input parameters, yes.
8 We can achieve melting temperatures of the 9 fuel, but these are very few, very far between, very 10 extreme conditions and so for the sequences and 11 scenarios that we model in the PRA, we simply don't 12 see melting temperatures occurring.
13 So we're, like I said, we're straying in 14 very extreme speculative space here. So --
15 MEMBER BROWN: Extremes do happen.
16 DR. SCHULTZ: So, Bill, that's another way 17 to consider uncertainty. In other words, you -- what 18 you're indicating is that this is a qualitative way to 19 drop down through all the possibilities.
20 If this, if this, if this, then -- that's 21 one way to handle uncertainties. The other way as you 22 described is you've done substantial number of 23 evaluations, uncertainty evaluations where you've 24 pushed the limit, if you will, pushed toward the limit 25 and found acceptable results.
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44 1 MEMBER BALLINGER: So what's the distance 2 between the bottom of the pressure vessel and the 3 containment?
4 MR. MULLIN: Something, like, less than a 5 foot.
6 MEMBER BALLINGER: Less than a foot? Is 7 there a way, again, being wildly speculative, for 8 creep deformation to close that gap without failing 9 the vessel?
10 MR. MULLIN: I don't -- I can't answer 11 that. I mean, I would expect if we were actually 12 going to fail the RPV lower head, it would probably 13 swell to maybe come down and meet the containment 14 lower head. I don't really know.
15 MEMBER BLEY: Well, apparently, he said 16 there's a structural report and I don't know if the 17 staff reviewed that an article last time I got it.
18 MR. MULLIN: It was part of the evaluation 19 of in-vessel retention. So as I was saying, even if 20 the upper CNV were to fail, from a steam explosion 21 type event, at that time and somewhat later in the 22 accident sequence, your radionuclides have been 23 released and have deposited largely.
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45 1 release by our calculations.
2 At least the instantaneous release of 3 everything airborne in there would not constitute a 4 large release.
5 So that's kind of it. Our Level 2 6 insights are this, the core damage events, they're 7 stabilized within the RPV, by our analyses.
8 And the remaining severe accident 9 phenomena do not challenge the CNV integrity. However, 10 even if we were to fail the RPV and the CNV, the 11 result is not going to be a large release, which 12 confirms our Level 2 event trees.
13 Thereby, the large release frequency is 14 dominated by containment bypass events. We have a 15 list of kind of key insights here.
16 There's a number of them. I'm not going 17 to read everything here, but the first one is 18 containment isolation.
19 As we've discussed before, these 20 containment isolation valves are very important.
21 Containment isolation as a function is very important 22 from a Level 2 standpoint.
23 As demonstrated in our event trees, if we 24 successfully isolate containment, we don't have a 25 large release.
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46 1 Furthermore, if we successfully isolate 2 containment, we generally -- we often avoid core 3 damage events altogether by cutting off the flow of 4 the release -- or of the -- with the LOCA outside 5 containment.
6 You can go to the next slide. We have 7 very good heat removal from the RPV, passive heat 8 removal. We have no insulating material from the RPV 9 because we evacuate the containment.
10 As a result, for all of our events, which 11 had successful containment isolation, we have water in 12 the containment and we completely submerge the RPV, 13 which helps stabilize all of our severe accidents in 14 the RPV.
15 To talk about our kind of severe accident 16 challenges specifically, we can't over-pressurize the 17 containment.
18 Even if we were to assume a RPV over-19 pressurization event where we fail the DHRS, we fail 20 our RSVs, get up to this ridiculously high RPV 21 pressure and fail it into containment, we've analyzed 22 that.
23 And even that won't result in a failure of 24 the containment once you equalize the pressures.
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47 1 enough oxygen in the containment.
2 In-vessel stream explosions do not occur 3 or do not challenge the containment. The core support 4 plate fails pretty early and for all these reasons 5 we've talked about before.
6 DR. SCHULTZ: Before you leave this slide, 7 Etienne, the -- one of the staff open items in 19.2 is 8 associated with hydrogen combustion external to the --
9 MR. MULLIN: Yes.
10 DR. SCHULTZ: -- to the vessel. You don't 11 mention that here, the staff refers to work that's 12 being -- going to be presented in Chapter 6.
13 Is that -- I mean, it's still -- that's a 14 piece that's still under review and it's open.
15 MS. BRISTOL: Yes. We submitted an RAI on 16 that topic and they have an additional supplemental 17 response and reviewing to confirm the results we 18 previously submitted.
19 CO-CHAIR CORRADINI: But just from a 20 scheduling standpoint, we're not going to hear about 21 that for this letter. Nor are we going to hear about 22 it in June. It's still being evaluated by the staff.
23 MS. BRISTOL: That's our understanding.
24 CO-CHAIR CORRADINI: Okay. So that's 25 going to remain an open item coming through --
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48 1 DR. SCHULTZ: Beyond June. Correct.
2 That's what I wanted.
3 CO-CHAIR CORRADINI: I wanted to make sure 4 you -- yeah.
5 DR. SCHULTZ: Okay. That's fine. So both 6 the open items are to be discussed later after now and 7 after June? When I say both, I meant the topical 8 report --
9 MS. BRISTOL: That's correct.
10 DR. SCHULTZ: -- coming in November. So 11 --
12 MS. BRISTOL: Yes.
13 DR. SCHULTZ: -- that's fine. Thank you.
14 CO-CHAIR CORRADINI: Where is the 15 bioshield redesign going to appear in second Chapter 16 3? When we do see the redesign, it'll be in Chapter 17 3? I'm sorry. I don't remember if it's three or six.
18 I thought three.
19 DR. SCHULTZ: The staff says six, but --
20 CO-CHAIR CORRADINI: Says six.
21 PARTICIPANT: I don't know.
22 MR. RAD: So I think it affects a couple 23 of areas, actually. So the bioshield redesign will 24 affect discussion in the structural area and I think 25 it also affects the discussion in high energy line NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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49 1 break.
2 PARTICIPANT: Okay.
3 MR. RAD: But the reality is it'll touch 4 most of the areas where the design of the bioshield is 5 discussed, which are several chapters.
6 MEMBER SKILLMAN: Etienne, on slide one of 7 five, in the event that the core is not covered, what 8 consideration is given to pyrophoricity?
9 MR. MULLIN: To what? Sorry.
10 MEMBER SKILLMAN: Pyrophoricity.
11 MR. MULLIN: Could you explain that?
12 MEMBER SKILLMAN: Sure. Core lights 13 itself on fire because of the chemistry of the core 14 materials whether eutectic or whatever, there's 15 sufficient chemical reaction and availability of 16 oxidizing material that the core sets itself on fire.
17 MR. MULLIN: Well, certainly, oxidation 18 reactions are modeled in our MELCOR simulations. I 19 don't know if that looks like fire, but it releases a 20 lot of heat.
21 MEMBER REMPE: Am I allowed to make 22 educated just information to help Dick? I know you're 23 coming from the TMI concerns when you mention that, 24 but usually pyrophoric reactions require very small 25 fragmented material. Right?
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50 1 MEMBER SKILLMAN: I don't know --
2 MEMBER REMPE: Very small pieces --
3 MEMBER SKILLMAN: Well just a minute.
4 Right? I don't know. Right or not right. Accurate?
5 We chased this for a good reason.
6 MEMBER REMPE: The TMI. Yes.
7 MEMBER SKILLMAN: I'm not trying to be 8 snarly here. Because the design is so different, the 9 core so small, most of the assumptions that you've 10 made regarding core size, DKE, maximizing heat 11 transfer, that type of thing, at least in my mind, 12 makes sense.
13 So I'm saying what else is out there that 14 might not have been considered and pyrophoricity is 15 something that we gave quite a bit of attention to 16 because of the potential of the added heat that that 17 reaction could contribute.
18 Particularly when we're headed towards 19 defueling. Dr. Rempe's accurate. It's normally a 20 fine particle interaction where there's moisture, 21 hydrogen and oxygen.
22 I understand that. But I'm just saying 23 this treatment is so thorough. Is there anything 24 that's not been considered that perhaps is a 25 vulnerability or an uncertainty that might surprise NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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51 1 all of us? And pyrophoricity is the one that comes to 2 mind.
3 MR. MULLIN: Well we have not looked into 4 that but we will. It's pretty interesting.
5 MEMBER SKILLMAN: Thank you.
6 MR. MULLIN: Thanks.
7 PARTICIPANT: Four of five, please.
8 PARTICIPANT: Okay.
9 MR. MULLIN: So here we're just kind of 10 talking about some of the other things we've talked 11 about. We skipped the -- let's go to the next one.
12 I know Sarah yesterday brought up that we 13 have one human action that is risk significant and 14 that shows up in the Level 2.
15 This is for an event where you have an 16 outside containment break CVCS line break outside 17 containment, you fail to isolate that line.
18 We can prevent core damage by actuating 19 ECCS and supplying coolant via the CFDS. So the 20 operator action to initiate containment flooding, 21 drain system, emergency coolant injection is -- it 22 comes up as risk significant. Yes.
23 DR. SCHULTZ: Can you help me understand 24 the last sentence under comment? What is the 25 mathematical limitation that is being described here?
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52 1 MS. BRISTOL: Yes. Mathematical. I think 2 that was just with respect to Fussell-Vesely being 20 3 percent of the total large release frequency, not 4 necessarily a limitation.
5 We provided candidates to the D-RAP panel 6 and so, you know, they could determine if they could 7 -- felt they were risk significant or not.
8 DR. SCHULTZ: Okay.
9 MS. BRISTOL: But it was just more, you 10 know, we didn't have human actions we mentioned 11 yesterday for internal events, Level 1, but for Level 12 2 in order of magnitude a Fussell-Vesely of 20 percent 13 threshold --
14 DR. SCHULTZ: Okay.
15 MS. BRISTOL: -- deemed this operator 16 action candidate for consideration.
17 DR. SCHULTZ: All right.
18 MEMBER BROWN: In other words, this is a 19 mathematical construct, not a physical construct that 20 you're basing operator action -- satisfactory 21 operation on just because that's the way you've 22 modeled the stuff as opposed to the physical actions?
23 MS. BRISTOL: Right. Twenty percent of a 24 small large release number.
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53 1 percent or five percent?
2 MS. BRISTOL: As was reviewed in the 3 topical report a couple years ago, we provided --
4 MEMBER BROWN: I didn't read that.
5 (Laughter.)
6 CO-CHAIR CORRADINI: You were here. It 7 was December of '16. We wrote a letter report on the 8 risk significant --
9 MEMBER BROWN: Okay. All right. I still 10 don't know whether it makes a difference or not. I'd 11 much rather see a physical basis for the operator 12 action being satisfactory based on mathematical 13 construct. That's my opinion.
14 CO-CHAIR SUNSERI: But in a sense what 15 you're going though, you use the PRA to figure out 16 what the critical steps are and then, the critical 17 steps then have some significance approach it.
18 And this is one of them and now they know 19 to focus their operator training program. They need 20 to time reactor operator actions, whatever it is to 21 make sure that they do this piece right going forward.
22 Right?
23 MR. MULLIN: I mean, so that's pretty much 24 it. Here's a list of our COL items, pretty generic.
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54 1 slide?
2 MS. BRISTOL: Yes. Overall, they just 3 confirmed that as the design progresses for the 4 combined license phase, we'll continue to look at the 5 PRA and these various inputs into the severe accidents 6 as well as the events we talked about yesterday.
7 And we'll confirm those for site-specific 8 differences that might arise from a specific site 9 versus the design stage.
10 MEMBER BLEY: All right. There have been 11 several different approaches proposed in other design 12 certs.
13 How do you lose the PRA from the designs 14 stage over to the COL stage? You probably say 15 something about that in 19 that I skipped over.
16 Can you tell me what's going to be 17 required when you actually have a site for the PRA?
18 Will you update the PRA based on site-specific 19 information or do you have some alternative set of 20 criteria that you'll have to meet?
21 MS. BRISTOL: We -- similarly to DC phase, 22 Reg Guide 1.206 discusses the expectations for PRA 23 application for DC as well as COL and so we'll follow 24 those similar steps and provide a PRA for the site-25 specific location.
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55 1 MEMBER BLEY: You will revise the PRA for 2 site-specific location?
3 MS. BRISTOL: That's correct.
4 MEMBER BLEY: At the COL stage? That's 5 what I was really asking.
6 MS. BRISTOL: Correct.
7 MEMBER BLEY: That's fine.
8 MS. BRISTOL: And similar to the COL items 9 yesterday, we talked, I know Member Rempe talked about 10 the key assumptions. So that's one example where 11 we're -- we'll verify those.
12 And to clarify that key assumption, there 13 are the tables in the FSAR. For instance, Level 2 was 14 191-21 or Level 1 was. Level 2's 28. All of those 15 assumptions will go through at the COL phase.
16 See if they are any different for a 17 specific versus what we assumed for DCA. We'd have 18 different information for external events or seismic 19 or that that we could incorporate at that time.
20 If there's testing, if there's additional 21 design information that we didn't have for the DC 22 application that we could apply for the COL 23 application, we would do that.
24 MEMBER BLEY: Thanks, Sarah.
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2 MEMBER REMPE: So I appreciate you 3 clarifying what you meant by key assumptions. And I 4 was thinking about this.
5 And so I think I'm allowed to bring this 6 up, but there's a table 19.1-something or other that 7 refers to a lot of assumptions resulting from severe 8 accident phenomena.
9 And if you actually go through and you 10 have the -- I'm always interested in how the COL 11 applicant will move forward and do they have to 12 justify that many of the assumptions related to severe 13 accident phenomena are indeed as conservative as 14 characterized?
15 And, again, this is a 19.1 table, not a 16 19.2 table. So I think I'm legitimate in this because 17 I'm just kind of wondering how -- I'm thinking about 18 ITAAC versus key assumptions.
19 And some of those key assumptions may be 20 difficult for a COL applicant to say -- they can 21 easily say I have no additional information, because 22 if you don't look, you won't find it.
23 But I'm wondering if it might be hard to 24 justify you've met that key assumption on some of 25 these cases where -- I mean, as Dennis mentioned NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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57 1 earlier, conservative sometimes is in the eye of the 2 beholder and when you're looking at when a release 3 will occur.
4 And I just am curious about that and how 5 that will progress.
6 MS. BRISTOL: We're unsure of the actual 7 table, so if can get that table, we can kind of 8 confirm --
9 (Simultaneous speaking.)
10 MEMBER REMPE: Okay. I don't have it in 11 front of me now, but I will find it for you.
12 CO-CHAIR DIMITRIJEVIC: That would be 13 important.
14 MEMBER REMPE: But it's a table --
15 PARTICIPANT: And you just sent for it.
16 MS. BRISTOL: Yes. But it is a table and 17 I'll find it. I had it up earlier. But I had --
18 MS. BRISTOL: We'll look and get back to 19 you.
20 MEMBER REMPE: Yes.
21 MS. BRISTOL: Thank you.
22 CO-CHAIR SUNSERI: Are you all done with 23 the -- are you done?
24 DR. SCHULTZ: Well I just have a -- I'm 25 just -- given the comments by Dennis and by Joy, I'm NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 1 trying to understand what a COL applicant would do 2 under Item 19.2-2?
3 They've got site-specific information and 4 that's going to cause them to define improvements in 5 the reliability of core and containment heat removal 6 systems.
7 Can you describe what that -- those 8 conclusions might be? What improvements they might do 9 at a COL stage?
10 MEMBER BROWN: Does that mean they can 11 change the design?
12 MS. BRISTOL: No.
13 MR. RAD: So a COL applicant can 14 absolutely depart from the certified design. That's 15 up to them. Right?
16 And so if they depart from that design or 17 if they have site-specific information that is outside 18 the assumptions in the PRA as provided in the 19 certified design, then they'll need to provide 20 information in accordance with the COL item.
21 DR. SCHULTZ: So this is a backstop is 22 what it is?
23 PARTICIPANT: Exactly.
24 DR. SCHULTZ: Yes.
25 MR. RAD: Absolutely.
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59 1 DR. SCHULTZ: Well it's a handoff of the 2 responsibility of maintaining the PRA from the --
3 MR. RAD: That's for --
4 (Simultaneous Speaking.)
5 DR. SCHULTZ: -- certified design to the 6 operational plan.
7 MR. RAD: That's correct. That's correct.
8 There are substantial departures that could be taken 9 in theory from the time the design is certified to the 10 time the design is licensed at the application stage.
11 Both would, of course, involve the NSSS 12 vendor, but that's up to the licensee to make those 13 determinations. Not up to NuScale today.
14 MEMBER REMPE: And the table of interest 15 is Table 19.1-29, Generic Sources of Level 2 Model 16 Uncertainty.
17 MS. BRISTOL: Thank you. I'll take a 18 look.
19 CO-CHAIR SUNSERI: All right. Any members 20 have any other questions for the NuScale team? All 21 right. Well we thank you for your presentation then.
22 We'll exchange this -- let's see. Yes.
23 We'll go ahead and exchange seats now with the staff 24 and bring up the staff for the 19.2 safety evaluation.
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60 1 ahead and get started.
2 MR. CRANSTON: Good morning --
3 CO-CHAIR SUNSERI: All right.
4 MR. CRANSTON: -- my name is --
5 CO-CHAIR SUNSERI: All right. Let's --
6 we're back in -- go ahead.
7 MR. CRANSTON: Good morning. My name is 8 Greg Cranston. I'm the senior project manager for the 9 NuScale project and we're continuing today with 10 Chapter 19, as you're aware, with 19.2, .3 and .5.
11 So at this time, I'd like to just turn 12 over to the staff for presentation. Jason?
13 MR. SCHAPEROW: Thank you, Greg. My name 14 is Jason Schaperow. I'm a member of the NRC technical 15 staff and I will present the staff's review of severe 16 accident mitigation for the NuScale design. Slide 4, 17 please.
18 MEMBER BLEY: Jason, do you have any place 19 in the -- I haven't looked at your slides yet, any 20 place in the slides where you going to talk about the 21 audit?
22 MR. SCHAPEROW: Yes. Actually, the very 23 first slide --
24 MEMBER BLEY: Okay. Good.
25 MR. SCHAPEROW: -- actually. Slide 4.
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61 1 MEMBER BLEY: I won't ask any questions 2 I'll wait for the slide.
3 MR. SCHAPEROW: Yes. So after hearing the 4 Committee yesterday, I want to try to maybe go a 5 little more detail what we did, certainly in the area 6 of severe accident mitigation.
7 We began our review in 2017. So roughly 8 two years ago or if we read the FSAR. And then we 9 read a series of detailed engineering reports that 10 NuScale provided.
11 NuScale provided these reports via 12 website, which we could access from our offices. So 13 basically, we sat for several weeks and we read these 14 reports.
15 And then we had meetings with NuScale to 16 ask our questions, all as part of this first audit we 17 did in 2017. When we met with NuScale it was -- well 18 we basically went across the street to the NuScale 19 office here in Rockville.
20 And they had this really nice setup where 21 they -- one wall of the room was basically this big 22 screen and maybe some of the ACRS have experienced 23 this. So basically, you're sitting in a room.
24 You're not physically there with them, but 25 there's a big screen and you're -- it looks like NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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62 1 you're at a big long table and they're at the other 2 end of it. Very impressive. So we asked a lot of 3 questions, whatever questions we had.
4 So, by going through this reviewing of the 5 reports and asking the questions, that enabled us to 6 kind of come up with what we thought our essential 7 questions were for the review.
8 And those essential questions we sent to 9 them. The NRC's formal RAI process, which you're 10 probably all familiar with.
11 MEMBER BLEY: Jason, among those reports, 12 did you folks review the one that was -- talked about 13 -- I think Etienne was talking about, on the 14 structural analysis report on the --
15 MR. SCHAPEROW: The structural analysis 16 report?
17 MEMBER BLEY: Yes. He said they're --
18 MR. SCHAPEROW: For in-vessel retention?
19 MEMBER BLEY: Yes.
20 MR. SCHAPEROW: Yes. And I should have --
21 so --
22 MEMBER BLEY: If you're going to talk 23 about that separately, that's great.
24 (Simultaneous speaking.)
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63 1 lot of detail on what they did exactly for in-vessel 2 retention, because we had some fundamental issues with 3 it. And I'll get into that in a slide or two.
4 MEMBER BLEY: Okay.
5 MR. SCHAPEROW: What we actually did do in 6 the area of in-vessel retention.
7 MEMBER BLEY: Okay. Because I had a 8 question and Ron had one, too. So we'll --
9 MR. SCHAPEROW: Yes, vice-chairman.
10 MR. SCHAPEROW: Okay. So we sent them our 11 request for information formally. Roughly, we had one 12 for each issue. For steam explosion, we sent them a 13 request for information.
14 We did one for in-vessel retention. One 15 for high pressure melt ejection. And we got their 16 responses and -- so we read their responses, but it 17 wasn't quite enough for us.
18 So we said we'd like to see more, hear 19 more. So we opened up a second audit. This was in 20 2018 and so we did the same thing.
21 They put the documents in there that they 22 had developed basically since we asked the questions, 23 these new documents or revised documents.
24 We read those. We sat down. We met with 25 them again and, you know, we -- at that point, we had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 1 pretty much had all our questions answered.
2 There may have been one or two REIs that 3 were not completely answered by the end of the audit.
4 I think maybe Dr. Bley noticed that.
5 MEMBER BLEY: Yes.
6 (Simultaneous speaking.)
7 MR. CRANSTON: There were two of them.
8 The one on --
9 MR. SCHAPEROW: The reason -- yes.
10 (Simultaneous speaking.)
11 MR. CRANSTON: -- retention, the one on 12 high pressure --
13 (Simultaneous speaking.)
14 MR. SCHAPEROW: So the reason was the 15 audit got started before we had finished reviewing all 16 the REI responses. So we weren't able to finish going 17 through those as part of the audit.
18 So we actually resolved -- finished 19 discussing those with NuScale as -- because the audit 20 was limited. Like, it was, like, a month or a month 21 and a half.
22 So that ended and then -- so we had a 23 couple of public meetings and closed meetings with 24 them afterwards, more formal, where we discussed their 25 -- some outstanding questions we had on those two NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 1 RAIs.
2 Okay. So another thing that we did as 3 part of our review at severe accident mitigation was 4 we did perform confirmatory analyses. We did MELCOR 5 analyses as we do typically do for each of the design 6 centers.
7 We also did some in-vessel retention 8 analyses and steam explosion analyses as we did for 9 AP1000 because we had some -- we actually had some 10 very similar issues with NuScale's in-vessel retention 11 analysis as we did with AP1000's.
12 And, again, I'll go through those as we go 13 through the slide presentation. Slide 5, please. So, 14 of course, any NRC review is guided by the NRC staff 15 guidance.
16 We have an SRP section that we follow.
17 There really isn't much in there on severe accidents.
18 19.0 is really aimed at the Level 1 PRA and what 19.0 19 does is it points to these two SECY papers from the 20 early '90s that lay out the basis for NRC's review of 21 new reactor designs.
22 This is all in the timeframe of CE80 plus, 23 AP600, ABWR, when we were getting the first new 24 reactor designs in back -- I guess they're not new 25 anymore, back in the early '90s.
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66 1 (Laughter.)
2 MR. SCHAPEROW: Of these the kind of the 3 central thing as far as severe accident mitigation 4 goes, is containment performance. That's what we 5 focus on.
6 And those SECYs and the associated SRMs 7 that lay out the containment performance goals and I 8 will talk about we tried to apply those to NuScale was 9 hard, because NuScale's not exactly like the other 10 designs.
11 Anyway. Slide 6, please. Okay. So we 12 started out we reviewed the NuScale design, as we 13 typically do, to confirm that they meet the 14 containment performance goals for new reactors.
15 And I list the goals here. There's two 16 goals. These goals were put into place because the 17 Commission wanted something for mitigation of severe 18 accidents.
19 The safety goals, the CDF, the LRF, those 20 could all be met on basis of a low CDF. If you have 21 a very good design, you prevent core damage, you can 22 meet all the safety goals.
23 So we just -- NRC decided we needed 24 something else. So to -- for as a backstop you might 25 say or if you did have a core damage accident, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 1 have to show your containment is robust.
2 So there were two goals. There was a 3 probabilistic goal and a deterministic goal. And 4 these have been used again through all the design 5 centers.
6 The probabilistic goal is that if there 7 was a core damage accident, you have to show that the 8 likelihood of containment failure, where bypass is 9 less than .1.
10 So that's our definition of a robust 11 containment from a probabilistic point of view. From 12 a deterministic point of view, you have to show -- the 13 safety analysis report has to show that the 14 containment will stay intact for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
15 So then they had this other provision, 16 which is what about after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />? Because after 24 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, the guidance is that it must continue to 18 provide a barrier against uncontrolled release.
19 So we puzzled -- I puzzled over this 20 myself when I first came from research over to NRO to 21 work on new reactor reviews and it turns out that this 22 was -- this, at least seems to me, I haven't found 23 anything -- absolute documentation, but this is about 24 the ABWR and the event system that it had that after 25 a certain amount of time, the event could lift and you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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68 1 could have a release.
2 Now that's maybe more of a controlled 3 release, because you have containment venting.
4 CO-CHAIR CORRADINI: Just for background, 5 did AP1000 address it? Which of the two is that 6 AP1000 address the possibility here for the goal? I 7 can't remember.
8 MR. SCHAPEROW: So AP1000, to my 9 recollection, did -- they satisfied both of the goals.
10 CO-CHAIR CORRADINI: They don't have to 11 satisfy both. It's one or the other?
12 MR. SCHAPEROW: It's both.
13 PARTICIPANT: Both.
14 MR. SCHAPEROW: It's both.
15 PARTICIPANT: Both.
16 MR. SCHAPEROW: It's both goals. Yes.
17 There actually was -- when we sent the Commission 18 Paper up, the Commissioner's like, well do you mean 19 one or both goals?
20 So actually, when they wrote back to us in 21 the SRM, we said we talked to the staff and we 22 understand the staff meets both goals. So they have 23 to show both.
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69 1 from Mark 1s, you know, looking at all the PRAs and 2 all. What do we want for new reactors?
3 We'd like to be more robust than the 4 operating reactors and this is what we came up with.
5 And again, this is for -- been applied to all the 6 different reactors.
7 And I -- this worked out well. We think 8 it's provided a reasonable metric for a safety 9 performance beyond core damage.
10 So we looked at NuScale's in-vessel 11 retention analysis as Etienne described in some detail 12 here and in previous meetings with ACRS.
13 We had issues with phenomenological 14 uncertainties. We went back, we took a look at what 15 we had done for AP1000 and, you know, we still had 16 kind of the same uncertainties.
17 One uncertainty deals with this potential 18 formation of a metallic layer inside of the core 19 debris either on top of it or maybe in the middle 20 somewhere.
21 And focusing -- conducting the heat 22 outward to the walls of the vessel and really focusing 23 the heat on a very, very small area of the vessel all 24 the way around, like, a little belt.
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70 1 reactions. There are some intermetallic reactions of 2 zirconium. Molten zirconium can react with steel, 3 forming -- and it can give off heat.
4 It's an excessive thermic reaction there 5 and the third uncertainty dealt with critical heat 6 flux. Now as Etienne pointed out, the experiments 7 that he used were for critical heat flux experiments 8 at atmospheric pressures.
9 And you can get additional benefit from --
10 if you're under pressure, you can get better heat 11 transfer. The issue that we had was with geometry.
12 So to me, the bottom of the containment 13 vessel in NuScale is about a one foot gap. And the 14 tests, I don't think they look quite like that.
15 I thought it was never a big tank. It was 16 -- there's more room around it. And similarly, under 17 the bottom of the containment, there is a -- there's 18 limited space there under the bottom of the 19 containment as well.
20 So we had some issues with this. And so 21 if you'll turn to slide 7, this is kind of the 22 punchline.
23 So we took a look at this again and said, 24 well is there another way we can review the design 25 without just applying the goals as they are stated in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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71 1 the SECY Papers?
2 And so we came up with alternative 3 criteria. On the probabilistic side, we want to keep 4 that probabilistic goal, but if in-vessel retention is 5 an issue, we're not going to apply this goal.
6 For the cases where in-vessel retention 7 was an issue, we would apply another goal, no large 8 release, which is really kind of the intent of these 9 containment goals is to provide a large release.
10 Not that the containment has to stay 11 intact. That's not really -- it's more about keeping 12 the fission products away from the people.
13 So if we can convince ourselves that the 14 -- we would have a large release, then we could 15 convince ourselves that this goal is satisfied.
16 We did send up a SECY Paper to the 17 Commission about a week ago on this. It's not been 18 made public.
19 There's a process where there's about a 20 two-week delay between when the paper's issued and 21 when it's made public for various reasons. I'm not an 22 expert on that.
23 MEMBER BLEY: But just for clarification, 24 we have a copy of that --
25 (Simultaneous speaking.)
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72 1 MR. SCHAPEROW: That's correct. The ACRS 2 -- you were on distribution. We notified ACRS, I 3 think --
4 MEMBER BLEY: '47, 190047.
5 MR. SCHAPEROW: Yes. So the ACRS's 6 already -- the Committee already has it, but again, it 7 has to go through this --
8 MEMBER BLEY: We've all read it. Don't 9 worry.
10 MR. SCHAPEROW: Bureaucratic process that 11 we have --
12 MEMBER BLEY: Some of us haven't seen it.
13 Is it in our -- for this meeting?
14 MR. SCHAPEROW: Yes.
15 MEMBER BLEY: Okay.
16 MR. SCHAPEROW: It's not long.
17 MEMBER BLEY: Missed it.
18 MR. SCHAPEROW: It's a five-page long 19 paper and I have -- she has a lot of background in it.
20 It goes -- actually goes through a lot of the stuff 21 that Etienne talked about today.
22 MEMBER BLEY: I think what Jason -- I 23 think it's important the Members read this, because 24 what he's going through briefly here, walks through 25 with background and discussion on how they've come to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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73 1 the thing.
2 And Vesna asked some questions yesterday, 3 which are directly relevant to how they chose a 4 different approach to largely --
5 (Simultaneous speaking.)
6 MR. SCHAPEROW: Yes. I mean, other 7 reactors have the, I believe APR1400 uses CCFP 8 definition. They use -- I think they might have used 9 two.
10 I know one of them was the large release 11 frequency divided by the CDF. I know one of the two 12 that APR1400 uses that. Again, this whole thing, you 13 know, about large, you know -- so anyway.
14 So there's a five-page SECY Paper that 15 just went up that provides, again, additional 16 background beyond what I'm going through here in the 17 slides.
18 Slide 8, please. So anyway. Going back 19 to containment lower head failure, so the possibility 20 that the core debris could leave the bottom of the 21 reactor, going into the containment lower plenum and 22 then, fail the containment lower head, we concluded no 23 large release.
24 Part of it is due to reactor pool 25 scrubbing, part of it is due to scrub -- to the --
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74 1 even before the core debris relocates down through 2 these two bottom heads, a lot of the volatile fission 3 products, which are the main risk, the cesium and the 4 iodine, a lot of these have left the fuel and have 5 deposited in the reactor somewhere.
6 So they're -- even before the bottom of 7 the containment fails, the stuff is -- some of the 8 threat is gone away and, of course, the reactor pool 9 scrubbing helps in addition to that.
10 With regard to steam explosion in the 11 containment, again, we judge that there's no large 12 release even if there was a steam explosion. Again, 13 due to this deposition of the fission product aerosols 14 in the reactor and the containment.
15 Slide 9, please. With regard to high 16 pressure melt ejection, we looked at NuScale's MELCOR 17 analyses. They did a wide variety of MELCOR analyses 18 for a wide variety of scenarios.
19 And we came to the conclusion that there 20 really would not be a significant pressure difference 21 between the reactor and the containment by the time 22 core debris had relocated downward into the lower 23 plenum.
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75 1 that there was insufficient energy released to fail 2 the containment upper head.
3 We reviewed NuScale's analysis. We 4 decided that it would be worth our while to perform 5 our own independent assessment.
6 What we did was we used the approach 7 developed by Theofanous, which is described in a 8 series of papers back from, I believe, the late 1980s.
9 They're also -- the papers are actually 10 put together in a NUREG. They're actually just one on 11 top of the other, the four papers. They're in NUREG-12 5030, NUREG/CR-5030.
13 We wrote a program, a small computer 14 program, using the Theofanous model, together with the 15 inputs for a large LWR. That model is and the inputs 16 are, again, documented in NUREG/CR-5030.
17 We encoded it and then, we changed some of 18 the inputs to scale it down to NuScale, because 19 NuScale, things are smaller. Geometry's a little 20 different.
21 And we performed this in-vessel steam 22 explosion analysis using Theofanous method and 23 NuScale's size and we concluded that the energy -- we 24 agreed with NuScale's conclusion that the energy 25 released was insufficient to shoot a slug up through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 1 and break the containment upper head.
2 DR. SCHULTZ: Jason, you -- clearly from 3 what you just described, you concluded that the size 4 features associated with NuScale didn't cause an issue 5 with regard to Theofanous approaches and evaluation 6 either analysis --
7 MR. SCHAPEROW: Yes.
8 DR. SCHULTZ: -- or conclusions.
9 MR. SCHAPEROW: Yes. Well there were some 10 differences. One of the differences in a large light 11 water reactor, there is an additional energy sink, 12 energy loss.
13 I'm presuming you understand this. I'm 14 sorry. I'm assuming you understand all the -- this 15 particular Theofanous Paper.
16 So the method is that you -- the core 17 debris falls into the water and releases energy and 18 then as the core goes up through the reactor and --
19 the chunk of core goes up through the reactor.
20 It hits different things. Energy is lost 21 at different places along the way. So one of the 22 energy losses for a large LWR is there's a shield plug 23 above the top of the reactor vessel, between the 24 reactor and the containments.
25 NuScale didn't have that energy loss, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 that was one thing we had to -- we changed. Another 2 thing we changed in the analysis was we didn't use the 3 conversion of thermal energy to mechanical energy from 4 the Theofanous step.
5 We cut that in half because there had been 6 work done since then that shows that the conversion of 7 thermal to mechanical energy would be less.
8 MEMBER SKILLMAN: Jason, there is a 75-ton 9 shield plug up there. That's a huge massive plug.
10 PARTICIPANT: That's outside containment.
11 MR. SCHAPEROW: Outside containment in the 12 --
13 (Simultaneous speaking.)
14 MR. SCHAPEROW: -- large LWR, there's a 15 hunk of -- there's a big plug between the reactor and 16 the container. But NuScale, there's really nothing 17 there.
18 It's an --
19 (Simultaneous speaking.)
20 MEMBER SKILLMAN: I understand.
21 MR. SCHAPEROW: -- there's some piping and 22 things.
23 MEMBER BLEY: Jason, a quick question. I 24 remember in one of those Theo papers, did you say 25 NUREG/CF?
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78 1 MR. SCHAPEROW: CR, NUREG/CR-5030.
2 MEMBER BLEY: I see.
3 MR. SCHAPEROW: It was a series of four 4 papers that were actually, basically, stapled together 5 and put in a NUREG.
6 MEMBER BLEY: That's the NUREG.
7 MR. SCHAPEROW: That's the NUREG. There's 8 --
9 MEMBER BLEY: Doesn't look like it's on 10 the public website. Okay. Thank you.
11 MR. SCHAPEROW: Yes. It's pretty old.
12 We'll --
13 (Simultaneous speaking.)
14 MEMBER BLEY: All right. Never mind.
15 MR. SCHAPEROW: Fifty -- that it was a 16 long time ago. But the -- it was a probabilistic 17 analysis with the deterministic stuff woven in.
18 It was a nice -- that was a very nice 19 piece of work from our -- I wasn't familiar with it 20 actually, until I started working on NuScale, but then 21 I had help from others.
22 One of the NRC staff members actually 23 worked for Theofanous back then, Hanry Wagage, who was 24 an NRO. He was a post-doc out there at ACSB.
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79 1 worked on this with me. He performed the in-vessel 2 steam explosion analysis for AP1000. He's up in 3 research.
4 So we had a -- I had significant help in 5 this area, because as most of you probably know, my 6 area's more in the fission product area and steam 7 generated tube integrity.
8 Okay. So bottom line is we performed 9 independent analysis using a different method than 10 NuScale did and we confirmed their conclusion. Slide 11 10, please.
12 The topic of severe accident induced steam 13 generated tube failure as I imagine you've heard many 14 presentations on this over the years for different 15 designs.
16 So NuScale's design, of course, looks 17 quite different than others because everything is in 18 one vessel. You can have a hot leg and a cold leg.
19 So we reviewed their steam generated tube failure 20 analysis.
21 We asked some questions. They actually --
22 we did their analysis at one point as a result of our 23 REI.
24 One issue that came up was -- I realize in 25 an earlier -- I noticed this, too, is that -- so in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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80 1 NuScale, during a severe accident, the high pressure 2 is on the outside of the tubes because they're primary 3 circuit and primary coolant is on the outside of the 4 tubes and their secondary coolant is on the inside of 5 the tubes.
6 So an issue came up about there's a 7 statement, I believe it's in the SER that talks about 8 the tubes being less susceptible to failure under 9 compression and under tension.
10 We didn't really evaluate this. We were 11 satisfied with their analysis on the basis of the 12 thermo-hydraulic conditions just -- it just wasn't 13 that hot and that high pressure.
14 It was just -- it was more mild thermal-15 hydraulic conditions. As you can tell from the 16 estimate of the probability of tube ruptures as well.
17 The other thing that stands out from 18 NuScale as being quite different is that the relief 19 valve that's on their secondary side actually relieves 20 in the containment, which is quite different than 21 large LWRs.
22 At least the ones that I'm familiar with, 23 the relief valve on the secondary side release into 24 the environment.
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81 1 the point of view of large release frequency because 2 if you have a tube failure, you don't necessarily get 3 a large release.
4 The tube failure can lead to release into 5 the containment. The only way to get a release 6 outside of the containment for a tube failure, is to 7 have the MSIDs not close or the feed water valves not 8 close.
9 Slide 11, please. So we also took a look 10 at NuScale's use of the MACCS code as part of their 11 large release evaluation for both 19.1 and 19.2.
12 One thing we noted was that they're using 13 it at a very closed in distance. This is an issue 14 that's come up, I imagine, at previous ACRS meetings.
15 I think I may have been here even. The 16 issue is that the modeling is -- MACCS has really been 17 assessed over the years for large LWRs, which maybe 18 they have larger exclusionary boundaries.
19 Their site is -- the site's bigger.
20 NuScale's site is compact and so the distance from the 21 center of the site and the site battery is only .167 22 miles.
23 So we -- again, we asked questions about 24 this and NuScale showed us as part of the audit, 25 second audit, they showed us some comparisons they had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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82 1 done between MACCS and ARCON.
2 And ARCON is a code which is used for 3 these close in distances. So they showed that in --
4 for the various cases they ran, that the concentration 5 down wind at .167 miles predicted by MACCS was always 6 higher than the concentrations predicted by ARCON at 7 those distances.
8 So that satisfied us. Also, as Etienne 9 had described this morning, there are large margins 10 between their large release definition and the doses 11 that they calculated with MACCS.
12 DR. SCHULTZ: Thank you, Jason. I had a 13 question because the way the SER read, at least the 14 way I read the SER, I was concerned that the staff had 15 done the ARCON evaluation --
16 MR. SCHAPEROW: No.
17 DR. SCHULTZ: -- but was the applicant --
18 MR. SCHAPEROW: Yes.
19 DR. SCHULTZ: -- demonstrating that for 20 the staff.
21 MR. SCHAPEROW: Yes.
22 DR. SCHULTZ: And there was one other 23 issue at about the same place and that was the --
24 using the large margins. The fact that there are 25 large margins to justify the assumptions here.
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83 1 If you could take a look at it, because it 2 appeared that that was used to differentiate between 3 the short distances or differences in short distances 4 that were used.
5 And I think if the right distances aren't 6 used in the calculation, then it ought to be correct, 7 we shouldn't --
8 MR. SCHAPEROW: The right distances were 9 used. They --
10 (Simultaneous speaking.)
11 MR. SCHAPEROW: -- the question is the 12 uncertainty in the prediction of the concentration --
13 DR. SCHULTZ: Okay.
14 MR. SCHAPEROW: -- could be -- there could 15 be some uncertainty, but we're not worried about that 16 uncertainty.
17 That's not an issue for us because, you 18 know, if your predicted dose is one rem and the dose 19 your comparing is 200 rem, we're not concerned about 20 -- that doesn't present an issue for us.
21 (Simultaneous speaking.)
22 DR. SCHULTZ: No. That's, again, if the 23 right distances were used, I don't have a concern, but 24 I was --
25 MR. SCHAPEROW: Yes.
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84 1 DR. SCHULTZ: -- confused by the way it 2 was written.
3 MR. SCHAPEROW: No. The correct distances 4 were certainly used.
5 DR. SCHULTZ: Good. Thank you.
6 MR. SCHAPEROW: We need to -- maybe we 7 need to clarify the SER.
8 CO-CHAIR SUNSERI: Yes. Jason, I have a 9 question, too.
10 MR. SCHAPEROW: Sure.
11 CO-CHAIR SUNSERI: Since this NuScale 12 definition of large release is kind of a key aspect of 13 the containment acceptability criteria that you 14 mentioned earlier, why did you choose not to 15 independently validate those calculations?
16 And essentially, it sounds to me like you 17 just accepted a presentation of NuScale's results. Is 18 that -- am I understanding you correctly?
19 MR. SCHAPEROW: So I guess I'm not sure --
20 what sort of independent confirmatory analysis are you 21 envisioning?
22 CO-CHAIR SUNSERI: Well --
23 MR. SCHAPEROW: I'm not really sure --
24 CO-CHAIR SUNSERI: -- like, rerun -- run 25 the MACCS calculation using -- inputting your own NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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85 1 parameters and assessing kind of independently --
2 (Simultaneous speaking.)
3 MR. SCHAPEROW: Yes. We didn't do --
4 that's correct. We did not do that.
5 CO-CHAIR SUNSERI: So am I correct in this 6 is a --
7 MR. SCHAPEROW: I don't know what -- how 8 to answer that other than -- let me think for a 9 second. I didn't identify any issues that would 10 suggest I would need to do something like that.
11 CO-CHAIR SUNSERI: Okay. So let me back 12 up for a second. Am I correct that the definition of 13 large release is a key criterion as far as assessing 14 the containment acceptability in accordance with the 15 two criterion you have --
16 (Simultaneous speaking.)
17 MR. SCHAPEROW: That's correct.
18 CO-CHAIR SUNSERI: All right.
19 MR. SCHAPEROW: And this is important.
20 CO-CHAIR SUNSERI: So that's an important 21 value from a regulatory and safety, public. So what 22 you're saying is we're kind of accepting the NuScale 23 analysis as opposed to independently verifying it?
24 MR. SCHAPEROW: There was one case --
25 CO-CHAIR SUNSERI: Am I right?
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86 1 MR. SCHAPEROW: -- where I did an 2 independent check. It was in the area of if -- I did 3 an independent check on one of NuScale's dose 4 calculations. It was the one for -- during refueling.
5 If you dropped the upper internals on the 6 core and you had a destructive on the core such that 7 the gap activity was released from each of the fuel 8 rods.
9 I did a confirmatory calculation on that.
10 I took the NuScale analysis -- and on the NuScale, the 11 one that -- the RADTRAD calculation that was done for 12 the fuel handling accident and I multiplied that times 13 the number of assemblies in the core.
14 Instead of dropping one assembly, I 15 assumed 37 were dropped and I calculated a dose for 16 that and I got a number that wasn't too different than 17 what NuScale had calculated with MACCS.
18 But yes. In a sense, I think I was 19 comparing two NuScale calculations together. I 20 believe they were from different code and NuScale 21 could correct me if I'm -- I'm pretty sure that the 22 field handling accident was done with the RADTRAD 23 code.
24 It's not something we typically do. I 25 don't remember any of our other DCA reviews where we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 did MACCS calculations for Chapter 19. We have -- we 2 are doing -- I believe we're doing some MACCS 3 confirmatory analysis for the environmental report.
4 But I really can't speak to that. I don't 5 know if -- well we have our reviewer of the 6 environmental report here in the room if he would like 7 to add something.
8 MR. PALMROSE: Don Palmrose, senior 9 reactor engineer. I'm the principal technical lead 10 for the environmental review looking at the outside 11 consequences going to the cost benefit analysis.
12 And we do do our own confirmatory 13 calculations using our own determined staff input 14 values for those things, calculations, to confirm what 15 NuScale has done.
16 CO-CHAIR SUNSERI: And I presume you 17 didn't see any issues then, with those calculations:
18 MR. PALMROSE: That review is still under 19 way.
20 CO-CHAIR SUNSERI: Okay. Thanks.
21 DR. SCHULTZ: Jason, what about 22 confirmatory calculations related to the accident 23 source term methodology report?
24 MR. SCHAPEROW: Yes. So first of all, 25 I'll open up broadly with that review's still under NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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88 1 way and we just got revision three of the report maybe 2 a month ago.
3 And we're pouring over it and asking 4 questions, but we have been looking at that report 5 since we got rev one and rev zero.
6 DR. SCHULTZ: Right.
7 MR. SCHAPEROW: So from our point of view 8 in our area, the 19.1 and 19.2 and severe accident 9 stuff, we've been looking at their development of 10 source term.
11 We've been looking at their deposition in 12 the containment. We have done a little bit of 13 independent analysis, which you can find in Chapter 5 14 of the MELCOR independent confirmatory analysis 15 report, which is cited in --
16 CO-CHAIR CORRADINI: And we'll talk about 17 that in closed session today.
18 MR. SCHAPEROW: Okay.
19 CO-CHAIR CORRADINI: Thank you.
20 MR. SCHAPEROW: So we do have some 21 information in there, because we, you know, we're very 22 just -- yes. NuScale's unique. It's a different 23 design.
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89 1 these -- for these, for Chapter 15. So that was of a 2 lot of interest to me, certainly.
3 Okay. Slide 12, please. Okay. So module 4 drop accidents. We took a look at what NuScale had 5 done for module drop accidents and had basically run 6 MELCOR with the module on its side. And they've done 7 some work.
8 So we took a look at it and we thought, do 9 we want to build a deck like that? And we said, no.
10 We decided we just didn't.
11 We thought, well let's focus on pool 12 scrubbing, because we think that's a big deal. So 13 that was what we directed our REI. So we met with 14 them.
15 We asked questioned. We sent them an REI.
16 We met with them. They sent a response. We met with 17 them. They sent another response. Finally, after a 18 couple of REI responses, we got all our questions 19 answered.
20 The kind of the crux of the issue for me 21 was that they were using a model for reactor pool 22 scrubbing, which was based on tests done with iodine 23 vapor from back in the early -- this was the paper 24 that goes to the -- they cited a paper from back in 25 1972.
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90 1 And, you know, based -- since then, you 2 know, the state of the art is that these releases are 3 aerosols and we made a fuss about that. And so 4 NuScale did some supplemental analysis.
5 And basically, to demonstrate that the 6 numbers they were using from these earlier studies 7 were okay.
8 They were conservative as far as 9 decontamination by a large pool of water like this.
10 And slide 13, please.
11 Finally, as we have done with pretty much 12 each of the DCA reviews, we did an independent 13 confirmatory analysis with MELCOR.
14 This is one of -- perhaps MELCOR's main --
15 the main reason we developed it back -- starting way 16 back in the '80s was so we can have a tool for 17 independent confirmatory analysis.
18 Of course, we've used it for a lot of 19 things since then, a lot of other things beyond 20 confirmatory analysis.
21 We performed simulations for NuScale for 22 three of the scenarios that NuScale reported in the 23 FSAR. The first scenario was inadvertent opening of 24 a reactor vent valve.
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91 1 containment. And the third was CVCS break outside 2 containment.
3 And we also did perform some sensitivity 4 calculations, let's see, as is customary and required, 5 because we learned some things along the way.
6 The staff simulations, we produced the 7 applicant simulations, but we did see some timing 8 shifts as is, again, not unusual for two different 9 people running a severe accident model.
10 We do believe that the differences are 11 unlikely to affect the applicant's severe accident 12 mitigation analysis. Slide 14. Thank you.
13 MS. GRADY: Good morning. This is Anne-14 Marie -- thank you. Good morning. This is Anne-Marie 15 Grady.
16 And as Dr. Corradini has already 17 mentioned, in June, we're going to come back to 18 describing our analyses for combustible gas control in 19 Chapter 6 and the exemption requests.
20 And most of our review on combustible gas 21 control will be discussed then. We'll be discussing 22 the scenarios that were considered, the severe 23 accident scenarios.
24 We'll be discussing a combustion event in 25 the containment leading to a detonation in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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92 1 containment, the effect of the detonation event on the 2 containment structure.
3 Those are all for scenarios that are for 4 intact containments. I'm here just to talk about 5 today one specific event, which is not for intact 6 containment.
7 And I'm only here discussing it in the 8 context of Chapter 19, because it's going -- it's only 9 being looked at from the point of view not of how it 10 challenged containment integrity, because it doesn't, 11 containment no longer has integrity, but whether or 12 not it would lead to multi-module risk.
13 Okay. So the scenario that is -- the 14 scenario that was evaluated by NuScale, as a 15 supporting study for their PRA, was CVCS line break 16 outside containment.
17 It was a multiple failure scenario. The 18 containment -- the reactor tripped. The containment 19 isolated and nothing else worked. No mitigation 20 whatsoever.
21 No ECCS. No DHRS. No CFDS injection. No 22 operator action. So multiple failure. So that's what 23 they're talking about here.
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93 1 all the time traditionally, because of 50.44C, but 2 there was no chance of -- there was no hydrogen 3 combustion.
4 No problem. However, they said, because 5 we now no longer have an intact containment where we 6 have hydrogen leaking out through the broken CVCS 7 line.
8 And it's now going into the area space 9 under the bioshield and the bioshield has plenty of 10 oxygen.
11 So they said in their calculation that the 12 conditions for combustion, i.e., the conditions that 13 MELCOR predicts, which are concentrations, hydrogen 14 and oxygen, would support combustion.
15 And their expectation, this is a 16 qualitative statement now, is that it would be 17 sufficiently low concentration, would get disbursed 18 throughout the reactor building and it would be a non-19 issue with respect to hydrogen combustion and they 20 stopped there.
21 So we sent them an RAI on this and asked 22 them to evaluate not just thinking that it was going 23 to be disbursed throughout the reactor building, but 24 evaluate the potential for combustion and/or 25 detonation underneath the space in the bioshield.
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94 1 And we did a confirmatory -- okay. They 2 responded to that by saying this was a low frequency 3 event, the concentrations would be very low, they 4 didn't see the need to evaluate it further and stopped 5 there.
6 But we didn't. So we asked our 7 contractor, ERI, to do a confirmatory calculation to 8 see whether or not there would be conditions that 9 would support combustion in the space underneath the 10 bioshield.
11 They have done that analysis. It's still 12 in draft form. It's almost completely reviewed, but 13 then, NuScale redesigned the bioshield.
14 So now we had an -- their response as to 15 what took place under the bioshield was based on their 16 old bioshield design.
17 Now we have a new bioshield design. We 18 have -- there's (Telephonic Interference.) there's a 19 rough --
20 CO-CHAIR CORRADINI: Somebody is having 21 their speakerphone on. Can you please go to mute, 22 that's on the open line? Thank you.
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95 1 today of the bioshield.
2 It's a proprietary document. So I don't 3 know how much to say. I think I could say at high 4 level, the primary difference is that now the design, 5 in fact, does not have an enclosed space that relied 6 on some action of some panels, but rather, it has 7 fixed vent openings.
8 That I can say and if we want to talk 9 about the details of it, we can talk about it in 10 closed session.
11 So that's where we are right now. We are 12 evaluating, using our contractor, whether or not, with 13 the new design, there would -- the hydrogen and oxygen 14 concentration would reach combustion when it's -- if 15 it doesn't, case closed.
16 If it does, then we have to look to see 17 whether or not it would ever get to the concentration 18 when it's -- of detonation and then, if we have 19 detonation, then we have to look more broadly as to 20 whether or not there could be structural damage such 21 that something could go shooting off and impact one of 22 the other modules.
23 I'm sure everybody in this room hopes 24 that's not what we're going to find. We're going to 25 be able to screen it out on concentrations, but right NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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96 1 now, we are having this evaluated.
2 We don't have the final response yet and 3 we're looking at it solely from the point of view of 4 multi-module risk. So that's why we're talking about 5 it in this chapter.
6 And the RAI that's listed here, 94.47, has 7 a lot of the technical details. It doesn't have a 8 drawing of the new bioshield.
9 However, we have access to that in the 10 electronic reading room and I've seen that and many 11 people -- many reviewers are concerned or interested 12 in the design of the bioshield structurally is, RPAC 13 is, fire protection is.
14 I mean, many people have looked at it. So 15 it's of great interest and as Zac Rad mentioned, it's 16 discussed in many places in the FSAR.
17 I don't -- I can't show you a figure if 18 you want to see the new design. But perhaps NuScale 19 could provide one for the closed session.
20 CO-CHAIR CORRADINI: We can wait, if 21 necessary, until the closed session.
22 MS. GRADY: Thank you.
23 CO-CHAIR SUNSERI: Anything else?
24 MS. GRADY: No.
25 CO-CHAIR SUNSERI: So your summary?
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97 1 MS. GRADY: Go ahead.
2 MR. SCHAPEROW: This is just -- so first 3 one Anne-Marie just talked about, we have -- let me 4 back up. So we have two open ends left, basically, 5 after reviewing 19.2.
6 The first one is what Anne-Marie talked 7 about with the shield redesign. We want -- we've 8 asked questions about multi-module risk resulting from 9 hydrogen combustion under the shield.
10 The second one we have not talked about 11 today. This is the -- as part of the accident source 12 from topical report review, NuScale is going to be 13 proposing changes to a Section in 19.2 called 14 Equipment Survivability.
15 They're going to be giving us new FSAR 16 pages. I think on March -- on May 23rd. So we're 17 going to be getting them, like, in a week or so. So 18 this -- we couldn't close -- as much as we like to 19 close out, we couldn't. So this one is under review 20 and, again, more to come on the accident source term.
21 Okay. That concludes our formal presentation.
22 CO-CHAIR SUNSERI: All right. Great.
23 Thank you very much. Members, any comments or 24 questions for staff? All right. So it's -- we're way 25 ahead of schedule.
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98 1 I'll check with -- let me check with 2 NuScale here. Are you guys ready to start the multi-3 module stuff before lunch?
4 PARTICIPANT: I think they're looking it 5 over.
6 PARTICIPANT: NuScale's input.
7 CO-CHAIR SUNSERI: NuScale Chapter 21.
8 Right? Okay. How about the staff? You guys do your 9 --
10 MR. CRANSTON: Let's just -- staff is not 11 here and that was going to be after lunch. So --
12 CO-CHAIR SUNSERI: Well let's back up. So 13 we're now at the stage where we're scheduled for 14 Chapter 21. Does NuScale have presentation for 15 Chapter 21?
16 PARTICIPANT: Yes.
17 MS. NORRIS: We will, but they're not --
18 this is Rebecca Norris of NuScale. We will, but 19 they're not here right now. We are ready for 19.3.
20 If we could do that instead?
21 CO-CHAIR SUNSERI: Yes. That's fine. So 22 what I would like to do is take a 15-minute break 23 until 25 till and then at 25 till, we convene with 24 whatever we can go with today.
25 MR. SCHAPEROW: May I interject?
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99 1 CO-CHAIR SUNSERI: Yes, sir.
2 MR. SCHAPEROW: I'm sorry. Jason 3 Schaperow. You've mentioned --
4 CO-CHAIR SUNSERI: Hold on.
5 CO-CHAIR SUNSERI: Yes, Jason, go ahead.
6 MR. SCHAPEROW: You mentioned closed 7 session and I'm -- do we -- now I have a break now, 8 because other sections we haven't presented. Do I 9 need to develop something for --
10 CO-CHAIR SUNSERI: This closed session is 11 scheduled for later this afternoon. So --
12 MR. SCHAPEROW: If you have certain plans 13 of what you'd like to hear about, that would enable me 14 to --
15 CO-CHAIR CORRADINI: So I would say for me 16 alone -- we've lost our other curious parties. But 17 for me alone, I think the number of assertions in the 18 MELCOR compilations.
19 MR. SCHAPEROW: Okay. Thank you.
20 CO-CHAIR CORRADINI: So for closed 21 session, we want to hear about that at the very least.
22 MR. SCHAPEROW: All right.
23 CO-CHAIR CORRADINI: I think Anne-Marie 24 volunteered NuScale who volunteered discussion of the 25 biofield.
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100 1 (Laughter.)
2 CO-CHAIR CORRADINI: To the extent that we 3 can understand where they're going with it, that would 4 just be good for the future. But I would think for 5 sure MELCOR.
6 MR. SCHAPEROW: Thank you.
7 CO-CHAIR CORRADINI: In the closed session 8 and that's scheduled at the end of the day.
9 MR. SCHAPEROW: Got it. I'll -- are we 10 going to --
11 PARTICIPANT: Sorry.
12 (Simultaneous speaking.)
13 FEMALE PARTICIPANG: -- question.
14 CO-CHAIR CORRADINI: They just want to 15 know if we're going to have a closed session. The 16 answer's yes.
17 MR. SCHAPEROW: Is there anything else you 18 want to hear about at the end of the day? Because I'm 19 going to be out on break for a while because you'll be 20 discussing other sections I wasn't working on.
21 MEMBER REMPE: So are we, you and me, 22 allowed --
23 CO-CHAIR CORRADINI: Yes.
24 MEMBER REMPE: -- to ask questions about 25 --
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101 1 CO-CHAIR CORRADINI: Yes.
2 MEMBER REMPE: -- the MELCOR analysis?
3 CO-CHAIR CORRADINI: Yes.
4 MEMBER REMPE: Detailed on some of the 5 input assumptions?
6 CO-CHAIR CORRANDINI: Yes.
7 MEMBER REMPE: Okay. If I'm allowed to, 8 yes. I'd like to discuss that a lot.
9 MR. SCHAPEROW: I will be here. Thank 10 you.
11 CO-CHAIR SUNSERI: All right. So what are 12 we -- all right. I'm only in charge here. So what 13 are we doing?
14 MR. SCHAPEROW: Break.
15 PARTICIPANT: Take the break.
16 CO-CHAIR SUNSERI: What are we doing after 17 break?
18 MR. SCHAPEROW: They're getting their 19 people here for --
20 (Simultaneous speaking.)
21 CO-CHAIR SUNSERI: For 19.3?
22 (Simultaneous speaking.)
23 CO-CHAIR SUNSERI: Okay. So at 25 till, 24 we will start with 19.3 25 PARTICIPANT: They're all it. They're all NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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102 1 over on NuScale.
2 MS. NORRIS: This is Rebecca Norris with 3 NuScale, we have a question. I don't know. Are we 4 still -- okay. We can also do 19.2 closed first, 5 which would be convenient because one of our --
6 PARTICIPANT: We can't do that.
7 MS. NORRIS: Just for scheduling?
8 PARTICIPANT: Are we still on the record?
9 PARTICIPANT: Well they are discussing 10 schedule. Is it scheduled to be on the record?
11 CO-CHAIR SUNSERI: No. We're going to go 12 with 19.3 after the break.
13 MS. NORRIS: Okay.
14 CO-CHAIR SUNSERI: We're on recess until 15 25 till.
16 (Whereupon, the above-entitled matter went 17 off the record at 10:21 a.m. and resumed at 10:40 18 a.m.)
19 CO-CHAIR SUNSERI: We're back in session.
20 I apologize for being a little late on that, but we 21 readjusted the schedule. We need a little extra time.
22 So the way I am told that the actions are 23 going to play out today is, we're going to do 19.3 24 next, NuScale, followed by staff. Then we'll roll 25 into 21, NuScale and staff. I don't know if we'll get NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 1 to that before lunch, but it depends on how fast this 2 goes. Then we'll go back to 19.5, NuScale and staff.
3 After that, we'll have the closed session. Everybody 4 on board with that?
5 All right. So let me turn it over to 6 Rebecca then, for the 19.3.
7 MS. BRISTOL: Thank you. So again, I'm 8 Rebecca Norris. We still have Zachary Rad, Sarah 9 Bristol, and Bill Galyean up here at the front table.
10 We are going to be presenting on SR Chapter 19.3, as 11 was already stated. This is the Regulatory Treatment 12 of Non-Safety Systems.
13 In Section 19.3, we evaluated the 14 regulatory treatment of non-safety systems and the 15 various criterion of that, and those from criterion 16 alpha, bravo, charlie, delta, and echo, there were no 17 RTNSS SSCs in the design that met that criteria for 18 the NuScale design.
19 CO-CHAIR CORRADINI: So I don't -- I want 20 to understand this, because every time we talk about 21 RTNSS I get confused. So let me make sure I've got it 22 right. Is that using the risk-significant measures 23 from your topical report, you've identified things 24 within the PRA that are risk significant. That's 25 correct?
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104 1 MS. BRISTOL: That's correct.
2 CO-CHAIR CORRADINI: So what is the 3 screening criteria, then, that it falls into or out of 4 RTNSS, given you had a list on slide -- I've got to 5 find the slide now --
6 MS. BRISTOL: The candidates from 19.1.
7 CO-CHAIR CORRADINI: And where you went 8 through in 19.1 -- I'm trying to understand how that 9 list then gets reviewed that either falls into or out 10 of RTNSS. Can you help us, please? Help me?
11 MS. BRISTOL: Yes. As the RTNSS criteria, 12 charlie, is SSC's functions relied on underpower 13 operation and shutdown conditions to meet condition 14 goals of core damage frequency of less than 1 E minus 15 4 per reactor year, and large release frequency of 16 less than 1 E minus 6 per year, and this is taking 17 credit for safety-related components.
18 So on the Level 1 insights and the Level 19 2 insights we've provided, we had a bullet that said 20 there were no additional components needed to meet 21 those thresholds.
22 CO-CHAIR CORRADINI: So just walk me 23 through a little slower. So on slide 24, you have an 24 important slide which went through what's important 25 from risk for me from a significance determination.
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105 1 But because of the definition of what would go into 2 RTNSS, none of those rose to the level that would be 3 RTNSS?
4 MS. BRISTOL: That is correct. They 5 didn't need the RTNSS criteria. They were still 6 candidates for risk significance, and so as an 7 example, as discussed in FSAR 17.4.3.3, there may be 8 risk- significant components as determined by the D-9 RAP panel, but they aren't classified as RTNSS 10 components.
11 They could still be risk significant; for 12 example, a crane is a component that is risk 13 significant, but not RTNSS.
14 As well as then there could be components 15 that the panel in 17 determined weren't risk 16 significant based on the PRA, but they could still 17 include them in D-RAP.
18 CO-CHAIR CORRADINI: Okay.
19 MS. BRISTOL: So for the purposes of 20 RTNSS, it is looking at the components needed to meet 21 the two criteria for core damage frequency 1 E minus 22 4, and large-release frequency 1 E minus 6.
23 And if you can solve your PRA with only 24 crediting safety-related components, if you can meet 25 those thresholds, then there are no additional NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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106 1 components needed to be classified as RTNSS.
2 CO-CHAIR CORRADINI: Okay. All right, I 3 think I've got it. Maybe I'm the only one who got it.
4 MS. BRISTOL: And the additional criterion 5 that the PRA looked at was criterion delta, and that 6 was the containment failure probability than 10 7 percent, and we met that as well.
8 CO-CHAIR SUNSERI: Members, any questions?
9 All right. Well, we thank you for 19.3 presentation.
10 Are you done?
11 MS. BRISTOL: There was also, similar to 12 19.1 and 19.2, we do have a COL item that confirms 13 that for a COL applicant, they will also look at the 14 PRA and the various criterion for RTNSS and confirm if 15 there is any additional components required for the 16 COL applicant, based on the criterion for that design.
17 CO-CHAIR SUNSERI: All right. Very well.
18 MS. BRISTOL: Thank you.
19 CO-CHAIR SUNSERI: So if we can have the 20 staff come up now?
21 MR. CRANSTON: I'm Greg Cranston. We're 22 ready to start on 19.3, and I'll just pass it on to 23 Odunayo to commence.
24 MR. AYEGBUSI: All right, good morning.
25 So the applicant had two slides; I have one slide, if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 you have any questions.
2 CO-CHAIR SUNSERI: Well, we'll be sure to 3 take about four hours, then.
4 MR. AYEGBUSI: So my name is Odunayo 5 Ayegbusi, and I'm a risk and liability analyst in the 6 Office of New Reactors, and I'm here to talk about 7 regulatory treatment of non-safety systems.
8 So the way I would actually like to start 9 is kind of going back to the question that was asked 10 to the applicant, and that is, this program or process 11 was put together as a result of potential concerns 12 with the performance of passive safety systems and 13 passive design class.
14 And so the idea behind it was to try to 15 identify non-safety-related SSCs that would be 16 required to support the passive safety systems, given 17 a set of criteria.
18 So these would be potentially active 19 systems; they're not safety-related, but if they met 20 the criteria, are determined to be, I would say, 21 safety significant to support the passive safety 22 systems.
23 And so our guidance talks about risk 24 significance. I actually want to stress that when our 25 guidance talks about risk significance in this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 1 particular case, it's not talking about just a PRA.
2 It's talking about an overall risk significance.
3 So when you look at the set of criteria 4 which I will go through, you will see that there are 5 some cases where it's not probabilistic; it's actually 6 deterministic. So I wanted to talk about that up 7 front.
8 So the other thing I wanted to mention is, 9 there is no regulation for RTNSS. So this RTNSS came 10 out as a result of SECY papers that went up to the 11 Commission that supported doing this, as I mentioned, 12 because of the potential concern or certainty of 13 passive systems.
14 So it's all policy-driven, and we 15 developed our SRP 19.3 to provide answers to staff on 16 how to review this. There's also guidance for the 17 contents of the application that the applicant uses.
18 So as I mentioned earlier, there is a set 19 of five criteria, and I was going to go through each 20 one and discuss what we looked at to come to our 21 conclusion about the application.
22 Actually before I go there, the other 23 major item I wanted to point out was, the way the 24 guidance is set up is, any SSC that is identified as 25 a RTNSS SSC gets placed into D-RAP. And it gets put NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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109 1 into the D-RAP, the Design Reliability Assurance 2 Program.
3 It gets put into that list, and the reason 4 for that is, when you look at our guidance, if it's a 5 RTNSS SSC, similarly with the D-RAP program, such SSCs 6 have to have certain treatments, such as augment 7 equality, QA for non-safety systems, feeding those 8 SSCs into the maintenance rule, creating short-term 9 availability controls for non-safety systems that fit 10 into the RTNSS program. So I wanted to just mention 11 that as well.
12 So for the first criterion it's completely 13 deterministic. There is some reliance on the PRA, but 14 the first criterion has to do with ATWS, which is the 15 50.62 rule and the station blackout rule, 50.63 rule.
16 In that case what we're looking for there 17 is, are there any non-safety active systems that the 18 design would rely on to respond to these events? And 19 in this particular case, the applicant did not 20 identify any.
21 When we reviewed this, the sections that 22 discuss ATWS and section that discussed station 23 blackout, the staff did not identify any SSCs that 24 would meet the requirements.
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110 1 Chapter 19, which is the PRA. By looking at the 2 analysis that's done for ATWS, and looking at if that 3 meets the 10 D minus 5 criteria.
4 So that's criterion alpha. Criterion 5 bravo has to do with any non-safety SSCs, active 6 systems that are requested, post-72 hours, up to seven 7 days. Again, reviewing the respective sections of the 8 design, the staff didn't identify any SSCS.
9 As the applicant has presented, the 10 ultimate heat sink supports the passive safety-related 11 systems for a period of greater than seven days. So 12 in that case, there were no SSCs identified there.
13 That criterion, bravo, also has us look at 14 seismic considerations. So we looked at the PRA-based 15 SMA, and in that case there were no non-safety-related 16 SSCs that were identified.
17 So criterion Charlie is a criterion that 18 utilizes the PRA. This criterion is basically looking 19 at can you perform a sensory study that shows that 20 using the safety systems alone, your results will 21 still be consistent with the Commission's safety 22 goals.
23 So in this case, if you ran your PRA with 24 no non-safety systems, you'll still meet the 25 Commission's safety goals of 10 to the minus 4 CDF, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 1 and 10 to the minus 6 for the LRF.
2 And in this case, again, the applicant 3 didn't identify any, and when we reviewed the 4 respective sections of Level 1 or Level 2 of the 5 applicant's results, there was no cause for the staff 6 to identify any SSCs that would meet this criteria.
7 So criterion delta: this has to do with 8 any non-safety-related SSCs that area required to meet 9 the continued performance goals. In addition with 10 that, any non-safety-related SSCs are required to 11 support, to respond to consider a bypass event during 12 a severe accident.
13 So again, when the staff reviewed sensory 14 studies for Level 1 and Level 2, as I mentioned in 15 regard to that, a focused PRA, the continual 16 performance goals when that continual performance goal 17 is really a 0.1 factor.
18 And when we reviewed 19.2, 19.2 is severe 19 accidents. This is a discussion about steam generator 20 tube failure. In response to that event, there are no 21 non-safety-related SSCs that are required to respond.
22 So again, this staff didn't identify any 23 SSCs that the applicant should have added to this 24 list.
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112 1 SSCs that rely on preventing significant adverse 2 system interactions between safety systems and non-3 safety systems. So in this case we looked at Section 4 6.2 of the SR. We looked at passive SSC design, and 5 we looked at potential isolation from non-safety 6 systems, and what we found was the applicant 7 identified that the pressurized heaters are a 8 potential area for an adverse interaction.
9 However, the pressurized heaters are in 10 the reactor vessel, so when you look at that event, 11 there is, in consideration of adverse issues, there 12 are systems in place that will prevent the heaters 13 becoming uncovered, and that's done by a safety 14 system. So again, to prevent that scenario no non-15 safety-related system is required.
16 Lastly, the ultimate heat sink is cooled 17 by non-safety systems, but as I said earlier, the 18 ultimate heat sink has been shown to provide passive 19 cooling for greater than seven days.
20 And then for the cooling systems, there 21 are no connections to the ultimate heat sink pool at 22 levels that would challenge the potential cooling of 23 the pool for the passive systems.
24 Lastly, I wanted to say that the 25 applicant, as our guidance has required, has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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113 1 identified a COL Action Item for a COL applicant to 2 perform a review of the DC RTNSS and then develop any 3 potential site-specific RTNSS SSCs and add that to 4 that list. So that's adequately captured in the DCA.
5 So in conclusion, as the slide says, based 6 on the staff's evaluation, we found the applicant's 7 application reasonable and adequate.
8 CO-CHAIR SUNSERI: Good discussion. Just 9 one question: in one of the first criterion you 10 mentioned was dealing with the performance of the 11 plant during a station blackout.
12 And I know it's been a long time since 13 we've reviewed this, but we did review a chapter on 14 highly-reliable DC electrical system, I think. So 15 that system, if you consider it in the scope of a 16 station blackout response still screamed out as not 17 needing to be -- I presume since it's a highly-18 reliable system, it's not a safety system, right? So 19 it would be a candidate for this criterion, but it 20 screamed out.
21 MR. AYEGBUSI: So for all of the five 22 criteria we have, for each one of them, the question 23 you're asking is, do you need a non-safety-related 24 SSC?
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114 1 design for the criterion you mentioned, the station 2 blackout, the highly-reliable DC system, battery 3 system, is not required in response to a station 4 blackout. So the answer in this case would be no.
5 CO-CHAIR SUNSERI: Fair enough, thank you.
6 Any other questions from members?
7 CO-CHAIR DIMITRIJEVIC: Actually, I have 8 two: one is an observation and one is sort -- even 9 though I understand how this happened, I'm a little 10 concerned about.
11 One observation is that when you this was 12 designed there was the 50.69 maybe wasn't for the 13 federal -- you know.
14 So when 50.69 is in operation, you have a 15 different categories of the safety and non-safety 16 systems which are very significant, it may happen that 17 these two things are in opposition to each other.
18 Because the one program, not many 19 applicants came with 50.69, and I understand why, 20 because the PRA are not in the stage to be applicable.
21 But if anybody applies 50.69, you will 22 have a situation the 50.69 takes component out of QA 23 program and all a bunch of regulations, and this 24 problem puts them back. That's an interesting thing 25 which may happen.
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115 1 Maybe the future will be lucky, and the 2 people actually using 50.69, then it will be an 3 interesting controversy between those two -- maybe.
4 I mean, some of those things are not -- even 50.69 5 uses an expert panel, I don't think some of those 6 criteria were used.
7 The other thing is, we had the exemption 8 from the GDC 27, the criticality, and that was a long 9 time into the accident, which we analyzed, and I 10 remember that, because I wrote exemptions from all 11 that down there.
12 And I was wondering now, because the 13 charging season was considered. But that's not a part 14 of the ATWS analysis; however, it's a part of the 15 exemption request.
16 So it's not in the PRA, but it is asking 17 for exemptions based on the technicality of 18 criticality of a long time after the shutdown, so I 19 don't know if that seven-day area where the charging 20 system would be considered, which is not of safety 21 systems. I'm just wondering about it.
22 CO-CHAIR SUNSERI: I think we have 23 somebody coming to respond to that.
24 MR. SCHMIDT: This is Jeff Schmidt from 25 Reactor Systems. So on the return to power and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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116 1 exemption for PDC 27 or GDC 27 technically, the 2 replacement of PDC 27, where we're going with that 3 from Chapter 15 sampling, and I think this will help 4 inform you beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as well, is that what we're 5 looking at is a potential return to power, and then 6 we're evaluating whether the staff goals are met at 7 that point.
8 So if the staff goals are met at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 9 and longer, then we don't consider it to be a safety 10 concern.
11 CO-CHAIR SUNSERI: All right. Anybody 12 else? No? Okay.
13 CO-CHAIR CORRADINI: He just had backup.
14 CO-CHAIR SUNSERI: All right.
15 CO-CHAIR DIMITRIJEVIC: I have to think 16 about it.
17 CO-CHAIR SUNSERI: Okay. All right. Ayo, 18 go ahead.
19 MR. AYEGBUSI: Just going back to the 20 observation on 50.69 and the interaction with this 21 program, this program is not a PRA-based program.
22 RTNSS is really looking for non-safety-related SSCs 23 that will respond to very specific, deterministic, and 24 probabilistic events; well, deterministic and 25 probabilistic areas.
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117 1 So I think --
2 CO-CHAIR DIMITRIJEVIC: It's sort of an 3 old-fashioned program, because you're just looking at 4 all the division between safety and non-safety. But 5 let's say, I hope, in the near future that the 6 division will be absolute, and we will have four 7 categories.
8 So now if somebody comes with those four 9 categories, then the question is, how compatible will 10 those divisions will be? That's just my -- let's say 11 maybe somebody in the COL phase will decide to apply 12 50.69, which will make perfect sense from my 13 perspective, because a lot can be saved in the 14 procurement of the components and things like that.
15 So let's say that some of the COL 16 applicant decides to apply 50.69. It will be 17 interesting how they interact with this program.
18 MR. AYEGBUSI: So I guess my point there 19 is, if you look at something like ATWS or a station 20 blackout, our review there is looking at the 21 deterministic aspect.
22 CO-CHAIR DIMITRIJEVIC: What about the 23 expert panels in the 50.69? You covered that, but 24 some of those, I noticed, would not be covered. So it 25 will be interesting to -- to merge all of these 50.69 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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118 1 -- treatment of the non-safety systems to look into 2 how do they interact with each other and how -- 50.69 3 should actually be involved in all of this.
4 MR. AYEGBUSI: Okay. Understood.
5 CO-CHAIR SUNSERI: Is that it? Anybody 6 else? Well, thank you all for your presentation, and 7 at this point we will transition back to NuScale for 8 Chapter 21.
9 So this part of the presentation is on the 10 schedule for about an hour. We will go all the way 11 through this section, and then we'll break for lunch 12 when they're finished, and then we'll take staff after 13 lunch.
14 (Pause.)
15 CO-CHAIR SUNSERI: It looks like Sarah 16 brought some new faces with you.
17 CO-CHAIR CORRADINI: It's not the usual 18 suspects.
19 CO-CHAIR SUNSERI: We may need to have 20 introductions. We don't have name cards for --
21 CO-CHAIR CORRADINI: We have one.
22 CO-CHAIR SUNSERI: So for the recorder, we 23 will need to state your name so he can write that down 24 and notes who's speaking when you make your 25 presentation. Sarah, whenever you're ready.
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119 1 So since this is a new crowd, let's go 2 through the procedure. You have a microphone; there's 3 a button that says Push right near the edge of it.
4 Not the light; the edge of the speaker. So push that, 5 the green light comes on. Use it only while you're 6 talking and turn it off when you're not, because of 7 the background noise.
8 MS. JOERGENSEN: My name is Nadja 9 Jorgensen. I am a licensing specialist with NuScale 10 Power. We're here today to present Chapter 21, which 11 is multimodule design considerations.
12 With me I have Zach Rad, director of 13 regulatory affairs; Chris Maxwell; he's from our plant 14 startup group. We have J.J. Arthur from NuScale 15 engineering, and Sarah Bristol, whom you are very 16 familiar with. I'll turn it over to Chris.
17 MR. MAXWELL: Hi, Chris Maxwell, plant 18 startup. With 12 identical modules and a separate 19 reactor in each module, it will be operated 20 independent of the state of construction or operating 21 conditions of those other modules. And given that we 22 have shared systems between the modules, the NuScale 23 Power plant design meets the definition of a modular 24 design in accordance with 10 CFR 52.1.
25 MEMBER REMPE: So when I saw this quote on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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120 1 your slide the other day, I just don't think the folks 2 who wrote 10 CFR 52.1 were thinking about, you might 3 have all 12 of the modules in a swimming pool, with 4 one of them swimming by the others to go for 5 refueling.
6 So despite the fact that you meet the 7 definition, I just don't think that the people who 8 wrote that had envisioned that when they started it.
9 It's just a comment. So some of my questions may be 10 based on that.
11 MEMBER BROWN: You've got to be careful 12 with the words co-located on the same site. This is 13 different from co-located on the same site. That's 14 kind of a stretch. I had the same comment as Joy did 15 when I read that.
16 There are acres between co-location on 17 currently actives, and here they're eyeball to 18 eyeball. That's kind of a stretch, saying you meet 19 that. Not that we're not going to do it, but that 20 statement is kind of out of sorts.
21 MR. MAXWELL: Well, with the modular 22 design, 10 CFR 52.47(c)(3) requires for modular 23 designs an evaluation of the modular operating 24 configurations to consider common or shared system, 25 interface requirements, system interactions, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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121 1 restrictions during construction startup.
2 We'll start with discussing the evaluation 3 of the shared systems. When designing those shared 4 system, only the systems which allow for safe 5 operation of one module independent the operation of 6 any of the other modules is permitted.
7 As a result, the ultimate heat sink is the 8 only safety-related system common to multiple modules.
9 Or, stated another way, with the exception of the 10 ultimate heat sink, the safety-related systems are 11 module-specific, and they're functionally independent 12 of share systems and other NPMs.
13 So for each of our modules there exists an 14 independent emergency core cooling system, a 15 containment system, a decay heat removal system, 16 module protection system, and a set of demineralized 17 water isolation valves, safety in valves, and chemical 18 environmental control system.
19 MEMBER MARCH-LEUBA: How do you treat AC 20 power?
21 MR. MAXWELL: AC power was evaluated for 22 multimodule conditions. We'll see here a list of the 23 systems; it was evaluated.
24 MEMBER MARCH-LEUBA: But it's shared among 25 the modules, right?
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122 1 MR. MAXWELL: Shared, but not safety-2 related; that's correct. We have no safety-related 3 electrical power.
4 MEMBER BLEY: Well, AC electrical power.
5 MR. MAXWELL: No safety-related AC or DC 6 power.
7 MEMBER BLEY: I thought there was --
8 CO-CHAIR CORRADINI: You wrote the letter.
9 MEMBER BLEY: Yes, I did, but I thought 10 there was one DC.
11 CO-CHAIR SUNSERI: You have some 1E stuff, 12 right? For --
13 CO-CHAIR CORRADINI: It was non-class 1E.
14 MR. MAXWELL: We have no class 1E power.
15 We have class 1E module protection system, but not 16 power supply.
17 Because the ultimate heat sink supports 18 multiple modules, accordingly it was designed with 19 multimodule demands in mind. Regarding multimodule 20 operations, the ultimate heat sink has sufficient 21 capacity to remove the heat from one module that's 22 experiencing a design-basis accident, while 23 simultaneously removing heat from the other 11 modules 24 while undergoing a controlled shutdown and cooldown.
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123 1 cooldown, long-term cooling will be provided by 2 safety-related ultimate heat sink and the module-3 specific safety-related systems, again, without 4 reliance on any other shared systems or any operator 5 action.
6 So an evaluation of each shared system 7 that has the potential for adverse system interaction 8 --
9 MEMBER BROWN: Go back. Excuse me; under 10 the DBA with 11 shutdown and cooldown, once they get 11 there, you said they will be maintained using their 12 explicit systems. You don't need any shared systems 13 at that point, once they're to that circuit? Did I 14 understand that correctly?
15 MR. MAXWELL: Yes, sir. The one shared 16 system is the ultimate heat sink.
17 MEMBER BROWN: I understand, the big tank 18 of water, but all the rest -- there's other shared 19 systems when you're in normal operation, none of those 20 are required once you have this DBA and the other ones 21 are being maintained shutdown?
22 MR. MAXWELL: That's correct.
23 MEMBER BROWN: Okay. Thank you.
24 CO-CHAIR DIMITRIJEVIC: Well, how about 25 non-DBA, you add? Does it have the chemical capacity NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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124 1 to do this with LOCA?
2 MR. MAXWELL: Yes, it's designed for -- in 3 Chapter 15 space for accident on all 12 modules has 4 sufficient capacity.
5 CO-CHAIR DIMITRIJEVIC: Okay. So even if 6 all 12 modules are inactive, then it has enough 7 capacity to remove the --
8 MR. MAXWELL: That's correct.
9 CO-CHAIR DIMITRIJEVIC: Okay.
10 MR. MAXWELL: Yes, not to be misled by 11 that statement, specifically talking about operating 12 conditions with a module experience is a design-basis 13 accident, and the idea is that the 11 modules could be 14 undergoing normal shutdown and cooldown with the 15 ultimate heat sink capacity.
16 And then once those modules are shut down 17 and cooled down, normal long-term cooling can exist 18 with the ultimate heat sink and the module-specific 19 safety-related systems.
20 CO-CHAIR DIMITRIJEVIC: Okay.
21 MR. MAXWELL: So again, the evaluation of 22 each shared system that has the potential for an 23 adverse system interaction or an undesirable 24 multimodule interaction was conducted.
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125 1 include the cooling water systems, radioactive waste 2 systems, building HVAC systems.
3 MEMBER REMPE: Before you leave that one, 4 just so I understand: for a system like the reactor 5 component cooling water system which -- there's two of 6 them, and each one supports six modules.
7 MR. MAXWELL: That's correct.
8 MEMBER REMPE: So if you have -- and 9 again, if you have a failure in it, does that mean you 10 have to shut down all six of those modules?
11 MR. MAXWELL: I'll answer your question in 12 two slides, if I may.
13 MEMBER REMPE: Oh, okay.
14 MR. MAXWELL: And maybe more than two, but 15 I've got a direct answer to that question.
16 MEMBER REMPE: Okay.
17 MR. MAXWELL: So in addition to these 18 systems, we have support systems are evaluated. All 19 of the electrical distribution and power systems, and 20 the com and I&C systems all evaluated.
21 CO-CHAIR SUNSERI: I don't see the --
22 where's the system for drawing the vacuum on these 23 containments at? Is that up there?
24 MR. MAXWELL: The containment acquisition 25 system is module-specific, not shared.
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126 1 CO-CHAIR SUNSERI: Oh, okay.
2 MR. MAXWELL: The result of the 3 evaluations demonstrate that the shared system 4 operation does not result in any adverse system 5 interaction such as a loss of a safety function or a 6 design-basis event and a simultaneous degradation of 7 a safety function, or a design-basis event and 8 simultaneous degradation of critical operator 9 information, or a design-basis event and a requirement 10 for an operator action outside the control room.
11 That's a summary of the evaluations of 12 those systems.
13 One category of systems would be those 14 shared systems that serve one NPM at a time. And for 15 those systems, they are equipped with isolation 16 features that prevent a direct module-to-module 17 interaction during normal operation.
18 That's to say that those systems, while 19 they are capable of supplying each of the modules or 20 six modules, they are designed to be in service with 21 only one module at a time.
22 So for an adverse multimodular interaction 23 to occur as a result of a failure in that shared 24 system, one of two conditions would have to exist.
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127 1 in an abnormal lineup prior to operation outside of 2 the normal operating procedures.
3 The second would be that multiple 4 concurrent failures would have to occur within either 5 the system or the support system.
6 MEMBER BLEY: Are there interlocks or 7 anything that help prevent operators from doing that?
8 MR. MAXWELL: Not all interlocks have been 9 defined at this stage of the design.
10 MEMBER BLEY: Do you plan to have 11 interlocks, or are you counting on your operating 12 procedures?
13 MR. MAXWELL: I guess I can't answer that 14 question at this time. I'm not sure what direction 15 plant ops plans to go with.
16 MEMBER BLEY: Okay.
17 CO-CHAIR CORRADINI: But I guess to answer 18 Dennis' question, it will either be operator 19 procedures or interlocks, but you don't know which is 20 which, or which ones would be in which category?
21 MR. MAXWELL: That's correct.
22 MEMBER BLEY: If it's strictly admin, then 23 it's going to end up a line drawn some time. It's 24 just going to happen.
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128 1 you about these reactor pool cooling system. What is 2 that system?
3 MR. MAXWELL: It typically takes a suction 4 from the spent fuel -- I'm sorry -- the reactor pool, 5 goes through the heat exchanger, cooled by cycling 6 water system and returns back to the -- it goes 7 through demineralizers for cleanup if necessary and 8 returns the water back to the ultimate heat sink.
9 CO-CHAIR DIMITRIJEVIC: So there is 10 another head sink on this heat exchangers, where it's 11 cooled, right?
12 MR. MAXWELL: Site cooling water system, 13 right. It has cooling towers associated with it.
14 Water circulated from those cooling towers --
15 CO-CHAIR DIMITRIJEVIC: What will be 16 consequence of losing that system?
17 MR. MAXWELL: Of site cooling water?
18 CO-CHAIR DIMITRIJEVIC: Of that -- yes, 19 the reactor pool cooling system.
20 MR. MAXWELL: The gradual heatup of the 21 ultimate heat sink.
22 CO-CHAIR DIMITRIJEVIC: So how long will 23 it take for this to be a problem?
24 MR. MAXWELL: We also have another system, 25 the spent fuel pool cooling system, and the reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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129 1 pool and the spent fuel cooling system communicate 2 with on another during normal operations. So you're 3 still removing heat when you have the spent fuel 4 cooling system in service.
5 CO-CHAIR CORRADINI: So you'd need a 6 failure of both to actually cause the pool to heat up?
7 I think that's where she's going with that.
8 MR. RAD: I think the answer to that is 9 going to be conditional, but we'll see what sort of 10 answer we can find for your away team.
11 MEMBER BLEY: Chris, your second bullet 12 there makes sense, but I don't think, on the next 13 slide, you tell us what you do if you have adverse 14 multimodule interactions.
15 MR. MAXWELL: Well, the point of the slide 16 is that --
17 MEMBER BLEY: You don't expect them?
18 MR. MAXWELL: That's correct.
19 MEMBER BLEY: But it could happen.
20 MR. MAXWELL: Operators would respond to 21 the conditions that were generated as a result of the 22 adverse interaction, and procedures and indicators 23 that would provide them --
24 MEMBER BLEY: Probably some set of 25 abnormal procedures will exist.
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130 1 MR. MAXWELL: Yes, sir.
2 MEMBER BLEY: Okay. Those are all going 3 to have to be in place for a COL?
4 MR. RAD: So the question is, what do you 5 mean by COL? Do you mean at the application stage, or 6 do you mean prior to loading fuel and starting up a 7 module?
8 MEMBER BLEY: Well, what I mean, first, is 9 before the COL applicant gets its license.
10 MR. RAD: Generally, the answer to that is 11 no. Usually in Chapter 13, if you look at 12 applications, there's a short description of what the 13 procedures will be, and then a commitment and 14 milestone to implement them prior to a particular 15 phase. It's usually post-licensing, though.
16 MEMBER BLEY: Yes. Then tomorrow, we'll 17 get into --
18 CO-CHAIR CORRADINI: ITAAC.
19 MEMBER BLEY: ITAAC. My mind goes blank 20 on acronyms. So likely to be accurate ITAACs on 21 those?
23 MEMBER BLEY: No ITAAC? Just a 24 commitment?
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131 1 ITAAC.
2 CO-CHAIR CORRADINI: Keep on going.
3 MR. MAXWELL: Yes, sir. So in a separate 4 category is design includes reactor cooling water 5 system, RCCW. It's the only shared system that 6 directly interfaces with multiple modules and is also 7 designed to simultaneously support more than one NPM 8 at a time.
9 CO-CHAIR CORRADINI: And there's one per 10 six?
11 MR. MAXWELL: That's correct, two 12 subsystems for the entire -- if there's a failure in 13 a leak in a cooler or a heat exchanger, either in the 14 system itself or support system, the RCCW system is 15 designed to be able to completely isolate that and not 16 impact the other five modules.
17 It's also designed with sufficient 18 capacity such that a single component failure doesn't 19 result in loss of cooling to more than one module at 20 a time.
21 MEMBER REMPE: Okay. So my question is, 22 you're saying that even though it's there to support 23 six modules, if it fails, five of those six modules 24 can still keep plugging along and chugging along?
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132 1 support those five modules, is my understanding.
2 You're not allowing any sort of crossties. And so 3 this isn't an essential item for the modules, the 4 other five, that can be serviced by the RCCW system to 5 be serviced by it? They can function safely without 6 it; is that true?
7 Commonwealth Safely, yes; however, not 8 continue to operate.
9 MEMBER REMPE: Oh, you do have to shut 10 them down?
11 MR. MAXWELL: That's correct.
12 MEMBER REMPE: Okay. So are there other 13 shared systems where one supports six modules, that if 14 it goes out, you're going to have to shut six modules?
15 Or if there's one that supports four modules?
16 This is more of an investment protection, 17 but it could tie to safety, because it seems like you 18 may have a lot of modules that can't continue 19 operating because of these shared systems, and you're 20 going to be seeing a lot of shutdowns that people have 21 not really thought about because of the way this plant 22 is configured.
23 CO-CHAIR DIMITRIJEVIC: Like a circulating 24 water system?
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133 1 system; a complete loss of one -- so the answer to the 2 question is, it depends on how far back you zoom from 3 the module itself. By that I mean, right circulating 4 water, if it cools the condensers on the balance of 5 plant side, if they've experienced a complete loss of 6 one subset of circulating clean water system, you 7 would have to shut down those six modules.
8 MEMBER BLEY: You say it's the only shared 9 system that can interface with modules, but I guess 10 that means excluding electric power.
11 MR. MAXWELL: It's the only one that 12 directly interfaces with the six modules; in this 13 case, the interface requirement being a direct piped 14 connection between the shared system and the module 15 itself.
16 MEMBER BLEY: So it's the only fluid 17 system that can --
18 MR. MAXWELL: That's true.
19 MEMBER BLEY: Go ahead.
20 MEMBER SKILLMAN: What about ventilation?
21 MR. MAXWELL: Ventilation is a spatial 22 coupling between the modules, but again, not a direct 23 interface with the modules themselves. Like a 24 malfunction in the -- and the important distinction 25 there is that a malfunction in an HVAC system affects NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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134 1 each of the modules individually, but doesn't result 2 in an adverse system interaction, module to module.
3 You have to respond to that condition for each module 4 independently.
5 CO-CHAIR SUNSERI: So I think, just 6 pulling the string -- and we'll do more on Joy's 7 question -- so if you lose component cooling water, 8 you have six modules that are going to be shutting 9 down, right?
10 MR. MAXWELL: That's correct.
11 CO-CHAIR SUNSERI: Your control room 12 staffing is not large enough to deal directly with six 13 modules, so you're going to have some of these going 14 into DRS cooling, and then some of them in a 15 monitoring mode? So that's where the safety nexus is, 16 right? The distraction or the work demand on the 17 operating crew to be able to handle six simultaneous 18 shutdowns, right? Have you looked at that safety 19 implication?
20 MR. MAXWELL: We have. We did a staffing 21 plan validation and an in-grade system verification, 22 which I was lucky enough to be a part of. We were 23 challenged with multimodule interactions, including 24 the loss of RCCW.
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135 1 complete loss of RCCW; it doesn't require an immediate 2 action or an immediate response. The necessary 3 response time is measured in tens of minutes, 4 approximately 40 minutes, because of the systems that 5 are cooled by RCCW.
6 It controls our control rod drive 7 mechanisms, cools the wetdown cooler for CVCS, and 8 various sample coolers.
9 So the driver of the shutdown is the loss 10 of cooling to the control rod drive mechanism, and the 11 staffing was adequate to respond to six modules, to 12 take the actions to trip the units, isolate 13 containment, initiate the decay heat removal system, 14 passively place six modules in cooling, so safe 15 shutdown.
16 CO-CHAIR CORRADINI: So to answer Matt's 17 question differently, if you lost component cooling 18 there would be procedures such that modules one 19 through six would sequentially be shut down in a 20 normal procedural fashion such that there's more than 21 enough time to do it for all six before any one of 22 them has a critical function that needs to be cooled; 23 am I understanding this correctly?
24 MR. MAXWELL: Yes, sir, that's correct, 25 and you won't challenge a safety function on a lost NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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136 1 RCCW. There won't be a safety or a critical safety 2 function challenge by the loss of this support system.
3 CO-CHAIR DIMITRIJEVIC: But it would be 4 for loss of circulating collecting water. You will 5 challenge your decay heat removal system.
6 MR. MAXWELL: Not safety related. We 7 still have the decay heat removal system and our 8 ultimate heat sink to provide the safety-related at 9 the decay heat removal. So there's no challenge to a 10 safety function.
11 CO-CHAIR DIMITRIJEVIC: Okay. But it's 12 challenged because it has to operate. I don't know 13 how you interpret challenge. I thought that you say 14 challenge, that means that it's not required. But it 15 is -- so did you analyze the similar situation for 16 loss of feed water to six units?
17 CO-CHAIR CORRADINI: I'm looking at his 18 definition. The way the applicant has defined the 19 definition, this is the only direct connection.
20 But circulating water is not a direction; 21 nonetheless, if it fails for six units --
22 MR. MAXWELL: That's correct.
23 CO-CHAIR CORRADINI: It might require an 24 orderly shutdown.
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137 1 require a little more human action than the operator.
2 I mean, I understand you have a danger in pass or 3 fail, more things are happening, so operators have to 4 be ready. That's happening on six units.
5 MR. MAXWELL: I think an important point 6 to make is that a loss of the RCCW system or the 7 circulating water system, neither of those require an 8 operator action to maintain the safety functions.
9 The plant is designed to respond if you 10 reach actuation set point for the module protection 11 system to trip the reactor, isolate containment, and 12 initiate decay heat removal system.
13 So the safety functions are maintained.
14 These are just the support systems to in support of 15 operation, not of safety.
16 MEMBER REMPE: So we've heard that the 17 procedures aren't going to be developed until the COL 18 applicant is engaged, et cetera, et cetera.
19 But there are a lot of these support 20 systems that will require that you have some 21 procedures to say, Okay, if you lose system X, just 22 like this RCCW system, you've got to trip, in a safe, 23 orderly manner, X number of modules, depending on what 24 they are.
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138 1 a list in the DCA 21 about initiating this triggered 2 by systems, and how many systems support how many 3 modules, but is there a list somewhere that will be 4 turned over to the COL applicant about all the 5 different systems that are going to require that X 6 number of modules be shut down?
7 Am I making myself clear, what I'm trying 8 to ask here?
9 MR. RAD: I think so. We'll be present of 10 the COL application development as an entity.
11 MEMBER REMPE: So I've heard that 12 statement before, but we are talking about a design 13 certification. And historically in the U.S., once you 14 have a certified design, it may be hard, but another 15 design developer could come in and take that certified 16 design and go forward.
17 So when the agency certifies your design, 18 we need to know what we're certifying. So that's why 19 I'm asking, is that information available as part of 20 the design certification?
21 MR. RAD: We can look and see if it's part 22 of the certified design or the approved information in 23 Tier 2, Part 2, in answer to your question.
24 MEMBER REMPE: Okay. Thank you.
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139 1 after this, but I'm just going to make a statement 2 just for you to consider.
3 I heard many times your group and other 4 groups come in here and say, Well, you know, at the 5 time of event, we don't have to take any immediate 6 actions.
7 But having been in a control room, I know 8 the operators -- it's against human nature not to do 9 anything, and there will be things that they have to 10 do. They will be monitoring systems and all that 11 stuff.
12 And I know from being in a control room 13 and from all the training and background that I've 14 had, I know they are at their best when they do that, 15 because this is what they've been trained for.
16 However, on the flip side of that is the 17 most error-likely situation you can put them in. It's 18 high workload; it's a high-stakes environment, so I 19 just ask you to consider that more -- not as casually 20 as I've been hearing it stated to us, well, we just 21 hands off and watch this, or let's not do anything, 22 really.
23 MR. RAD: I think this is a good topic for 24 the onsite visit, because we've put a tremendous 25 amount of work into it. This is one of our longest-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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140 1 running programs. It's a tremendous investment for 2 us, and I think our program for HFE and all the way 3 out to the ISV program that was just discussed, I 4 think it would be very important for you guys to come 5 out and witness directly the work that we've put into 6 this, because it's not something that we've not taken 7 seriously. Like I said, it's one of our longest-8 running programs.
9 MR. MAXWELL: On that same note, I 10 mentioned the staffing plan validation that I was a 11 part of prior. I was previously a licensed SRO at a 12 commercial facility, as were the other individuals 13 during the staffing plan validation.
14 And I'll say that I had a modicum of 15 skepticism myself, having been an operating control 16 room. But I think the point that was just made is 17 that then, when you experience it and realize that 18 there are no immediate actions, there isn't that 19 pressure for the operators to take action because of 20 the design, that there wasn't a sense of time 21 pressure. You just didn't feel the same pressure that 22 I was used to, coming from a traditional commercial 23 design.
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141 1 effect is the same situation? You don't have to worry 2 about supplying batteries, starting generators, doing 3 things like that?
4 MR. MAXWELL: Yes. The design as such 5 doesn't require additional water, any operator action, 6 or electrical power, AC or DC, to maintain the key 7 safety functions.
8 So in those scenarios that we were put 9 under, we didn't know what scenario was going to be 10 run on us with our standard staffing, and we had a 11 loss of all AC power.
12 We experienced loss of circulating water, 13 loss of RCCW, multimodule, modules in ATWS with other 14 modules undergoing events, and just because of the 15 element that I spoke of, the design and the design of 16 the HSI, the interface, the information available to 17 the operators removed that pressure, because you could 18 see the automatic response of the systems and that the 19 safety functions were occurring without any operator 20 action as designed.
21 CO-CHAIR DIMITRIJEVIC: So how long would 22 your batteries last without AC?
23 MR. MAXWELL: Well, in our highly reliable 24 DC power system we have 72-hour batteries and 24-hour 25 batteries, and they have fully-redundant backups as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 well, so --
2 CO-CHAIR DIMITRIJEVIC: And you don't need 3 to shed some loads and everything to preserve them?
4 MR. MAXWELL: No. Designed to maintain a 5 minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the loads that are on them at 6 the event initiation.
7 MR. MAXWELL: The next element that was 8 evaluated was concerns during construction startup, 9 and during the construction startup, phases that occur 10 prior to the initial module fuel load, the shared 11 module-specific systems that are in the reactor 12 building, the control building, and the rad waste 13 building will be substantially completed, again, prior 14 to that initial fuel load.
15 The construction method and the phased 16 expansion of the modules provides the assurance that 17 the operating configuration is not materially 18 different from that assumed in our safety analysis, 19 and that the independence of each module's safety-20 related system is maintained.
21 MEMBER BROWN: Can you backtrack that 22 slide? I think one of the difficulties of 23 understanding how the --
24 CO-CHAIR SUNSERI: Do you have your green 25 light on?
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143 1 MEMBER BROWN: Thank you. I had it on; 2 she told me to turn it off, my mother hen over here.
3 One of the difficulties I had was, there's 4 no integrated architecture that you're showing how all 5 the shared systems are integrated across all 12 6 modules, and therefore how they are isolated, for 7 instance. The description of isolation is vague, if 8 non-existent.
9 So it's just a little bit -- the words are 10 fine, but in terms of agreeing with the concept, which 11 I don't have any problem with, but how it's done, at 12 least in a conceptual manner, it is not -- you just 13 state in Chapter 21 that they're going to do this.
14 Have I misstated that anywhere? I don't 15 remember seeing a giant, 12-module full facility 16 operation showing an architecture that shows how all 17 these shared systems are set up and how they will be 18 available for isolation, if necessary, from module to 19 module.
20 MR. MAXWELL: Right. That drawing does 21 not exist in Chapter 21.
22 MEMBER BROWN: That's what I thought.
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144 1 of showing that you've got reliable ability to isolate 2 and to not have non-adverse reactions on more than 3 one, other than the RCCW system, based on their plant 4 configuration. That just doesn't exist. So the words 5 are nice, but the implementation is non-existent.
6 MEMBER BLEY: It might be, from what they 7 said earlier, Charlie, it might be just direction to 8 the operator --
9 MEMBER BROWN: And that may not be --
10 that's not necessarily a good conclusion to come to.
11 When you have multiple and complex interactions 12 between 12 different reactors -- I can only speak to 13 that as I was a Naval nuclear guy, and the Enterprise 14 had eight reactor plants on it.
15 MEMBER BLEY: I can speak real directly to 16 that.
17 MEMBER BROWN: I know you can speak real 18 directly to that, and we fought like the devil to be 19 try to ensure that those systems, that could be cross-20 connected, whether they be electrical or fluid systems 21 -- and Dennis had direct relations to it in his 22 operations -- was very complex, relative to how you do 23 it, and you've got a similar situation with 12, and 24 two dozen different systems that you're integrating on 25 this.
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145 1 MEMBER BLEY: And if they're not normally 2 cross-connected as they inadvertently or on purpose 3 get cross-connected someday, you might not have the 4 right to balance as you would expect, and water can 5 move really fast.
6 MEMBER BROWN: So I've got real difficulty 7 with fundamentally agreeing with Chapter 21 the way 8 it's presently configured. I have no idea what we'll 9 do. That's a personal opinion; that's not a committee 10 opinion, by the way.
11 MR. MAXWELL: That information is 12 available in the equipment drawings and the P&IDs, 13 detailed P&IDs that have the valves and show those 14 isolations.
15 MEMBER BROWN: But it's like -- pardon?
16 CO-CHAIR CORRADINI: I was going to say, 17 I think all Charlie is saying is, If I go to Chapter 18 21 in isolation, there's nothing there to graphically 19 explain what you're describing.
20 MEMBER BROWN: It's like telling me the 21 module protection system is independent, four 22 channels, does all these nifty things, but I don't 23 have an architecture that demonstrates that. There's 24 no architecture that demonstrates that, and this is 25 very complex in terms of overall plant operations.
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146 1 That's just my thought process.
2 MEMBER BLEY: It would be very helpful to 3 see the P&IDs, but --
4 MEMBER BROWN: That's not enough. I know, 5 but if you look at those, and you try to stack all 6 those together individually one at a time, you just 7 don't get --- it's difficult -- I'm supposed to get 8 all my drawing tables out and start drawing all these 9 things and showing them all at the same time? I 10 didn't think so.
11 I've said my piece; I'll stop there, and 12 you can go on.
13 CO-CHAIR SUNSERI: All right, continue, 14 please.
15 MEMBER SKILLMAN: Before you change that 16 slide, let me ask this. Several meetings ago, one of 17 the lines in questioning had to do with how the build 18 out is occurring. Let's for instance say you've got 19 two, or three or four modules in operation and you're 20 bringing number five into the plant. When number five 21 arrives from the manufacturer, is number five fully 22 tested and ready to be inserted into its operations 23 bay? Or does number five find a place in the pool 24 where there is additional testing to confirm that five 25 is fit for duty before it is moved to its operating NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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147 1 bay?
2 MR. MAXWELL: The plan is that module five 3 is placed in the module five operating bay. It's 4 connected and then, as part of the initial test 5 program -- there's a set of pre-operational tests that 6 must be performed to verify that -- and ITAAC that 7 must completed prior to loading fuel for module five.
8 MEMBER SKILLMAN: Thank you.
9 CO-CHAIR CORRADINI: Actually, the way you 10 said it makes it -- I guess I hadn't thought about it 11 that way. But you can -- I have just a couple slides, 12 so somewhere in there you want to answer it. So, to 13 satisfy this pre-operational procedure -- pre-14 operational testing, and to satisfy ITAACs, it's a 15 module-by-module satisfaction?
16 MR. MAXWELL: There's a -- we will discuss 17 it in more detail with Chapter 14.
18 CO-CHAIR CORRADINI: Why don't we just 19 wait? Let's just wait. I can bring it up tomorrow.
20 MR. MAXWELL: Okay. Next slide, please.
21 MEMBER BLEY: I am still back where 22 Charlie was, and I have a question for Sarah. Did the 23 PRA consider -- or why shouldn't it have considered --
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148 1 some initiating event, added to the likelihood of an 2 initiating event, or complicated things in any way?
3 (No audible response.)
4 MEMBER BLEY: You can think about it till 5 the next time we see you, if you want.
6 MS. BRISTOL: Okay.
7 (Simultaneous speaking.)
8 MEMBER BLEY: I am interested in your --
9 CO-CHAIR CORRADINI: I'm not understanding 10 the question. Help me.
11 MEMBER BLEY: Okay. You're running the 12 plant --
13 CO-CHAIR CORRADINI: Yes.
14 MEMBER BLEY: You open the wrong valves and 15 now you've got one of these shared systems or more 16 connected to more than one unit. Which isn't supposed 17 to happen. Can that to some event? Or increase the 18 likelihood of some event that should have been modeled 19 in a particular -
20 CO-CHAIR CORRADINI: The question is, is 21 that sort of initiator considered.
22 MEMBER BLEY: Yes. And if not, why -- why 23 didn't it need to be considered? Or, its impact on 24 some initiating event frequency.
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149 1 mentioned, with examples of circ water or component 2 cooling water, we do look at a general transient. You 3 know, and so we have the initiating event frequency of 4 about one a year of these numerous trips that could 5 impact the modules. And so, if that was a shared 6 system that failed, that would be captured in the 7 single-module PRA as a general transient. And that's 8 how the plant would respond. We didn't find anything 9 specific with the loss of one of these support systems 10 that would impact or present a different event 11 sequence.
12 CO-CHAIR DIMITRIJEVIC: You also have a 13 loss of support sequence --
14 PARTICIPANT: Vesna, do you have your mic 15 on?
16 PARTICIPANT: Your mic is off.
17 MS. BRISTOL: For the loss of support 18 system initiating event frequency, we looked at the 19 loss of responding systems. So the loss of CVCS or 20 the loss of instrument air in order to respond to the 21 event. The loss of a shared system impact on a module 22 to create the initiating event was captured as a 23 general transient. The loss of support system 24 initiator was the loss of a support system needing to 25 respond to the initiating event.
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150 1 MEMBER DIMITRIJEVIC: And loss of feed 2 water?
3 MS. BRISTOL: That is correct.
4 MEMBER DIMITRIJEVIC: So do you have 5 actually fault trees for initiating events for some of 6 those systems?
7 MS. BRISTOL: We didn't calculate any --
8 outside of ECC -- as a spurious ECCS initiator, we 9 didn't have any support system faults redeveloped 10 initiators.
11 MEMBER DIMITRIJEVIC: It was just based on 12 genetic data?
13 MS. BRISTOL: Generic data, yes.
14 (Pause.)
15 MR. MAXWELL: The analysis, though, of the 16 shared systems that we talked about previously was 17 conducted for six systems -- or, six modules that were 18 talked about, subsystems -- or 12 modules for the 19 support systems that support all 12. And as a result 20 it continues to be bounding as you install subsequent 21 NPMs. As a result there's restrictions in the 22 operating configuration, or interface requirements, 23 aren't necessary to ensure that the operating NPMs 24 continue to operate safely during the installation 25 testing or startup of the subsequent NPMs. And that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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151 1 -- that concludes the Chapter 12 presentation.
2 CO-CHAIR CORRADINI: Thank you. Members, 3 any other comments?
4 MEMBER MARCH-LEUBA: So, different topic, 5 but since we have a qualified reactor operator here, 6 you lose power to your reactor and all the valves 7 close and open, depending on which direction, and you 8 go into DHRS. What mode are you in? In operator 9 action? You're definitely not in mode 1 anymore. Let 10 me remind you, mode 1 is operation, mode 2 is hot 11 shutdown and mode 3 is safe shutdown, transition --
12 MR. MAXWELL: Assuming 100-percent power, 13 the event initiation you would be in mode 2, heading 14 towards mode 3, safe shutdown.
15 MEMBER MARCH-LEUBA: Right. So you will 16 need to establish a boron concentration consistent 17 with safe shut-down before you transition from 2 to 3, 18 correct?
19 MR. MAXWELL: No, the module operates with 20 the required boron concentration --
21 MEMBER MARCH-LEUBA: For hot shutdown.
22 MR. MAXWELL: It's safe -- for safe 23 shutdown. We establish safe shutdown without 24 additional boron.
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152 1 shutdown boron?
2 MR. MAXWELL: That's correct.
3 MEMBER BLEY: But not cold shutdown, then?
4 MEMBER MARCH-LEUBA: No, not cold, but 5 safe -- safe shutdown -
6 CO-CHAIR CORRADINI: Is a temperature 7 defined quantity.
8 MEMBER MARCH-LEUBA: Yes.
9 MR. MAXWELL: I am sorry, what's?
10 CO-CHAIR CORRADINI: No, I am just 11 clarifying what I thought was --- when you say safe 12 shutdown, it is not cold shutdown. It is at a 13 specified temperature, which I cannot remember.
14 MR. MAXWELL: Four hundred and twenty 15 degrees. That's correct. That's safe shutdown.
16 MEMBER MARCH-LEUBA: Safe shutdown is 17 below -- less than 420.
18 MR. MAXWELL: That's correct.
19 MEMBER MARCH-LEUBA: And that -- it has no 20 lower limit.
21 MR. MAXWELL: That's correct.
22 MEMBER MARCH-LEUBA: So it could go to 40 23 degrees Fahrenheit.
24 MR. MAXWELL: That's correct.
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153 1 normal operation without boron concentration?
2 MR. MAXWELL: That's correct.
3 MEMBER MARCH-LEUBA: I don't believe you.
4 (Laughter.)
5 MEMBER MARCH-LEUBA: This is a very calm, 6 honest statement coming from this microphone. I don't 7 think that is correct, but we will come back to it.
8 CO-CHAIR DIMITRIJEVIC: MY question is --
9 I mean, I am not sure -- when I was reading first time 10 this chapter, I was expecting also more than just 11 shared system and I was expecting -- because you have 12 this other consideration which will be like, you know, 13 module drop during refueling affecting other modules 14 -- or, accident in one module affecting other modules.
15 How come that's not part of this section?
16 MR. MAXWELL: As far as an accident in one 17 module impacting other modules, that is part of the 18 interfacing evaluation to the point that one module 19 can't affect -- there isn't a viable interface between 20 the two that causes a multi-module interaction between 21 these shared systems.
22 CO-CHAIR DIMITRIJEVIC: Well, they're just 23 considering this hydrogen explosion impact, so I mean, 24 that is consideration for -- I mean, you know, it's 25 not -- well, you may conclude that there is no NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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154 1 problem, but there should be discussion somewhere.
2 That's what I was thinking. Why is the discussion --
3 like, that not the part of this session? That's only 4 my question. I mean, I don't -- module drop impact on 5 other two modules, that's all part of the module 6 design. So I was thinking, those things will nicely 7 belong to this session to cover all the issues with 8 the multi-module design.
9 MR. RAD: So, when we wrote this 10 application, we invented this discussion. And we 11 decided to limit it to the scope given here and put 12 those sort of severe accidents in Chapter 19, which 13 were previously discussed. That's all. It was just 14 a matter of deciding which --
15 (Simultaneous speaking.)
16 CO-CHAIR DIMITRIJEVIC: All right, because 17 you know somebody would check this and say, okay, what 18 is the -- what type of issue I can see with multi-19 modules and the person saying multi-module design 20 consideration, and then there is -- not all 21 considerations are here. That's my point. So. I 22 mean, if you wanted to do, you should make it like 23 really to cover all the issues experienced with the --
24 analyzes -- I mean, only know ones from 19. Maybe 25 they're in the other chapters, also, bad issues -
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155 1 MR. RAD: It is already somewhat redundant 2 in that GDC-5 is discussed with each of the systems.
3 And so it's -- in this case, already redundant in a 4 lot of cases to a lot of the system descriptions. And 5 so it -- in some ways, it's just a summary of the 6 discussions and the system descriptions that are 7 provided throughout the FSAR today. So, I understand 8 what you're saying though. I understand the comment.
9 CO-CHAIR DIMITRIJEVIC: Especially when 10 you had a peer review that told you you guys are 11 making first step for the industry in the multi-12 module, so -- I know it's not easy to be first, but we 13 appreciate --
14 (Laughter.)
15 CO-CHAIR SUNSERI: Any other comments from 16 members?
17 MEMBER MARCH-LEUBA: Yes, just a comment 18 to Zach. When we go there in July -- to your place, 19 I will bring this issue of the -- of the temperature 20 after shutdown and boron concentration with you guys.
21 If you can be ready for it. I will ask it.
22 CO-CHAIR SUNSERI: I understand.
23 (Simultaneous speaking.)
24 MEMBER MARCH-LEUBA: I don't think the 25 answer we have today is --
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156 1 CO-CHAIR SUNSERI: Yes, and I think -- and 2 I -- just kind of being a -- listening from the side 3 here. I don't think you all are connecting. So maybe 4 if you could write down the question, so they clearly 5 --
6 (Simultaneous speaking.)
7 MR. RAD: I think it would be helpful for 8 us to answer rather than off the top of someone's head 9 -- to answer specifically, you know, given the 10 appropriate information right in front of us.
11 MEMBER MARCH-LEUBA: We were told that the 12 Core Operating Limits Report, the COLR, hasn't been 13 developed yet. And I suspect that the COLR will have 14 some boron concentration that you must achieve to have 15 such safe shutdown based on calculations you provide 16 in Chapter 4 and I would like to hear more about that 17 once we are there.
18 MR. RAD: That's great. Yes, I believe 19 the tech specs say in accordance with the Core 20 Operating Limits Report.
21 MEMBER MARCH-LEUBA: We don't have a --
22 you don't have a COLR. No, the tech spec is in 23 accordance with the COLR because it says thou shall 24 satisfy the COLR. The question is, what is in the 25 COLR?
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157 1 (Simultaneous speaking.)
2 CO-CHAIR SUNSERI: Let's coordinate 3 through Snodderly so that when we show up in --
4 (Simultaneous speaking.)
5 CO-CHAIR SUNSERI: -- they'll know what we 6 are talking about.
7 CO-CHAIR DIMITRIJEVIC: And you are going 8 to tell us a little more about this reactor or pooling 9 -- pool-cooling system, right? You said that they --
10 CO-CHAIR CORRADINI: We have it all --
11 it's all planned as part of the visit.
12 CO-CHAIR DIMITRIJEVIC: It's plan as a 13 visit -- it's not going to be discussed in closed 14 session or anything?
15 CO-CHAIR CORRADINI: What are we talking 16 about?
17 CO-CHAIR DIMITRIJEVIC: Reactor pool 18 cooling system, which I ask in the beginning of this.
19 CO-CHAIR CORRADINI: We can talk about it 20 in closed session if you would like to.
21 MS. JOERGENSEN: So I have -- I just have 22 one comment. So you asked if you lost site cooling, 23 what the impact on the -- on the UHS was.
24 CO-CHAIR DIMITRIJEVIC: Uh-huh.
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158 1 they said the pool temperature would rise very slowly 2 at a rate less than a degree per hour, which would 3 give the operator sufficient time to respond.
4 CO-CHAIR DIMITRIJEVIC: That's -- this is 5 not a safety system reactor pool --
6 (Simultaneous speaking.)
7 CO-CHAIR DIMITRIJEVIC: Raised one degree.
8 So 72 degrees for your mission time. Right?
9 Something to think about.
10 (Simultaneous speaking.)
11 CO-CHAIR DIMITRIJEVIC: So in accident 12 condition, that's not necessary, you're saying.
13 MR. MAXWELL: That's correct. It's not --
14 reactor pool cooling system is not required in 15 accident condition.
16 CO-CHAIR DIMITRIJEVIC: All right.
17 MR. MAXWELL: Also, just to point out that 18 as that -- as pool temperature escalated, there's tech 19 spec limits on ultimate heat sink temperature that 20 will direct the operators to conduct controlled 21 shutdowns, which they'd use the normal heat sink for 22 prior to reaching any safety limits.
23 CO-CHAIR DIMITRIJEVIC: What do you call 24 normal heat sink?
25 MR. MAXWELL: Condenser.
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159 1 CO-CHAIR DIMITRIJEVIC: Oh, condenser, 2 okay.
3 CO-CHAIR SUNSERI: Any other member 4 comments or questions?
5 (No audible response.)
6 CO-CHAIR SUNSERI: All right, at this 7 point we are going to recess for lunch. We will 8 reconvene at 1:00 and we will begin with the staff's 9 presentation on Chapter 21. There is about an hour-10 and-a-half of technical presentations remaining in the 11 agenda today, and then we will go into a public 12 comments session followed by the closed session. So 13 please plan appropriately for the after-lunch time.
14 Thank you. You're recessed.
15 (Whereupon, the above-entitled matter went 16 off the record at 11:54 a.m. and resumed at 1:00 p.m.)
17 CO-CHAIR SUNSERI: All right, 1:00, let's 18 reconvene the session here on Chapter 21, multi-module 19 design with a review by the staff. And I will turn it 20 over to Mr. Chowdhury.
21 MR. CHOWDHURY: Yes, good afternoon. My 22 name is Prosanta Chowdhury. I am the NRO project 23 manager for Chapter 21. I have been with the agency 24 for about 15 years -- 11 of which I have been working 25 in the NRO as a project manager. Prior to that I was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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160 1 employed by the State of Louisiana for their radiation 2 protection program involving nuclear power plants 3 surrounding the state, and I worked there for 18-and-4 a-half years.
5 I have a master's degree in electrical 6 engineering, and one in nuclear engineering. With 7 that, this is Chapter 21 of NuScale Design 8 Certification Application that the agency is reviewing 9 right now. And Chapter 21 is multi-module design 10 considerations. We have had the applicant's 11 presentation before lunch today.
12 Before I go into my slides, right next to 13 me is Marie Pohida, one of the key reviewers of multi-14 module interactions. I do have other staff present in 15 the audience and I will mention their names in one of 16 my slides and the specific questions related to those 17 areas that they have reviewed. They will be able to 18 answer those questions. I request them to come up to 19 the microphone and state their names and affiliation, 20 and then answer questions or make comments.
21 The technical staff that were involved in 22 reviewing aspects of Chapter 21, Multi-Module Design 23 Considerations, are Hanry Wagage, and I hope Hanry is 24 in the audience. If not, he will be here soon.
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161 1 in the back. Nan Chien and Chang Li -- Nan is there?
2 Okay. Sheila Ray, electrical. Angelo Stubbs, Ryan 3 Nolan and Michelle Hart and Marie Pohida is next to 4 me. Please don't ask me why Marie is sitting here and 5 nobody else.
6 CO-CHAIR CORRADINI: It says Chapter 19?
7 MR. CHOWDHURY: Chapter 19, yes. So, 8 anyway, the lead project manager is Greg Cranton, he 9 is sitting here. And he will be helping us answer any 10 questions there from the project management side. And 11 I am the chapter project manager. Next is -- I 12 believe you have seen the three-page safety evaluation 13 report on Chapter 21. And it is really not an 14 evaluation of Chapter 21 -- Applicant's Chapter 21 in 15 SER 20 -- Chapter 21 by the staff. It's -- it merely 16 points to where some of the shared systems have been 17 evaluated by the -- by the staff, and findings have 18 been documented. I would like to point out that 19 Chapter 15 has not been issued yet. But there are a 20 couple of sections in Chapter 15 the staff has 21 confirmed that they have addressed some of the aspects 22 of the design that's pertinent to Chapter 21.
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162 1 handled in those systems. And the failure of shared 2 systems that are not safety related is considered in 3 Chapter 15. Once again, Chapter 15 has not been 4 issued yet. Chapter -- Section 15.0.0 contains the 5 staff's review of categorization and classification of 6 design basis events. 15.0.3 is radiological 7 consequences of design basis accidents. That's why 8 the staff's review is documented. And then, 19.1.4.9, 9 Evaluation of Multimodule Risk, where staff discuss 10 the multimodule risk, including internal and external 11 events.
12 So the two concluding points, as far as 13 Chapter 21 SE goes, is that the staff used standard 14 review plan as well as prior applicable design-15 specific legal standards, DSRS, and then they used 16 these guidance to -- to conduct their review and 17 document their findings -- in those sections that I 18 have already mentioned. And a list of reactor names.
19 That's all the slides I have. And with that I will 20 open this for any questions from members of the 21 committee.
22 CO-CHAIR SUNSERI: So I have -- this is 23 going to be an easy one for you. So on page 3 -- can 24 you go back to page 3?
25 (Pause.)
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163 1 CO-CHAIR SUNSERI: This is -- I love the 2 table, actually. It clearly defines, you know, how 3 many systems, or essentially, modules. But I have a 4 question. So in a couple of cases you say six each 5 for two independent subsystems. Can you clarify that 6 nomenclature? Because in other cases you just say six 7 nuclear power modules. What do you mean by six each 8 for two independent subsystem? Like it is containment 9 flooding and drain number 5.
10 MR. CHOWDHURY: Yes. Okay, so I am --
11 Angela, did you -- or, Hanry Wagage is going to 12 explain the nomenclature there.
13 MR. WAGAGE: My name is Hanry Wagage.
14 What it means it that there are two different systems.
15 Each system has six units.
16 CO-CHAIR SUNSERI: Okay, so it's really 17 just -- term number three is the same? Is that --
18 MR. WAGAGE: Number three --
19 (Pause.)
20 MR. CHIEN: Nan Chien from NRC. There is 21 no information from NuScale, but my understanding is 22 one system for turbine. There is one turbine 23 building. There is only one turbine building.
24 CO-CHAIR CORRADINI: I think the question 25 we're asking is, is the interpretation of six each of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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164 1 two independent subsystems the same as what you say 2 under number 3?
3 MR. CHIEN: No, six 4 CO-CHAIR SUNSERI: So six should be 12 for 5 number 3. Okay.
6 (Simultaneous speaking.)
7 MR. CHOWDHURY: But there are two turbine 8 buildings.
9 CO-CHAIR DIMITRIJEVIC: So it's the same 10 HVAC system.
11 CO-CHAIR SUNSERI: Oh, yes. Right. Okay.
12 MR. CHOWDHURY: There are two turbine 13 buildings.
14 CO-CHAIR DIMITRIJEVIC: So there are two 15 independent systems, right? Turbine HVAC systems?
16 PARTICIPANT: Yes.
17 MR. CRANSTON: This is Gary Cranston.
18 There are two turbine buildings. Each turbine 19 building serves six modules.
20 CO-CHAIR DIMITRIJEVIC: And has 21 independent HVAC system.
22 MR. CRANSTON: They're independent. The 23 two different buildings are independent -- one on each 24 side of the reactor building.
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165 1 question.
2 CO-CHAIR SUNSERI: No, no, I understand.
3 I understand that, though.
4 MR. MAXWELL: Number 3 should be six each 5 for two independent subsystems as well.
6 CO-CHAIR SUNSERI: Okay, but I understand 7 the distinction they're trying to make now. Just all 8 the rest of it is in one building. This happens to be 9 two, right? Okay. I don't have any other questions.
10 CO-CHAIR DIMITRIJEVIC: I have a question.
11 It's not directly maybe connected to this, but I just 12 noticed that we never discussed this control room 13 habitability system. Then was that -- this loss of 14 that which will result in that control room 15 evacuation?
16 (Pause.)
17 MR. CHOWDHURY: Hanry?
18 CO-CHAIR DIMITRIJEVIC: And how was that 19 treated? Would -- obviously the operators would trip 20 before they evacuate control room, right? Even they 21 may not be needed.
22 MR. CHOWDHURY: I am looking at the 23 Chapters 6. Staff.
24 CO-CHAIR CORRADINI: You're asking 25 NuScale?
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166 1 CO-CHAIR DIMITRIJEVIC: Well, somebody.
2 (Pause.)
3 MR. MAXWELL: Again, Chris Maxwell from 4 NuScale. A note, a loss of control room habitability 5 system wouldn't require evacuation of the control 6 room. It's rather a loss of all electrical power or 7 a event -- toxic gas event -- would initiate control 8 room habitability system, compressed air, pressurizing 9 the control room to maintain the -- the environment 10 within the control room suitable for habitation.
11 CO-CHAIR DIMITRIJEVIC: So loss of that 12 system would require -- is that what you say?
13 MR. MAXWELL: No, not -- normal operating 14 conditions, if the control room habitability system 15 was rendered inoperable, then normal control room HVAC 16 system would maintain humidity and temperature in the 17 control room for the operators.
18 CO-CHAIR DIMITRIJEVIC: Okay. Which is 19 number 6 versus number 8, if that is -- there are two 20 redundant system, you say.
21 MEMBER BLEY: Well, no the -- the --
22 number 6 one is -- only comes into play if you have a 23 problem with the normal HVAC, or you have contaminants 24 come in.
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167 1 normal operating system. Number 8 --
2 (Simultaneous speaking.)
3 MEMBER BLEY: Oh, I am sorry. Yes.
4 MR. MAXWELL: Eight comes into play if we 5 have toxic gas --
6 (Simultaneous speaking.)
7 CO-CHAIR DIMITRIJEVIC: Toxic releases.
8 Okay, so let's discuss number 6. If they lost 9 ventilation, would that require evacuation of the 10 control room?
11 MR. MAXWELL: No, it would not require a 12 control room evacuation.
13 CO-CHAIR CORRADINI: I think what he is 14 saying is, 6 is normal operation. If that is 15 compromised, then 8 would be activated to keep the 16 control room -
17 CO-CHAIR DIMITRIJEVIC: I thought 8 is 18 only if they have a toxic releases somewhere.
19 MR. MAXWELL: Normally the control 20 habitability system -- number 8 is in a stand-by 21 lineup with isolation valves presenting the compressed 22 air from being released into the control room. So the 23 normal control room HVAC system maintains the normal 24 environment. It is only upon that loss of power, 25 which renders the normal control room HVAC system NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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168 1 inoperable -- or, on a toxic gas or fire -- that the 2 valves in the control room habitability system then 3 open and supply the compressed air into the control 4 room.
5 MEMBER BLEY: How long does that last?
6 MR. MAXWELL: A minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
7 MEMBER BLEY: That's a lot.
8 (Simultaneous speaking.)
9 CO-CHAIR SUNSERI: Okay, good. Anybody 10 else?
11 (No audible response.)
12 CO-CHAIR SUNSERI: All right, well thank 13 you for the presentation.
14 PARTICIPANT: Thank you.
15 CO-CHAIR SUNSERI: Now we are going to 16 transition back to NuScale where we will discuss 17 Chapter 19.5, Adequacy as Design Features in 18 Functional Capability Identified and Described for 19 Withstanding Aircraft Impacts. Open session.
20 (Pause.)
21 MR. BRYAN: Good afternoon. My name is 22 Marty Bryan. I am the Licensing Project Manager for 23 NuScale for 6 and 19.5, and today we are going to 24 discuss adequacy of design features and functional 25 capability identified and described for withstanding NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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169 1 aircraft impacts. Our presenter will be Amber Berger.
2 MS. BERGER: Hello, I am Amber Berger, a 3 civil structural engineer with NuScale Power.
4 CO-CHAIR SUNSERI: Could you pull that 5 just a little bit closer to you?
6 MS. BERGER: Sure. So quick introduction 7 and a background of what we will present today. So 8 the NuScale Power Plant has been designed for the 9 potential effects of a beyond-design basis large 10 commercial aircraft impact, following the guidance of 11 10 CFR -- Title 10 CFR 50, 150-A, which requires that 12 the reactor core remains cooled, or the containment 13 remains intact -- and that the spent fuel pool -- or, 14 the spent fuel remains cool, or that the spent fool 15 integrity is maintained.
16 We have followed design-specific 17 assessment requirements of Reg Guide 1.217, which is 18 the guidance for beyond-design basis aircraft impact.
19 And that endorses NEI 07-13, which is the methodology 20 for aircraft impact assessment for a new plant design.
21 We have followed NEI 07 guidance, taking no exceptions 22 from the methodologies. And just as a note, the 23 aircraft impact has actually informed the design of 24 our plant, such as requiring five-foot thick walls for 25 the exterior concrete walls of the reactor building.
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170 1 The scope of our assessment --
2 CO-CHAIR CORRADINI: So just to make 3 sure, the outside reactor walls are treated as 4 potentially a -- the equivalent of a missile shield, 5 as a robust barrier?
6 MS. BERGER: Yes.
7 CO-CHAIR CORRADINI: Okay. Thank you.
8 MS. BERGER: The reactor building has been 9 assessed for three different areas of postulated 10 aircraft impact. One is the physical damage of the 11 airplane actually hitting the reactor building, 12 exterior walls, and the roof. We also assessed for 13 shock damage, which is the shock-induced vibrations on 14 the structure systems and components housed within the 15 reactor building, and we've assessed for fire damage, 16 which is the aviation fed -- fuel-fed fire from 17 aircraft impact.
18 The methodology, as I mentioned, is taken 19 from NEI 07-13. The reactor building is our structure 20 of concern, so the reactor building is what we've 21 assessed for the aircraft impact. The reactor 22 building houses nuclear power modules, the NuScale 23 power modules, houses the ultimate heat sink and it 24 houses the spent fuel pool. So, all of these systems 25 inside the reactor building are what we're trying to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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171 1 protect.
2 The impact locations of the aircraft 3 hitting the reactor building were screened based on 4 NEI 07 guidance. The rad waste building, which is a 5 building directly to the west of the reactor building, 6 is considered as an intervening structure for the 7 structural portion of our assessment. We did not --
8 conservatively, we did not consider it an intervening 9 structure for the shock or the fire assessment. No 10 credit is taken for the control building or the 11 turbine generator buildings, which, on the next slide, 12 we show a schematic view of our site plant, showing 13 you the -- a site layout.
14 The building right in the middle, the 15 long, rectangular gray building, that's the reactor 16 building. That's the building of concern for us --
17 the structure of concern. To the west of that -- so, 18 north is up and to the left. The building -- the 19 small gray building low and to the left of the reactor 20 building is the rad waste building. It's a heavily 21 reinforced concrete building directly adjacent, about 22 30 feet from the reactor building. That's the 23 building that we've used as an intervening structure 24 for a structural assessment.
25 MEMBER BLEY: Can you -- whoever's got the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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172 1 mouse, can you point at that one?
2 MEMBER MARCH-LEUBA: Yeah, you --
3 MS. BERGER: Yeah.
4 PARTICIPANT: It's the lower gray box.
5 MEMBER MARCH-LEUBA: Amber, you can get 6 the mouse and point.
7 MS. BERGER: Okay.
8 MEMBER MARCH-LEUBA: It's long enough.
9 MS. BERGER: Sure. That's great.
10 PARTICIPANT: Can you see it there?
11 MEMBER MARCH-LEUBA: Yeah.
12 MS. BERGER: Okay. Thank you. Okay. So, 13 I'll start from the beginning. The reactor building 14 is the long, skinny gray building.
15 MEMBER BLEY: You said that was heavily 16 reinforced. The same as the reactor building, or?
17 MS. BERGER: Not quite as heavily 18 reinforced, no.
19 MEMBER BLEY: How thick's the --
20 MS. BERGER: The control building walls 21 are three-foot thick.
22 MEMBER BLEY: Three foot?
23 MS. BERGER: Yeah. With two layers of 24 number 11, approximately. No shear tie reinforcing.
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173 1 the roofs?
2 MS. BERGER: Not for aircraft impact, no.
3 MEMBER BLEY: Because the angle has got to 4 be two --
5 MS. BERGER: Yes. Yep.
6 (Simultaneous speaking.)
7 CO-CHAIR CORRADINI: Could you say that 8 louder? I didn't understand.
9 MEMBER BLEY: I asked about the roof and 10 I asked why it didn't have to be considered. And the 11 answer was because the angle is going to be a glancing 12 blow.
13 CO-CHAIR CORRADINI: So, that -- the 14 requirement is the roof does not need to be fortified 15 in a similar manner as the walls?
16 MS. BERGER: Yes. Flat roofs.
17 MEMBER BROWN: That's the control room 18 building, right? You're talking about the control 19 room building right now?
20 MS. BERGER: Yeah. I believe so. The 21 flat roofs --
22 MEMBER BLEY: Well, I was first going to 23 talk about that, then I was going to ask about the 24 reactor building that's --
25 MS. BERGER: Got it. Yes.
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174 1 MEMBER BLEY: -- on the roof.
2 MS. BERGER: Yep.
3 MEMBER BLEY: Same thing?
4 MS. BERGER: Well, so, on the reactor 5 building, we do have some sloped portions of the roof 6 that are not completely horizontal. And we did assess 7 the sloped portions of the roof for aircraft impact.
8 MEMBER BLEY: Okay. How thick's the roof 9 on the reactor building?
10 MS. BERGER: It's four-feet thick.
11 MEMBER BLEY: Okay. So, that's pretty 12 hefty. Okay.
13 CO-CHAIR CORRADINI: So, this is an impact 14 barrier, but yet not pressure-retaining.
15 MS. BERGER: Pressure-retaining for what 16 type of --
17 CO-CHAIR CORRADINI: Internal pressure.
18 It's a leaky, strong building.
19 MS. BERGER: Well, we prevent fire from 20 getting into the reactor building -- any pressurized 21 fire from getting into the reactor building -- the 22 main area of the reactor building.
23 CO-CHAIR CORRADINI: Okay.
24 MS. BERGER: And so, it doesn't need to be 25 pressure-retaining for an aircraft impact perspective.
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175 1 CO-CHAIR CORRADINI: Okay. Because the 2 reason I'm asking like that is the next question is, 3 if it's pressure-retaining, does it have a leak rate 4 specification?
5 MS. BERGER: We don't need it to be 6 pressure-retaining.
7 CO-CHAIR CORRADINI: Okay.
8 MS. BERGER: Yeah.
9 CO-CHAIR CORRADINI: Okay.
10 MS. BERGER: So it's not.
11 CO-CHAIR CORRADINI: So, how do you prove 12 that what's outside can't get inside?
13 MS. BERGER: This is sort of later on in 14 the slides. We can start right now, if you'd like.
15 CO-CHAIR CORRADINI: That's fine. We'll 16 wait. No, no. You can tell us to wait.
17 MS. BERGER: Okay. So I'll continue --
18 actually, go back to the previous slide, please. Just 19 to finish out this slide, the turbine generator 20 buildings are the long, larger brown buildings north 21 and south of the reactor building. Those will be 22 steel buildings. We did not credit those as 23 intervening structures, so we ignored those in our 24 assessment. And also, to the --
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176 1 directly in from the sides there and hit the reactor 2 building.
3 MS. BERGER: Yes. Yep.
4 MEMBER BLEY: You took no credit for any -
5 -
6 MEMBER BROWN: What did she say?
7 (Simultaneous speaking.)
8 MEMBER BROWN: -- buildings.
9 CO-CHAIR CORRADINI: The yellow buildings 10 aren't credited.
11 MEMBER BLEY: It's like they aren't there.
12 MEMBER BROWN: Okay. For aircraft, it's -
13 - I got it. Thank you.
14 MS. BERGER: Got it.
15 CO-CHAIR SUNSERI: So, let me just pause 16 for just a second. I don't know where the line 17 between sensitive information and non-sensitive 18 information is on this topic is, but don't let us ask 19 you questions to expose sensitive information and --
20 MS. BERGER: Appreciate it. Thank you.
21 CO-CHAIR SUNSERI: Okay. Thank you.
22 CO-CHAIR CORRADINI: But we leave it to 23 you to tell us that it's not appropriate.
24 MS. BERGER: Okay. Will do. Thanks.
25 CO-CHAIR CORRADINI: So, now I have a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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177 1 question.
2 MS. BERGER: Okay.
3 CO-CHAIR CORRADINI: So, because of the 4 low profile of the design, does that essentially 5 affect how Dennis was asking the question about angle 6 of attack, or angle of --
7 (Simultaneous speaking.)
8 MS. BERGER: The low profile of the 9 reactor building? The height of the reactor building, 10 essentially?
11 PARTICIPANT: Smaller target.
12 MS. BERGER: Well, we've assessed the 13 walls of the reactor building for aircraft impact 14 strikes along all four sides of the reactor building.
15 16 CO-CHAIR CORRADINI: But the fact that 17 it's a low profile, you don't have to worry about it 18 coming in, in unusual orientations.
19 MS. BERGER: No. We follow the NEI 20 guidance.
21 CO-CHAIR CORRADINI: Okay.
22 MS. BERGER: So --
23 CO-CHAIR CORRADINI: Okay.
24 MS. BERGER: Yeah.
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178 1 it this way. Does the NEI guidance have a height of 2 the building aspect too, so that if I have a low 3 height I have to follow one set of procedures for 4 analysis versus a higher height?
5 MS. BERGER: There's no specific 6 stipulation in the guidance for that.
7 CO-CHAIR CORRADINI: Okay. I couldn't 8 remember.
9 MS. BERGER: We've assessed the 10 building above grade. All of the building above grade 11 is susceptible to aircraft impact.
12 CO-CHAIR CORRADINI: Okay. Thank you.
13 MS. BERGER: Yeah.
14 MEMBER MARCH-LEUBA: And as a mechanical 15 engineer, would have been cheaper to dig 20 foot 16 deeper and not have any vertical walls?
17 MS. BERGER: Yeah. We looked into that, 18 actually. We did an aircraft impact design decision 19 paper and we considered that.
20 MEMBER MARCH-LEUBA: If there is no 21 vertical walls, you don't have to do the assessment, 22 right?
23 MS. BERGER: Sure. Yes. But then your 24 entire building is low grade. Yeah. There's 25 advantages and disadvantages, and we did do those.
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179 1 MEMBER MARCH-LEUBA: And this is cheaper, 2 4-foot reinforced concrete?
3 MS. BERGER: Yes.
4 MEMBER MARCH-LEUBA: I thought digging was 5 cheap.
6 PARTICIPANT: No.
7 MS. BERGER: It's the operational impacts 8 that are more expensive. Yeah.
9 MEMBER BLEY: And just so you don't 10 forget, we covered the control building --
11 MS. BERGER: Yep.
12 MEMBER BLEY: -- the two turbine halls, 13 and the you got two more buildings at the back end, 14 there.
15 MS. BERGER: Well, so, actually, the 16 control building is this brown building that's sort of 17 behind the reactor building that's hard to see in this 18 graphic. The control building, again, is -- we did 19 not use it as an intervening structure.
20 MEMBER BLEY: Okay.
21 MS. BERGER: Just pointing that out. And 22 then adjacent to that's the central utility building, 23 not considered in our analysis.
24 MEMBER BLEY: Oh, that's not considered 25 either.
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180 1 MS. BERGER: Yeah. Yeah. So, really, 2 only -- we only considered intervening structure on 3 one side of the building, that's the red waste 4 building here on the west, and we only considered that 5 for the structural assessment. Next slide.
6 So, the assessment results for physical 7 damage, we assessed local damage of the aircraft 8 hitting the exterior walls of the reactor building and 9 the sloped portion of the roof and found there's no 10 perforation of the aircraft and no scabbing effects 11 that would cause --
12 MEMBER BLEY: How detailed an analysis did 13 you do on that?
14 MS. BERGER: For structures?
15 MEMBER BLEY: Yeah.
16 MS. BERGER: Let's see here. Going down.
17 Is it in this slide? Oh, second -- next slide. We 18 did a detailed finite element analysis model, 19 considered --
20 MEMBER BLEY: So not even the rotors get 21 through the building.
22 MS. BERGER: No.
23 MEMBER BLEY: And you were right at 24 scabbing, so don't forget that. Tell us about that.
25 MS. BERGER: Talk about the scabbing?
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181 1 MEMBER BLEY: Yeah.
2 MS. BERGER: We had done an assessment for 3 scabbing and found that it doesn't cause an issue to 4 meet the 10 C.F.R. guidance or requirements. So, that 5 was the local assessment. We also did the global 6 assessments using our detailed finite element model, 7 using -- and using the NRC-specific force-time 8 history.
9 The reactor building external walls 10 prevent physical damage from entering the reactor 11 building, so we completely keep the plan out of the 12 building proper. There are no internal missiles for 13 secondary impact, which would be a scabbing issue. We 14 have no impact to the containment boundary or the 15 containments' boundaries. Our spent fuel --
16 MEMBER BLEY: There's nothing on any walls 17 that could get knocked loose or be sufficient?
18 MS. BERGER: That's part of the shock 19 assessment.
20 MEMBER BLEY: Okay. I'll wait for that.
21 MS. BERGER: Thank you. And our spent 22 fuel pool is protected as it's inside the reactor 23 building and it's completely below grade, so there's 24 no issues -- physical damage to the spent fuel pool.
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182 1 cannot be dislodged from its rails.
2 MEMBER SKILLMAN: Is that because of the 3 seismic clamps that hold that bridge and trolley in 4 place?
5 MS. BERGER: Not entirely. It's really 6 just the -- we've done -- we did an analysis to 7 determine what the displacement on the rails would be 8 from the aircraft hitting the exterior of the reactor 9 building. And that displacement geometrically is less 10 than the length of the bridge rails.
11 MEMBER SKILLMAN: Thank you.
12 MS. BERGER: Yeah. And pertaining to the 13 shock damage, so the aircraft impact causes short-14 duration, high-acceleration, high-frequency 15 vibrations, and this would be the sort of damage that 16 would cause equipment to malfunction inside the 17 reactor building, cause equipment to fall off walls.
18 For the core cooling, we do an at-power 19 and shutdown scenarios. That was considered in our 20 assessment. For our plant, there's no active 21 equipment that's required for -- to successfully keep 22 the cores cool.
23 MEMBER BLEY: But would some of the 24 equipment be damages?
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183 1 strikes around the reactor building. There is one 2 strike on the west side of the building where the rad 3 waste building is. Taking no credit for the rad waste 4 building as in intervening structure, some of the 5 shock from that impact does reach into the room where 6 the core cooling equipment is on the 50-foot 7 elevation, two floors below grade. And when the shock 8 footprint touches on that equipment, we assume that it 9 loses its function.
10 MEMBER BLEY: Okay.
11 MS. BERGER: It loses its ability to 12 perform that function. And then the general jolt for 13 our core cooling is the adequate heat removal shown 14 for all strikes. The cores are -- remain cool based 15 on our passive designs.
16 MEMBER SKILLMAN: For that portion of the 17 analysis, is the equipment so fragile that, even with 18 that shock loading, it would fail? Or do you just 19 simply assume that it fails?
20 MS. BERGER: We assume -- it's part of the 21 guidance you assume that it fails. There's two levels 22 of shock. There's near shock, which is -- mechanical 23 is more susceptible to the near shock. And if near 24 shock -- or, sorry. Vice versa.
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184 1 phone line that is open. Could you please mute your 2 phone line?
3 MEMBER SKILLMAN: You were saying, two 4 levels of shock?
5 MS. BERGER: Yeah. Two levels of shock, 6 and I said it wrong the first time. I apologize. The 7 far shock is effective to mechanical equipment. If 8 the far shock profile has the potential to reach that 9 equipment, the assumption in the guidance is you lose 10 the capacity of -- or, the capabilities of that 11 system.
12 MEMBER SKILLMAN: Thank you.
13 MS. BERGER: And then, additionally, spent 14 fuel for the shock damage, the spent fuel pool 15 integrity is maintained for all strikes. The shock 16 profiles don't actually even reach the spent fuel 17 pool. And next slide.
18 So, fire damage. The results for the fire 19 damage assessment, we've designed and located three-20 hour fire barriers and three-hour 5 psi fire barriers 21 to prevent the propagation of fire into the reactor 22 building. We've also designed and located 5 psi fast 23 acting dampers inside of the reactor building HVAC air 24 intakes where the louvers are on the exterior of the 25 building. Those prevent those -- blast dampers NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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185 1 prevent pressurized fire from getting into the reactor 2 building.
3 MEMBER BLEY: What activates them?
4 MS. BERGER: Pardon?
5 MEMBER BLEY: What activates them?
6 MS. BERGER: The pressure.
7 MEMBER BLEY: The pressure from the blast?
8 Okay.
9 MS. BERGER: Yes. Yeah. We also have on 10 the exterior of the building, at all exterior wall 11 penetrations, we've placed concrete shrouds -- five-12 foot thick, heavily-reinforced concrete shrouds, to 13 protect pipe penetrations, HVAC penetrations, prevents 14 physical damage from getting inside those -- into 15 those -- through those penetrations, and it also 16 prevents fire from propagating into the reactor 17 building.
18 MEMBER BLEY: I don't remember. At each 19 of the turbine halls, the main steam lines, do they go 20 underground and then come up in the building?
21 MS. BERGER: They don't. They come out 22 the level at-grade, and then they go straight across 23 that grade, yes. Well, they don't go straight across.
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186 1 that we have located.
2 MEMBER BLEY: Okay. So they could be 3 ripped up by this thing.
4 MS. BERGER: The main seam lines?
5 MEMBER BLEY: Yeah.
6 MS. BERGER: Yeah. Past the shrouds.
7 Past the exterior of the reactor building, yeah. And 8 then the last bullet for the fire damage assessment 9 is, on that -- on all four corners of the reactor 10 building, we have personnel doors. Mostly are for 11 emergency egress. Fire can get inside of those 12 personnel doors.
13 But then we have stairwells at each door 14 and five -- three-hour fire barriers so that 15 unpressurized fire can get into stairwells. It 16 propagates up and down, vertically, through the 17 building, but it stays contained within the stairwell, 18 so it never gets into the reactor building main area.
19 And so, we keep fire out of the fuel -- the spent fuel 20 pool area and the modules area -- the operating module 21 bay.
22 And then another part of our assessment is 23 that required -- operator action is required prior to 24 the impact of an aircraft, and that is part of the 25 methodology of any --
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187 1 MEMBER BLEY: You must have some kind of 2 equipment hatches somewhere for letting -- bringing in 3 new equipment.
4 MS. BERGER: We have the west end of the 5 building and adjacent -- between the reactor building 6 and rad waste building, we have a barrier, 7 approximately, 20-foot-by-20-foot door that has a door 8 that is made of -- it's steel-concrete composite and 9 it's been designed for the physical damage of an 10 aircraft impact, and it will be sealed to prevent 11 pressurized fire from entering the reactor building, 12 too. But that's the largest -- other than the 13 personnel doors --
14 MEMBER BLEY: What's the longest it sits 15 open if you're moving things in and out?
16 MS. BERGER: I don't know the answer to 17 that. I'm sorry --
18 MEMBER BLEY: So, it could be a 19 substantial period of vulnerability.
20 MS. BERGER: No. I wouldn't expect so 21 because it's -- that door also separates the rad 22 boundary, right? So, we don't want that door to be 23 open -- it would not be open for normal operations or 24 anything. We need that --
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188 1 equipment and you shut down all the reactors before 2 you brought in --
3 MR. RAD: So, there's a period of time 4 between when the plant is warned of an impending 5 impact that operator actions occur and the plant is 6 placed in certain conditions. And we would expect 7 that this would be one of the actions that would be 8 taken, would be to close this door.
9 (Simultaneous speaking.)
10 MEMBER MARCH-LEUBA: I'm confused. Who 11 warns you? How do you know it's coming?
12 MR. RAD: They.
13 MEMBER MARCH-LEUBA: Who warns you that 14 an --
15 MR. RAD: Okay. Right. So, that's a 16 matter of plant operations, right? So, the Federal 17 Government warns you that there's an impending 18 aircraft impact. But that's --
19 PARTICIPANT: Like the FAA?
20 MR. RAD: Yeah. Yeah.
21 MEMBER MARCH-LEUBA: The airport tower?
22 CO-CHAIR SUNSERI: There's a protocol.
23 There's an industry protocol.
24 PARTICIPANT: Yeah.
25 MEMBER BLEY: Unless you're the first one.
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189 1 Go ahead.
2 MS. BERGER: Okay. Next slide. In 3 conclusion, the design and functional capabilities of 4 our plant provide adequate protection to public health 5 and safety. NuScale Plant meets 10 C.F.R. 5150 6 regulations by both maintaining containment integrity 7 and maintaining core cooling capability. Only one of 8 these is required to meet the guidance. We also 9 maintain spent fuel pool integrity.
10 And then, for most postulated aircraft 11 impact strikes, spend fuel pool -- active spend fuel 12 pool cooling is maintained, meeting all four of the 10 13 C.F.R. requirements.
14 MEMBER BLEY: Which implies there's some 15 of these strikes that wipe out your cooling and fuel 16 pool.
17 MS. BERGER: Just that one strike. One 18 out of 17 strikes affects the spent fuel pool cooling 19 equipment.
20 MR. CHOWDHURY: From shock that we don't 21 take credit for that equipment.
22 MEMBER BLEY: Okay.
23 MS. BERGER: Yeah.
24 MEMBER BLEY: But you would still have the 25 main pool cooling --
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190 1 MS. BERGER: Still have -- well, we still 2 have our ultimate heat sink and we still have --
3 MEMBER BLEY: That's what I meant.
4 MS. BERGER: Yeah. Yeah. And that is the 5 rest of our slides and acronyms. Appreciate any 6 questions or clarifications.
7 CO-CHAIR SUNSERI: Members, any additional 8 questions?
9 MS. BERGER: Great.
10 CO-CHAIR SUNSERI: Thank you.
11 MS. BERGER: Appreciate it. Thank you.
12 CO-CHAIR SUNSERI: We'll transition for 13 the staff to come up with their safety evaluations.
14 Whenever you're ready.
15 MR. CRANSTON: Good afternoon. I'm Greg -
16 -
17 CO-CHAIR SUNSERI: Your microphone.
18 MR. CRANSTON: I think I blew mine out.
19 CO-CHAIR SUNSERI: Hey, I mess it up.
20 MR. CRANSTON: Good afternoon. I'm Greg 21 Cranston, project manager for the NuScale project.
22 We're here for the continuation of Chapter 19 with 23 19.5, and Ryan is going to make the presentation.
24 MR. NOLAN: All right. Thanks, Greg. My 25 name's Ryan Nolan. I'll be giving the staff's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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191 1 presentation on our Safety Evaluation Section 19.5, 2 Aircraft Impact Assessment. You can go to the next 3 slide. The 51.50 requires a design-specific aircraft 4 impact assessment and incorporation of design features 5 to ensure that the reactor core remains cool or 6 containment remains intact, and spend fuel pool 7 cooling or spent fuel pool integrity is maintained.
8 The introduction of the Safety Evaluation 9 tries to step through what it is we do in the 10 licensing review and what we're making a finding on.
11 And in this case, we're looking at -- that the 12 Applicant identified design features, described design 13 features, and stated how those design features meet 14 the assessment. The assessment is not submitted to 15 us. We don't review the assessment or the adequacy of 16 the design features as part of our licensing review.
17 We just ensure -- it's more of a 51.50 Bravo is the 18 content of application. That's what our licensing 19 review is on in 19.5. And so --
20 CO-CHAIR CORRADINI: So you're reviewing 21 that they followed procedures.
22 MR. NOLAN: We review to ensure that 23 they've identified the design features in the 24 licensing basis document in 19.5, that they've 25 described those design features.
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192 1 CO-CHAIR CORRADINI: All right.
2 MR. NOLAN: So, as far as the design 3 features, they're typically broken down into three 4 areas: fire protection, structural, and systems. For 5 example, fire protection, they would identify -- they 6 do identify which three-hour walls, 5-psi barriers are 7 credited as design features.
8 The structure, I think the main one here 9 is the exterior reactor building walls, that they're 10 identified as key design features. You know, 11 dimensions, rebar size, things like that. And then as 12 far as systems, they would identify any of the systems 13 that the assessment relies on to meet the 14 requirements. That would be decay heat removal 15 system, ability to trip the reactor, things like that.
16 As far as how they meet the requirements, 17 you know, we ensure that the systems or SSCs that were 18 identified as key design features, that they are 19 linked to the requirements. And in this case, NuScale 20 has committed to meeting the requirements by ensuring 21 the reactor core remains cooled and containment 22 remains intact, and then spent fuel pool integrity is 23 maintained. We review to ensure that the assessment 24 looked at shock damage, that they followed the 25 appropriate guidance, and that the assessment was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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193 1 performed by qualified individuals.
2 There's one confirmatory item, and this 3 was more of just a clarification on the use of 4 instrumentation and monitoring, and whether it was a 5 key design feature, or if it was just there as a 6 monitoring aspect of it. And so, this is just to 7 identify that there is a confirmatory item. We 8 reviewed the RAI. We found it acceptable. And so, 9 we'll close it out when we see your Provision 3 of the 10 FSAR.
11 CO-CHAIR CORRADINI: I don't think I 12 understand the RAI. Can you help me again?
13 MR. NOLAN: Sure. A lot of our Phase 1 14 RAIs were more clarification on the description of 15 systems and are they considered a key design feature 16 or is it kind of just there as part of normal 17 operations. And the clarification that we received 18 from NuScale is that the decay heat removal system 19 doesn't rely on instrumentation. So, it's a clear, 20 passive system. Doesn't rely on any external power.
21 The instrumentation is there just for monitoring.
22 CO-CHAIR CORRADINI: Okay. Thank you.
23 MR. NOLAN: It's not needed to meet the 24 core cooling requirement.
25 CO-CHAIR CORRADINI: Okay. I got it.
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194 1 MR. NOLAN: So, in conclusion, we found 2 that the Applicant adequately identified and described 3 the design features and that it's reasonably 4 formulated and performed by qualified individuals.
5 CO-CHAIR SUNSERI: That it?
6 MR. NOLAN: Yeah.
7 CO-CHAIR SUNSERI: Members, questions?
8 All right. Well, thank you very much for the 9 presentation. At this point, we have completed all 10 the prepared presentations -- or, we've gotten through 11 all the prepared presentations, so now I'll open up 12 the floor to public comments. We'll first start with 13 the members of the public that are in the room. And 14 while we are asking for their comments, we will also 15 open the phone line.
16 So, any members of the public in the room 17 that would care to make a statement? Okay. We'll 18 turn to the phone line now. Any members of the public 19 on the phone line that would care to make a statement?
20 Now is the opportunity. And if there is a member out 21 there, they could just acknowledge that the phone line 22 was open, we would appreciate that.
23 PARTICIPANT: Phone line is open.
24 CO-CHAIR SUNSERI: So, no comments to be 25 had?
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195 1 MR. SNODDERLY: If there are no comments, 2 we're going to close the public line at this time.
3 CO-CHAIR SUNSERI: Okay. All right. So, 4 the public comments are done.
5 Subcommittee, let's go around the table 6 now for the open portion of this meeting and hear from 7 individual members on any comments you'd like to make.
8 We'll start with Vesna.
9 CO-CHAIR DIMITRIJEVIC: I don't have any 10 additional comments now.
11 CO-CHAIR SUNSERI: Dennis?
12 MEMBER BLEY: Yeah. I'm still -- just 13 continuing from yesterday, but, you know, the big 14 things that are sitting there for me area the ECCS 15 valves and the passive system performance. Excuse me.
16 I've heard the phrase, investment protection issue, 17 from a number of people. And just a caution, you 18 know? We've seen a lot of times when these things we 19 call investment protection have led to events that 20 have occurred, they quickly morph into safety concerns 21 and everybody gets spun up.
22 The other thing is, if I'm -- I sell you 23 a plant and we scuff up -- we get the plant in a 24 situation that makes you shut down all your reactors, 25 the executives of that utility aren't going to be very NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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196 1 happy with their vendor friend. So, thinking hard 2 about those things to make sure they're very unlikely, 3 yeah, makes -- probably worthwhile.
4 MEMBER MARCH-LEUBA: No further comments.
5 CO-CHAIR SUNSERI: Charlie?
6 MEMBER BROWN: No further comments other 7 than what I had earlier.
8 CO-CHAIR SUNSERI: Okay. Joy?
9 MEMBER REMPE: So, I -- again, I believe 10 today we talked about a concern about exactly which 11 shared systems failures could require modules to be 12 shut down because they've lost that share system. And 13 I believe the Applicant did agree to look into whether 14 they had to provide such a list and whether it is part 15 of the Tier 2 required documents.
16 And so, I'm interested in seeing that 17 answer. And maybe -- again, if it's not required, 18 maybe that's something where we have a gap and it 19 should be required, and it's something for us to think 20 about, so I'd like to have that follow-up.
21 And then, I think the comments I had 22 yesterday, one of them got addressed. But the 23 question I had about a list of what other 24 instrumentation would give the operators an indicator 25 that -- or, would we, not the operators, excuse me, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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197 1 would cause the module protection system to generate 2 a signal if the -- if you didn't have the pressurized 3 water level, and I still would like to hear that 4 answer, too. And so, I think that's it.
5 MEMBER BROWN: Matt, when you're done.
6 CO-CHAIR SUNSERI: Thank you. I was just 7 taking a note here --
8 (Simultaneous speaking.)
9 MEMBER BROWN: Yeah. I just want to 10 backtrack for a minute. I'm not sure how I covered 11 this during the multi-module design consideration 12 relative to integration of and the lack of an 13 architecture in anything that showed that. One of the 14 things -- or, two of the things, actually, that I kind 15 of expected when I ready Chapter 21 was a discussion 16 of how abhorrent behavior or abnormal behavior, such 17 as interlocks that would be to ensure certain things.
18 You couldn't start the system in another 19 plant without turning it off in the previous --
20 whatever's required, as well as an annunciations 21 and/or alarms or warnings that you might say is 22 something inadvertently energizes itself in another 23 one of the modules. And that's my concern on the 24 whole chapter is that it's sparse on how those shared 25 systems are integrated into the overall control and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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198 1 operations in the overall facility in the plant 2 instead of just operators are going to control 3 everything and it's going to be just fine. I have --
4 very skeptical on that point.
5 That's all. I just wanted to amplify 6 that -- that's what I meant by architectures. How 7 they were integrated, how you interlock them, what 8 warnings and indications and monitoring you provide 9 that give operators a heads up that, hey, they're 10 heading into some situation or condition that they're 11 not aware of.
12 CO-CHAIR SUNSERI: Okay.
13 MEMBER BROWN: That's all I have.
14 CO-CHAIR SUNSERI: Yep. Thank you, 15 Charlie. Mike?
16 CO-CHAIR CORRADINI: I want to ask 17 Charlie. So, you're saying it doesn't exist or you 18 couldn't find it in that chapter?
19 MEMBER BROWN: It doesn't exist. He said 20 there's no diagram. And if you read the chapter, it 21 doesn't talk about specifics of how you ensure that 22 those systems are alerted, alarmed, interlocked, or 23 what have you. At least, I didn't find it.
24 CO-CHAIR SUNSERI: Well, he said it exists 25 in discrete drawing --
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199 1 (Simultaneous speaking.)
2 MEMBER BROWN: It's some type of P&IDs of 3 some kind, if you go take 27 drawings and try to 4 stitch them all together yourself in the backroom.
5 CO-CHAIR CORRADINI: But what I'm trying 6 to understand is this a recommendation that you want 7 something added to the DCA, or?
8 MEMBER BROWN: I think the -- my personal 9 opinion, if -- this is my personal opinion. If I had 10 been doing this, I would've provided an architecture 11 diagram in Chapter 21 identifying how interlocks and 12 other alarms would be used to ensure operating -- one 13 control room, three operators, you got to give them 14 some operation instead of having it.
15 And that's just based on experience of 16 watching at least one aircraft carrier trying to 17 operate eight reactor plants and steam systems, and 18 some of it with various means of cross-connecting.
19 So, that's what I would've expected to see in the 20 chapter, and it was absent.
21 MEMBER BLEY: As to Matt's point, you 22 know, you know, they said they do have, you know, 23 P&IDs, which we haven't seen, which would show you the 24 valves one might use. They've said -- they haven't 25 decided whether they're going to control that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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200 1 administratively or by interlocks --
2 (Simultaneous speaking.)
3 CO-CHAIR CORRADINI: For the various 4 shared systems.
5 MEMBER BROWN: For the shared systems, 6 yeah.
7 MEMBER BLEY: And I -- you know, from 8 Charlie's point, I don't think we can insist we can 9 insist they have that in the DAC.
10 PARTICIPANT: That's fine.
11 MEMBER BLEY: But we might want to raise 12 it as something to watch out for later at COL stage.
13 MEMBER BROWN: Okay. Yeah. I'll 14 recognize that we can't dictate what they put in the 15 DAC.
16 CO-CHAIR CORRADINI: Well, you can try.
17 But I just want to sure I understood what you're 18 asking.
19 MEMBER BROWN: I've never been hesitant.
20 We can discuss that later.
21 CO-CHAIR CORRADINI: So, for the full 22 committee meeting, I think we need -- three's only two 23 open items as far as I remember for 19.2. I think it 24 would benefit the members of the committee that aren't 25 here today to hear about those open items and their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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201 1 path to resolution. I think one is relative to the 2 bioshield redesign, which affects a number of things.
3 But I think at least from the standpoint of hydrogen 4 control, this is something that the full committee 5 ought to hear.
6 And the second thing is, relative to their 7 path forward, relative to, I think it was, how the 8 source term is affected. So, I think those two things 9 ought to be in the full committee meeting as part of 10 the presentation to the -- to us and the other 11 members. That's it.
12 CO-CHAIR SUNSERI: Dick?
13 MEMBER SKILLMAN: The concerns that I have 14 about the load drop I already expressed yesterday, so 15 I don't think it is constructive to waste our precious 16 resources now. But that's where the bulk of my 17 concerns lie. Thank you.
18 CO-CHAIR SUNSERI: Thanks. Ron?
19 MEMBER BALLINGER: No additional 20 questions.
21 CO-CHAIR SUNSERI: okay Before I go to 22 Steve, I'm going to come back to Joy.
23 MEMBER REMPE: Oh, I just wanted to 24 follow-up about Mike's comment about open items. And 25 one thing I forgot to mention today was there's an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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202 1 open item, also, about the electrical penetrations and 2 their effect on source term, and I think that's one 3 I'd like to hear about, too.
4 CO-CHAIR SUNSERI: And before I close, 5 I'll go to Steve with any comments you want to make.
6 DR. SCHULTZ: Just a general comment on 7 the aspects of Chapter 19. And it is a general 8 comment. And that is, each of the presentations tend 9 to start off discussion about how best estimate 10 evaluation is going to be done because we're talking 11 about probabilistic risk assessment approach. But 12 somewhere along the line, each of the discussions, we 13 get into handling uncertainty or doing evaluations.
14 And we get to the difference between the goal, the 15 margin, and get into general discussions about how we 16 have plenty of margin, so we don't have to worry about 17 the details of the specific calculation.
18 And I think looking toward -- if you will, 19 looking toward the long-term, the future, it's going 20 to be very important before all of this gets wrapped 21 up. The best-estimate evaluation is very, very clear.
22 And the uncertainties are more clear than what they 23 are today, and that a clear engineering evaluation can 24 be demonstrated so that the margins are really known, 25 not qualitatively expressed.
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203 1 CO-CHAIR SUNSERI: Okay. Very good.
2 CO-CHAIR DIMITRIJEVIC: Matt?
3 CO-CHAIR SUNSERI: Yes, Ves. When you 4 asked me first time, you -- I told you I have any 5 additional comments on 19.5. Since we are summarizing 6 to this --
7 CO-CHAIR SUNSERI: Yeah. If I wasn't 8 clear on that, I apologize. Yeah. The whole --
9 PARTICIPANT: The day.
10 CO-CHAIR SUNSERI: The day, yeah.
11 CO-CHAIR DIMITRIJEVIC: So I can -- I 12 mean, I summarized yesterday, but I had --
13 (Simultaneous speaking.)
14 CO-CHAIR SUNSERI: And you can have 15 another -- yeah. This is like bridge. We keep 16 bidding, you know? Go ahead.
17 CO-CHAIR CORRADINI: Go ahead.
18 PARTICIPANT: What did she just say?
19 CO-CHAIR DIMITRIJEVIC: I just said --
20 (Simultaneous speaking.)
21 PARTICIPANT: It's like bridge. We just 22 keep bidding.
23 CO-CHAIR DIMITRIJEVIC: That's all right.
24 The thing is that what I still have left on the --
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204 1 alpha committee. First I have these comments on the 2 SER itself. There's the three statements where 3 they're technically questionable, and one of them is 4 just technically wrong, which I would like to see 5 change before we --
6 The other thing is that we have -- since 7 then, I learn more about, you know -- even today, I 8 learn more about the certain aspects of the ECCS 9 valves and passive heat removal, and analyzing single 10 units. So one of my concerns was that the analysis of 11 the PRA single unit is not taking in account that 12 there is multiple units and that multiple units could 13 be affected.
14 I don't think that's going to be my big 15 concern about what I heard today. How much time do I 16 have and how did answer? But that's definitely true.
17 This just analyzes the single units and presence of 18 other units, most not consider in the sum of the 19 aspects of the PRA. So, I don't think that will make 20 difference. It would not be my -- I bought it up 21 yesterday, but today I don't think it's going to be my 22 big concern.
23 My bigger concern is that I don't think 24 the sensitivity and uncertainties are completely done, 25 and I'm especially concerned about uncertainties in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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205 1 the passive heat removal because it's very small, and 2 uncertainty in the total result is very small. That 3 would not affect conclusions, you know? Even if 4 uncertainty was done, you know, be the less optimism.
5 It will not change the conclusion. However, these 6 should be showing high uncertainly. That's something 7 you would expect that we didn't see it. It's unusual, 8 so I may make comment on that. I'm not sure how to 9 face it.
10 My other thing is that things which are 11 extremely important from the safety valve, the ECCS 12 values, there was some non-conservative assumptions 13 made, which I learned today. And that those non-14 conservative assumptions, again, are not going to 15 impact the results in the sense that --
16 CO-CHAIR CORRADINI: Can you remind us 17 what you mean by the non-conservative assumptions?
18 CO-CHAIR DIMITRIJEVIC: Well, you know, we 19 were talking about the trip valve versus the valve 20 itself. And the trip valve has a much higher fail 21 probability than that was assumed about ECCS valve 22 fail probability. However, the trip valve is combined 23 with these assumptions that there is some probability 24 the valve will open on delta low delta P and it 25 doesn't need signal and trip valve.
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206 1 But that's assumption and pilot valve was 2 assume, and it wasn't consider in common cause. It 3 could have a impact on the total result. It's a big 4 difference in the failure rate. So there is some non-5 conservative modeling, which would not impact results, 6 you know, because results are so low.
7 CO-CHAIR SUNSERI: Yeah. I thought -- but 8 I thought I heard them say, though, that -- or, like, 9 the pilot valve versus the main valve, that there were 10 different industry-accepted frequencies for those 11 events that --
12 (Simultaneous speaking.)
13 CO-CHAIR DIMITRIJEVIC: Yeah. It's the 14 trip valve, which never had frequency. It was much 15 higher frequency. It was -- that's considered always 16 necessary because it was, as we will say, PRA ended.
17 So, they say, okay, this trip valve fails, there is 18 still probability the valve will open on the low delta 19 P. And because they don't know what the 20 probabilities, it was assumed that it's 0.1.
21 So, now this valve was put in common cause 22 group, and that's a high probability. However, it's 23 multiplied by this 0.1 for two valves, which is 10 to 24 minus 2, lowering impact of that valve significantly.
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207 1 might just -- this is what I -- I think we're 2 repeating what we want to see for the full committee.
3 But I think what I'd ask is, we want to understand 4 better how they got to a number, which you view is 5 non-conservative, so that we hear in the full 6 committee with a better explanation.
7 CO-CHAIR DIMITRIJEVIC: Yeah. I 8 understood that today. I got information.
9 CO-CHAIR CORRADINI: Okay.
10 CO-CHAIR DIMITRIJEVIC: So, I have it, 11 information, and I'm really not comfortable with this.
12 But the question is, so, okay. So, maybe they're not 13 conservative, but the 10 to minus 9 and 10 to minus 4, 14 we are talking hundred thousand -- you know, five 15 orders of magnitude.
16 So, the question is -- I mean, you know, 17 I will point, but this is one of the things which I'm 18 still uncomfortable. There are a lot of things I'm 19 uncomfortable, but you do the sensitivities analysis 20 and you only concentrate on one parameter, then you 21 don't know how combination of your different 22 parameters. So, you say, okay. I will fail or, you 23 know, I will increase common cause.
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208 1 combination of sensitivity cases, and some of them can 2 be combined and they show different results. That's 3 on my major commons which I will, do, do, do, write 4 down, and try to break up --
5 (Simultaneous speaking.)
6 CO-CHAIR CORRADINI: But I think the first 7 one you brought up -- I want to make sure we're on the 8 same page. The first one you brought up, though, I --
9 yesterday, at least, my recommendation is that, for 10 the full committee, the Applicant and the staff have 11 got to at least explain their view of how -- what they 12 find is acceptable relative to the reliability and the 13 effect of degradation. I think we've got to have an 14 exposition of this.
15 The actual valve qualification is later.
16 But to answer your questions about what's non-17 conservative, we've got to have that -- at least 18 discuss with, again, the full committee, just so we're 19 all on the same page with what we sense we hear.
20 CO-CHAIR DIMITRIJEVIC: Okay.
21 CO-CHAIR CORRADINI: Okay?
22 CO-CHAIR DIMITRIJEVIC: Well, I -- you 23 know, the question is also, how do we feel about 24 making comments which will not affect the main 25 conclusions. I mean -- you know?
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209 1 CO-CHAIR CORRADINI: So, let me reverse 2 it. Part of the reason I'm asking these questions is 3 I want to make sure that we give the right guidance to 4 the staff and the Applicant for the full committee.
5 So, for example, I wouldn't look forward -- just to 6 get out of the box, I wouldn't look forward for 7 anything in 19.3 to 19.5. I'd rather they concentrate 8 their time on the open items in 19.1, the open items 9 in 19.2, these questions about valve reliability 10 estimates for their things, versus just, essentially, 11 going through details that aren't necessary.
12 CO-CHAIR DIMITRIJEVIC: Well, you know, 13 for that one the other question is also this passive, 14 which I want to say, they gave us the slice, which we 15 saw in October, which says, okay, this is probability 16 that we have inadequate heat removal. And they say, 17 okay, we did the Monte Carlo, blah, blah, blah, and we 18 came out with these two numbers -- and I run 19 factories, too. That's mean, A, we almost know 20 exactly what that number is because this is very 21 small --
22 (Simultaneous speaking.)
23 CO-CHAIR CORRADINI: Are we in closed 24 session? You're not talking about anything we need to 25 be in closed session about, are we?
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433
210 1 CO-CHAIR DIMITRIJEVIC: Are we in -- no.
2 CO-CHAIR SUNSERI: It doesn't say.
3 CO-CHAIR DIMITRIJEVIC: No.
4 CO-CHAIR CORRADINI: Okay.
5 CO-CHAIR DIMITRIJEVIC: I didn't say --
6 CO-CHAIR SUNSERI: Well, what's it say 7 that it's on?
8 CO-CHAIR CORRADINI: Yeah. We're closed.
9 We're closed.
10 (Simultaneous speaking.)
11 CO-CHAIR CORRADINI: So, why don't we 12 wait --
13 CO-CHAIR DIMITRIJEVIC: Well, I didn't not 14 to use any information which I did not use in the open 15 session.
16 CO-CHAIR CORRADINI: Why don't we wait on 17 discussion?
18 CO-CHAIR DIMITRIJEVIC: All right.
19 CO-CHAIR SUNSERI: All right. Very good.
20 Well, so I think -- Jose? You got any -- we're going 21 to wrap this up. So, I think we had a very good 22 session today. I mean, the presentations informed our 23 thinking quite a bit. And, obviously, we have some 24 things to consider as we formulate our final -- our 25 thoughts on what -- you know, areas impact reasonable NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433
211 1 assurance or not as far as safety goes. And we look 2 forward to the discussing these items further in the 3 full committee meeting. So, thanks to everyone that 4 made a presentation today. And at this point in time, 5 we're going to go out of the open session and into the 6 closed session. Well, I suggest we take a 15-minute 7 break till 2:15, and then we'll start with the closed 8 session.
9 CO-CHAIR CORRADINI: And then we rely on 10 the Applicant and the staff to tell us who's allowed 11 in the room.
12 CO-CHAIR SUNSERI: That's correct. All 13 right. We're at recess until 2:15.
14 (Whereupon, the above-entitled matter went 15 off the record at 1:59 p.m.)
16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433
LO-0519- 65373 May 09, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee meeting on May 14 and 15, 2019.
The materials support NuScales presentation of Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, of the NuScale Design Certification Application. is the nonproprietary presentation entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Rebecca Norris at 541-452-7539 or at rnorris@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-7H4 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12 : ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0519- 65373 :
ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Nonproprietary ACRS Subcommittee Presentation:
NuScale FSAR Chapter 19 Probabilistic Risk Assessment and Severe Accident Evaluation May 14 and 15, 2019 1
PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
NuScale Nonproprietary Chapter 19 Section Title Comment Probabilistic Risk Assessment and Severe 19.0 Overview Accident Evaluation 19.1 Probabilistic Risk Assessment Level 1, 2 Thermal hydraulic &
19.2 Severe Accident Evaluation phenomenological analyses 19.3 Regulatory Treatment of Nonsafety Systems No RTNSS SSCs Strategies and Guidance to Address Loss of Addressed in 19.4 Large Areas of the Plant due to Explosions and Chapter 20 Fires Adequacy of Design Features and Functional 19.5 Capabilities Identified and Described for Overview Withstanding Aircraft Impacts 2
PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
NuScale Nonproprietary ACRS Subcommittee Presentation:
NuScale FSAR Chapter 19.0 and19.1 Probabilistic Risk Assessment May 14, 2019 3
PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
NuScale Nonproprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris Supervisor, Licensing 4
PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
NuScale Nonproprietary Section 19.0: Probabilistic Risk Assessment and Severe Accident Evaluation
- Developed in accordance with applicable regulations, regulatory guidance, and industry standards
- Performed for a single module
- Considered all modes of operation for both internal and external initiating events
- Provides risk insights including those related to risk-significant systems, components, human actions, relevant programs (e.g., RTNSS, SAMDA), and multiple module risk
- PRA demonstrates that the NuScale design exceeds NRC safety goals with significant margin 5
PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Section 19.1: Probabilistic Risk Assessment
- Objective: to assess risks associated with all modes and all hazards for a single NuScale Power Module (NPM)
- Level-1 (CDF) and Level-2 (LRF)
- Full power, internal events (FP-IE)
- Low power and shutdown (LPSD)
- Include crane failure
- Internal fire
- Internal flood
- External flood
- Seismic margins assessment (PRA based) 6 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary PRA Quality Process
- NuScale PRA quality procedure
- Follows guidance provided in NRC Regulatory Guide 1.174
- NuScale PRA follows guidance provided by
- ASME/ANS PRA standard
- NRC Regulatory Guide 1.200 and Interim Staff Guidance 028
- Self-assessment documented by notebook authors
- Self-assessment independently reviewed/verified by outside consultants
- PRA reviewed by outside, independent expert panel 7
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NuScale Nonproprietary PRA Expert Peer Review Group
- Separate and independent from PRA standard self-assessment reviewers
- Expert review group members:
- George Apostolakis (chairman)
- Mark Cunningham
- Rick Grantom
- Dave Moore
- Per Peterson 8
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NuScale Nonproprietary Expert Panel Findings
- Review group authored a final report
- No major concerns or objections
- Minor points that were raised include
>> NuScale multi-module risk approach represents an important first step in advancing the state-of-the-art
>> There are more detailed and sophisticated HRA methods available compared to what was done in the NuScale PRA
>> The terms CDF and LRF are tied to current large reactors and use of these terms in the NuScale design may be misleading 9
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NuScale Nonproprietary Independent Self Assessment
- External review of the NuScale PRA self-assessment against the high level and supporting requirements of the ASME PRA Standard
- In general, there was agreement, and in fact, in some cases, a higher capability category than identified was believed to be met. However, there were also some instances of a lack of concurrence, and possible enhancements were provided
- NuScale was able to incorporate those recommendations into the design certification PRA 10 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Initiating Event Analysis
- Multiple sources of input used to identify potential initiating events (IEs)
- NuScale design-specific master logic diagram
- NuScale design-specific simplified system-level failure modes and effects analysis (FMEA)
- Traditional lists of PRA initiating events
- Continuous focus (over the years of NuScale design and PRA development) on identifying potential initiating events and hazards 11 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Full Power Internal Initiating Events
- Spurious opening of ECCS valve
- Loss of DC power
- Loss of offsite power
- Steam generator tube failure
- LOCA (other) inside CNV
- Secondary-side line break (i.e., feedwater or main steam)
- General reactor trip
- Loss of support system (e.g., instrument air, AC power bus) 12 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Accident Sequence Analysis
- Initiating events and subsequent plant responses evaluated
- Key safety functions identified
- Fuel assembly heat removal, reactivity control, containment integrity
- End states of the accident sequences defined
- Level-1: core damage frequency (CDF)
- Level-2: large release frequency (LRF)
- Event trees constructed for each of the initiating events associated with system successes or failures to accomplish the applicable safety functions 13 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Success Criteria
- The Level 1 PRA overall success criterion is the prevention of core damage, defined by maintaining a peak cladding temperature less than 2,200 degrees Fahrenheit
- This is demonstrated for a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> mission time
- System success criteria is determined by the minimum system availability required to prevent core damage
- The Level 1 success criteria evaluation is built upon a comprehensive simulation suite of more than 40 unique accident sequences
- The Level 2 success criterion is large release defined as a source term resulting in acute whole body 200 rem dose to the maximally exposed individual stationary at the reactor site boundary for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 14 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Success Criteria
- PRA success criteria simulations use NuScales safety-related NRELAP5 code with an input model that starts with NuScales safety-related input model
- The PRA simulations augment the safety-related input model with additional nonsafety-related models for beyond-design-basis phenomena
- Chemical and volume control system (CVCS) and containment flooding and drain system (CFDS) models
- Multi-dimensional core thermal hydraulic and neutronic models are used to simulate complex beyond design basis transients such as ATWS 15 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Human Reliability Analysis
- Human actions are not credited in the evaluation of design basis events
- Human actions only relevant to beyond design basis analyses
- Human error probabilities for beyond design basis events based on methodologies provided in NUREG/CR-4772 and NUREG/CR-6883
- Latent human errors and recovery actions
- As a modeling convenience, when quantifying the PRA model, the bounding human error probability of the complete set of post-initiator human failure events, is used for all independently modeled post-initiator human failure events
- Risk significant human action candidates input to D-RAP
- Operator fails to initiate CFDS injection
- Operator fails to initiate CVCS injection 16 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Post-initiator HEPs in PRA Quantification
- Time available (based on bounding scenarios) for human actions range from 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
- To simplify the quantification of the PRA model, bounding value of the set of HEPs used to quantify all post-initiator HEPs Event Description Value EF HEP01 Human error probability for first HFE in cutset 4.0E-03 10 HEP02 Human error probability for second HFE in cutset 1.5E-01 3 HEP03 Human error probability for third HFE in cutset 0.5 -
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NuScale Nonproprietary NuScale PRA Human Errors Modeled (Pre-Initiator)
Name Description Operator misaligns MDP 0001A CFDS train A manual CFDS--HFE-0001A-UTM-N valves during test and maintenance Operator misaligns MDP 0001B CFDS train B manual CFDS--HFE-0002A-UTM-N valves during test and maintenance Operator misaligns MDP 0002A CVCS train A manual CVCS--HFE-0001A-UTM-N valves during test and maintenance Operator misaligns MDP 0002B CVCS train B manual CVCS--HFE-0002A-UTM-N valves during test and maintenance Operator misaligns CTG 0003X EHVS combustion EHVS--HFE-0001A-UTM-N turbine generator during test and maintenance Operator misaligns DGN 0001X ELVS standby diesel ELVS--HFE-0001A-UTM-N generator during test and maintenance Operator misaligns DGN 0002X ELVS standby diesel ELVS--HFE-0002A-UTM-N generator during test and maintenance Operator miscalibrates safety function modules during MPS---HFE-0001A-UTM-S test and maintenance 18 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary NuScale PRA Human Errors Modeled (Post-Initiator)
Name Description Context Operator fails to unisolate and Used for LOCA-OC (2 IEs), SGTFs, CFDS--HFE-0001C-FOP-N initiate CFDS injection and transients (1 IE)
Used for LOCA-IC (3 IEs), LOCA-OC (letdown) (1 IE), transients (1 IE)
Operator fails to unisolate and CVCS--HFE-0001C-FOP-N and secondary steam line break (1 initiate CVCS injection IE) upon failure of ECCS, and SGTFs Operator fails to locally Local unisolation due to lack of CVCS--HFE-0002C-FOP-N unisolate and initiate CVCS control from a partial loss of DC injection power Operator fails to open ECCS Backup action to MPS autofunction ECCS--HFE-0001C-FTO-N valves failure Backup local action to control room Operator fails to start/load EHVS--HFE-0001C-FTS-N initiation failure during loss of offsite combustion turbine generator power Backup local action to control room Operator fails to start/load ELVS--HFE-0001C-FTS-N initiation failure during loss of offsite backup diesel generator power 19 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Data Sources
- Industry information (e.g., NUREG/CR-6928, LERs) where applicable
- Common cause failure (CCF) modeling based NUREG/CR-5497
- Design-specific analyses
- Passive safety system reliability (i.e., ECCS, DHRS)
- Unique events (e.g., steam generator tube failure) 20 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Quantification
- Including CCF models, failure data correlations and uncertainty analyses
- Using the ASME/ANS PRA Standard convergence criterion, a truncation value of 1E-15 per module year was used for the CDF 21 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Uncertainty Analyses
- Addressed using both quantitative uncertainty analyses and sensitivity studies
- SAPHIRE PRA code has capability for propagating parametric uncertainties
- Sometimes augmented using sensitivity studies (e.g., SGTF)
- Thermal hydraulic analyses typically use bounding inputs
- Uncertainty addressed in all modes and all hazards of single module PRA
- Multi-module risk quantification uses conservative, bounding estimates 22 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Parametric Uncertainty
- The data parameters include initiating event frequencies, component failure probabilities, CCF events and their alpha factors, and human error probabilities
- Initiating event frequencies that rely on generic industry data were assigned an expanded uncertainty distribution (i.e., lognormal error factor = 10)
- SAPHIRE has the built-in ability to perform an uncertainty analysis
- Includes correlating failure probabilities
- After cutsets were generated in SAPHIRE, an uncertainty analyses was performed using the Latin Hypercube uncertainty sampling methodology.
23 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Importance
- Systems
- CNTS (containment isolation valves), ECCS, MPS, and UHS
- Components
- ECCS RVVs and RRVs
- DHRS actuation valves
- RSVs
- CVCS and CES containment isolation valves
- Combustion turbine generator
- Other events and initiators (FV>20%)
- RBC, LOCA inside CNV, LOCA outside CNV, LOOP, internal fires, internal flood
- Human actions (FV>20%)
- CVCS actuation and CFDS actuation (Level 2 and LPSD) 24 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Sensitivity Studies Parameter Parameter Change CDF Result LRF Result Base Case 2.7E-10 1.7E-11 HEP All HEPs set to FALSE 2.0E-10 1.0E-11 HEP All HEPs set to TRUE 3.2E-8 2.8E-9 CCF All CCFs set to FALSE 5.4E-12 1.2E-12 CCF All CCFs set to max value of 0.002 4.2E-6 3.7E-8 LOOP-IE LOOP frequency set to 1 per year 2.2E-9 1.7E-11 (base = 3.1E-2 per year)
LOCA-IC-IE LOCA inside CNV frequency increased 1 3.4E-10 1.7E-11 order of magnitude SGTF-IE SGTF frequency increased to generic value 2.8E-10 2.2E-11 ECCS & ECCS and DHRS passive heat transfer 3.2E-10 1.7E-11 DHRS PSSR failure increased 1 order of magnitude I&C sensors Failure probability of sensors was increased 2.8E-10 1.7E-11 an order of magnitude 25 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 2 Methodology
- Analysis indicates that the only applicable containment vessel (CNV) failure mechanisms are containment bypass events and failure of containment isolation
- No bridge trees or Level 1 plant damage state binning
- Level 2 event tree is directly linked to the Level 1 event trees Core Damage Sequences Core Damage Cutset Mapped to Containment Isolation - CIVs # End State Comments Release Size Close (Phase - PH1) (Phase - PH1)
CD CD-T01 CNTS-T01 1 CD Core Damage LEVEL2-ET 2 NR RC1:CD with Isolation 3 LR RC2:CD with Release 26 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary External Hazards
- External events are evaluated using Level 1 PRA model and the following methodologies
- Internal fire: NUREG/CR-6850
- Internal flood: Part 3 of ASME/ANS RA-Sa-2009
- External flood: Part 8 of ASME/ANS RA-Sa-2009
- High winds: Part 7 of ASME/ANS RA-Sa-2009
- Seismic margin assessment: Part 5 of ASME/ANS RA-Sa-2009 27 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Seismic Risk Evaluation
- NuScale performed a PRA-based seismic margin assessment (SMA)
- Design-specific fragility calculations were performed for SSCs that contribute to the seismic margin
- Consulted with Simpson, Gumpertz & Heger, Rizzo Associates, and Stevenson and Associates
- Generic capacities with design-specific response factors were used for other SSCs
- DC/COL-ISG-020 seismic margin goal: high confidence of low probability of failure (HCLPF) value of 1.67 times the certified seismic design response spectra (CSDRS)
- Corresponds to 0.84g peak ground acceleration (PGA) 28 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary SMA Methodology
- PRA-based SMA uses internal event logic, seismically-induced initiators, and maps seismic failures to random failures
- HCLPF: high confidence (95%) of low probability (5%) of failure
- HCLPF can also be interpreted as a 1% probability of failure at the mean (or best-estimate) confidence level (i.e., at the HCLPF PGA there is a 1% probability of core damage)
- Evaluated at the sequence level using min-max criteria
- Seismic margin determined by those seismic failures that would result in a conditional core damage probability of greater than 1%
- Structural fragilities evaluated for those SSCs that contact the module, are located above the module, or where collapse might damage the module (which is assumed to result in core damage) 29 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Seismic Risk Evaluation Seismic plant response SMA results
- Induced initiator event trees
- Plant level HCLPF: 0.88g
- Structural failures
- Structural failures dominate
- LOCAs - Crane
- Loss of offsite power - Exterior walls
- Seismic failure mapping - Bay walls
- No pre-screening (all PRA cutsets - Module supports included in the SMA)
- Evaluated at 14 ground motion with random failures dominate levels ranging from 0.05g to 3.5g results
- Negligible seismic risk from low power and shutdown states 30 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Select NuScale SMA Structures 31 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Fragility Calculation Parameters
- Design calculations for demand/capacity (D/C) ratio inputs
- Uses bounding, conservative values
- For fragility purposes, design calculates are adjusted to median-centered values, uncertainties quantified
- Structural response factor variables
- Ground motion response
- Damping
- Modeling
- Mode combination
- Time history simulation
- Foundation-structure interaction
- Earthquake component combination 32 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Fragility Calculation Parameters
- Capacity variables
- Strength
- Ductility
- Earthquake scale factor (ESF)
- Used in wall calculations, where capacity changes with demand
- Ratio by which the seismic demand must increase for overall demand to equal capacity
- Static demand + ESF
- seismic demand = static capacity +/- ESF
- dynamic capacity (sign is dependent on load in compression / tension)
- Used to calculated median capacity Am 33 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Low Power and Shutdown
- Potential initiating events are those considered for full power and those unique to LPSD
- Reduced inventory (drain down) events not applicable
- No reduced inventory operations in the NuScale design
- Evaluated external events shown to be not important
- Dropped module event most significant CDF contributor
- Relatively high level of conservatism embedded in analysis 34 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Dropped Module Evaluation
- Drop probability developed based on conceptual reactor building crane design
- Core damage conservatively assumed for dropped module
- For a horizontal module the core partially uncovers
- Containment assumed to fail in a manner that prevents pool water incursion but allows radionuclide release
- Maximum radiological release much less than large release due to pool scrubbing effect
- Up to two operating modules theoretically could be struck by free-falling module, potentially inducing LOCA or transient in struck module 35 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Postulated Dropped Module Impacts
- Potential damage to the decay heat removal system (DHRS) because the heat exchangers are located external to the containment and face central pool channel
- Likelihood is an insignificant contributor to the modeled frequency of secondary side line break initiating event
- Potential damage to the chemical and volume control system (CVCS) piping where the piping penetrates the bay wall as a result of movement of the struck module
- Likelihood is an insignificant contributor to the modeled frequency of the CVCS pipe break outside containment initiating event
- Considering the probability of a load drop, the contribution of a potential module drop to the initiating event frequencies of an operating module is judged to be negligible both in absolute terms and in comparison to the frequency of a randomly occurring initiating events 36 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Multiple Module Evaluation
- Each NPM comprises a separate, independent RPV and CNV, and is serviced by separate, independent safety systems
- Systematic evaluation performed per SRP 19.0
- Single module PRA with bounding multi-module adjustment factors (MMAF) applied to each and every basic and initiating event
- MMAF value of 1.0 for SSCs shared amongst multiple modules and plant wide initiating events (e.g., LOOP)
- MMAF values from 0.1 to 0.3 for SSCs with potential coupling mechanisms between modules (e.g., potential for common cause failures)
- Smallest applied MMAF of 0.01 to events that would nominally be considered independent (e.g., pipe failures) 37 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 1 Insights
- NuScale design exceeds NRC core damage frequency safety goal with significant margin
- Full power internal event CDF 3.0E-10/mcyr
- External initiator CDFs: 1.0E-09 to 6.1E-11/mcyr
- LPSD CDF dominated by module drop event: 8.8E-08/mcyr
- Focused PRA CDF (no credit for nonsafety-related systems): 3.1E-06/mcyr
- Approximately equivalent to a long-term station blackout with no recovery of ac power
- Multiple module CDF factor: 0.13 38 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 2 Insights
- NuScale design exceeds NRC large release frequency safety goal with significant margin
- Full power internal event LRF 2.3E-11/mcyr
- External initiator LRFs: 4.3E-11 to <1E-15/mcyr
- Module drop event does not result in large release
- Focused PRA LRF (no credit for nonsafety-related systems): 1.6E-07/mcyr
- Approximately equivalent to a long-term station blackout with no recovery of ac power
- Multiple module LRF factor: 0.01 39 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 1 Key Insights (1 of 2)
Design Feature/Insight Comment Failure to scram events (ATWS) do Core characteristics result in ATWS power levels that are not lead directly to core damage. comparable to decay heat levels. Heat transfer from the containment vessel (CNV) to reactor pool is adequate to prevent core damage and most ATWS sequences require approximately the same system success criteria as non-ATWS events.
Passive heat removal capability is RSV cycling transfers adequate RCS water to the CNV to allow sufficient to prevent core damage if heat transfer through the RPV to the CNV and ultimately to a reactor safety valve (RSV) cycles. reactor pool to remove decay heat.
Post-accident heat removal through The steam generators and DHRS provide effective heat removal steam generators or decay heat paths to prevent core damage, but are unnecessary if RSV removal system (DHRS) is cycling allows heat transfer to reactor pool. Passive, fail-safe unnecessary if RSVs cycle. DHRS provides a natural circulation closed loop system that does not require pumps, power, or additional water.
Passive, fail-safe emergency core The ECCS consists of 5 valves that fail-safe on a loss of power cooling system (ECCS) functions to and provides a natural circulation path through the core and preserve RCS inventory, which is CNV, thus providing heat transfer to the reactor pool. The sufficient to allow core cooling closed-loop system does not need additional inventory.
without RCS makeup from external source.
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NuScale Nonproprietary Level 1 Key Insights (2 of 2)
Design Feature/Insight Comment Containment isolation preserves Containment isolation eliminates the potential for breaks outside RCS inventory for core cooling of containment to result in loss of RCS inventory. For breaks without external makeup. inside of containment, containment isolation is not necessary to support passive core cooling and heat removal.
Passive, fail-safe safety systems Safety-related mitigating systems are fail-safe on loss of power (ECCS, DHRS, RSVs) include and do not require supporting systems such as lube oil, air or redundancy and do not need support HVAC to function. No single failure results in a loss of system systems, including electric power or function.
operator actions.
There are no risk significant, post- No operator actions, including backup and recovery actions, are initiator human actions associated risk significant to the CDF because of passive system reliability with the full-power PRA. and fail-safe system design.
Risk significant structures, systems The module response to external events is comparable to the and components (SSCs) for external response to internal event due to the passive features of the events are largely the same as those design and independence from support systems such as power.
found risk significant for internal Additional systems and components have been identified as risk events. significant for external events due to a conservative evaluation.
Active systems providing makeup Inventory addition is possible by the active systems chemical inventory to the RPV are not risk and volume control system (CVCS) and containment flooding significant. and drain system (CFDS). Due to the reliability of the passive safety systems, the active systems providing this backup function were found not to be risk significant.
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NuScale Nonproprietary Section 19.1 COL Items Item Number Description COL Item 19.1-1 A COL applicant that references the NuScale Power Plant design certification will identify and describe the use of the probabilistic risk assessment in support of licensee programs being implemented during the COL application phase.
COL Item 19.1-2 A COL applicant that references the NuScale Power Plant design certification will identify and describe specific risk-informed applications being implemented during the COL application phase.
COL Item 19.1-3 A COL applicant that references the NuScale Power Plant design certification will specify and describe the use of the probabilistic risk assessment in support of licensee programs during the construction phase (from issuance of the COL up to initial fuel loading).
COL Item 19.1-4 A COL applicant that references the NuScale Power Plant design certification will specify and describe risk-informed applications during the construction phase (from issuance of the COL up to initial fuel loading).
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NuScale Nonproprietary Section 19.1 COL Items Item Number Description COL Item 19.1-5 A COL applicant that references the NuScale Power Plant design certification will specify and describe the use of the probabilistic risk assessment in support of licensee programs during the operational phase (from initial fuel loading through commercial operation).
COL Item 19.1-6 A COL applicant that references the NuScale Power Plant design certification will specify and describe risk-informed applications during the operational phase (from initial fuel loading through commercial operation).
COL Item 19.1-7 A COL applicant that references the NuScale Power Plant design certification will evaluate site-specific external event hazards (e.g.,
liquefaction, slope failure), screen those for risk-significance, and evaluate the risk associated with external hazards that are not bounded by the design certification.
COL Item 19.1-8 A COL applicant that references the NuScale Power Plant design certification will confirm the validity of the key assumptions and data used in the design certification application probabilistic risk assessment (PRA) and modify, as necessary, for applicability to the as-built, as-operated PRA.
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NuScale Nonproprietary ACRS Subcommittee Presentation:
NuScale FSAR Chapter 19.2 Severe Accident Evaluation May 15, 2019 44 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris Supervisor, Licensing 45 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Section 19.2: Severe Accident Evaluation
- Accident sequences resulting in core damage are evaluated in the Level 2 PRA for potential to challenge containment vessel (CNV) integrity and result in a large radionuclide release
- Large release defined as source term resulting in acute whole body 200 rem dose to the maximally exposed individual stationary at the reactor site boundary for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />
- MACCS off-site consequence calculations demonstrate that sequences with intact CNV are not large release
- CNV bypass accidents counted as large release (simplification for convenience)
- There are no unique phenomenological challenges that are introduced by the NuScale design 46 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Use of MELCOR
- Provides a best estimate evaluation of severe accident challenges to CNV
- Informs conservative evaluations of severe accident challenges
- Provides a physical basis for parameters
- Timing of core damage, core relocation
- Quantity of relocated material, composition of relocated material
- System pressures, temperatures, quantity of hydrogen produced
- Evaluations use limiting values from database of simulations that each involve bounding/conservative simplifications
- End of cycle decay heat load
- DHRS not credited to slow down accident progression
- Evaluations also consider parameters that bound all results observed from database of simulations
- 100% of fuel UO2 relocates at first observed relocation time from database
- Assume debris is molten, pure UO2 composed of no filler materials (e.g., steel, zirconium)
- No credit for water in lower plenum at time of relocation 47 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary MELCOR Model Development
- Thermal-hydraulics modeling developed from NRELAP5 model
- Matching elevations, volumes, flow areas, frictional losses, heat structure material, surface area, thickness, heated diameters, etc
- Benchmarking of steady-state operation and transients demonstrate reasonable to excellent agreement with NRELAP5
- Goal is to approximately match NRELAP5 accident simulation to the point of core damage and then extend simulation into severe accident space
- Severe accident modeling based on appropriate and accurate modeling of NPM design characteristics
- Decay power curve, core component masses and locations, radionuclide inventory, core flow geometry
- Incorporates modeling best practices from
- MELCOR code development staff and industry leading subject matter experts
- State-of-the-Art Reactor Consequence Analyses (SOARCA) reports
- MELCOR guides, manuals, assessments 48 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary In-Vessel Retention (IVR)
- Conservative analysis demonstrates that RPV lower head integrity is maintained if core debris relocates to lower plenum
- Maximum heat flux remains below critical heat flux (CHF) on exterior surface
- Heat generation rate based on conservative assumptions/inputs (e.g., 100% core UO2 - no upward radiation heat losses)
- Assumed CHF threshold conservatively does not credit high absolute pressure and large subcooling in CNV
- With effective external vessel cooling, the lower head remains intact and the severe accident progression is stabilized in RPV 49 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Severe Accident Phenomena CNV integrity not challenged by severe accident phenomena
- In-vessel fuel-coolant interactions (FCI) (i.e., steam explosion) are not sufficiently energetic to induce alpha mode failure due to factors including:
- Small core size, low debris temperatures, small drop height, shallow pool, relatively high system pressure
- Containment overpressure does not occur
- High pressure steel CNV designed for most limiting LOCA blowdown which exceeds maximum severe accident pressures
- Submergence of CNV in UHS provides highly effective pressure suppression
- No concrete interactions to generate non-condensable gases 50 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Consideration of Uncertainty
- If IVR in RPV fails
- High pressure melt ejection (leading to direct containment heating) does not occur because there is no driving pressure differential
- Energetic ex-vessel FCI not likely for similar reasons as in-vessel FCI
- Debris relocated to CNV would be retained by CNV lower head
- Effective external cooling of CNV by reactor pool
- If lower CNV fails
- Pool scrubbing minimizes release
- If upper CNV fails
- Instantaneous release of entire airborne radionuclide inventory in module at time of postulated CNV failure would not constitute a large release 51 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 2 Insights
- Core damage events are stabilized within the RPV
- Severe accident phenomena do not challenge CNV integrity
- Large release does not occur even if RPV and CNV are postulated to fail
- The large release frequency is dominated by containment bypass events 52 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 2 Key Insights (1 of 5)
Design Feature/Insight Comment Containment Isolation The primary purpose of CNTS is to If coolant remains primarily within the RPV, then the core retain primary coolant inventory is covered. If the core is not covered in the RPV then within the CNV. With primary sufficient primary coolant is in the CNV to submerge the coolant inventory maintained in the outside of the lower RPV and establish conductive heat RPV or CNV, cooling of core debris removal from the core debris to the coolant in the CNV is ensured. through the RPV wall.
CNTS terminates releases through Containment penetrations through which releases are penetrations leading outside assumed to occur that dominate risk include those that containment. bypass containment such as CVCS (injection and discharge) and paths through the steam generator tubes (main steam and feedwater piping). Isolation of normally open valves in these penetrations prevents releases from bypassing containment.
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NuScale Nonproprietary Level 2 Key Insights (2 of 5)
Design Feature/Insight Comment Passive Heat Removal The RPV has no insulating material Retaining primary coolant in the containment results in collection and passive heat removal capability of sufficient RCS water in the CNV to allow heat transfer through from the RPV to the CNV is RPV to CNV and ultimately UHS to remove heat generated in sufficient to prevent core debris from the fuel regardless of its location.
penetrating the reactor vessel.
The CNV is uninsulated and passive heat removal capability from the CNV to the UHS is sufficient to prevent the containment from pressurizing and or core debris from penetrating the containment 54 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Level 2 Key Insights (3 of 5)
Design Feature/Insight Comment Severe Accident Containment Challenges (1 of 2)
Primary coolant system Addition of water to the containment from external overpressure failure cannot lead to sources (CFDS) results in submergence of the reactor vessel overpressurization of containment and establishes passive heat removal through the containment (i.e., loss of decay heat removal wall to the reactor pool. Even if containment flooding is not through the steam generators plus successful, the RPV failure mode is such that containment failure of the RSVs to open). ultimate capacity would not be exceeded.
Hydrogen combustion is not likely as There is very little oxygen available (oxygen generated from the containment is normally radiolysis is only a long-term issue) and containment is steam evacuated. inerted under severe accident conditions. In addition, conservative AICC analyses predict containment pressures that do not exceed the design pressure.
In-vessel steam explosions are not Core support failure is expected before the fuel has a chance to likely due to core support design and become molten. With the core uncovered there is little water in volume of lower vessel head. the bottom of the RPV with which core debris can interact.
HPME cannot occur Submergence of the lower RPV With passive heat removal establishes passive heat from the reactor to removal and prevents core containment established, the debris from exiting the RPV. No reactor is depressurized ex-vessel challenges occur if the even if core debris is core remains within the vessel. postulated to exit the vessel.
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NuScale Nonproprietary Level 2 Key Insights (4 of 5)
Design Feature/Insight Comment Severe Accident Containment Challenges (2 of 2)
Ex-vessel steam explosion does not Submergence of the lower RPV establishes passive heat occur with a submerged RPV. removal and prevents core debris from exiting the RPV. No ex-vessel challenges occur if the core remains within the vessel.
Overpressure of containment due to There is no concrete in the containment with which the core non-condensable gas generation is debris could interact and generate non-condensable gases.
not applicable to the NuScale design.
Basemat penetration is not There is no basemat making up the containment boundary. This applicable to the NuScale design. issue is addressed as a part of considering protection against contact of core debris with the containment wall.
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NuScale Nonproprietary Level 2 Key Insights (5 of 5)
Design Feature/Insight Comment Support Systems, Human Action, External Events Support systems are not needed for Safety-related mitigating systems are fail-safe on loss of power safety-related system functions (i.e., and do not require supporting systems such as lube oil, containment isolation) important to instrument air, or HVAC to function.
the Level 2 PRA.
With one exception, there are no risk Operator actions, including backup and recovery actions, are significant, post-accident human not significant to the Level 2 analysis because of passive actions associated with the full- system reliability and fail-safe system design. The operator power internal events Level 2 PRA. action to align CFDS during a CVCS break outside containment The exception is alignment of CFDS meets the risk significance thresholds because of a during accident sequences in which mathematical limitation of the calculation of the Fussell-Vesely isolation of a broken CVCS line measure of importance outside containment fails, ECCS is successful but coolant inventory in containment needs replenishment in order to maintain natural circulation between CNV and the RPV.
Risk significant SSC for external The module response to external events is comparable to the events are largely the same as those response to internal event due to the passive features of the found risk significant for internal design which are not affected by the external events and plant events systems that are protected against external event challenges.
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NuScale Nonproprietary Section 19.2 COL Items Item Number Description COL Item 19.2-1 A COL applicant that references the NuScale Power Plant design certification will develop severe accident management guidelines and other administrative controls to define the response to beyond-design-basis events.
COL Item 19.2-2 A COL applicant that references the NuScale Power Plant design certification will use the site-specific probabilistic risk assessment to evaluate and identify improvements in the reliability of core and containment heat removal systems as specified by 10 CFR 50.34(f)(1)(i).
COL Item 19.2-3 A COL applicant that references the NuScale Power Plant design certification will evaluate severe accident mitigation design alternatives screened as not required for design certification application.
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NuScale Nonproprietary ACRS Subcommittee Presentation:
NuScale FSAR Chapter 19.3 Regulatory Treatment of Nonsafety Systems May 15, 2019 59 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris Supervisor, Licensing 60 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Section 19.3
- There are no RTNSS SSCs in the NuScale design
- None of the five RTNSS criteria were met by any NuScale SSC
- RTNSS is also discussed in FSAR 17.4.3.3 61 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Section 19.3 COL Item Item Number Description COL Item 19.3-1 A COL applicant that references the NuScale Power Plant design certification will identify site-specific regulatory treatment of nonsafety systems (RTNSS) structures, systems, and components and applicable RTNSS process controls.
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NuScale Nonproprietary ACRS Subcommittee Presentation:
NuScale FSAR Chapter 19.5 Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts May 15, 2019 63 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Presentation Team Amber Berger Civil/Structural Engineer Rebecca Norris Supervisor, Licensing Marty Bryan Licensing Project Manager 64 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Introduction and Background
- Plant design for potential effects of beyond design basis large commercial aircraft impact [10 CFR 50.150(a)]
- The reactor core remains cooled, or the containment remains intact
- Spent fuel cooling or spent fuel pool integrity is maintained
- NEI 07-13 methods followed with no exceptions
- Aircraft impact informed the plant design 65 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Scope
- Reactor Building assessed for effects in three areas for postulated aircraft impact
- Physical damage
- Shock damage from shock-induced vibration on structures, systems, and components
- Fire damage from aviation fuel-fed fire 66 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Methodology
- Reactor Building is structure of concern
- NuScale Power Modules
- Spent fuel pool
- Impact locations
- Screening by NEI 07-13
- Radioactive Waste Building (RWB) is intervening structure to mitigate physical damage to RXB, conservatively do not credit RWB in shock assessment
- No credit taken for CRB or TGB 67 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary NuScale Site Plan North NuScale DCA Tier 2 Figure 1.2-1 Conceptual Site Layout 68 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Results
- Physical damage
- Local assessment per NEI formulas for perforation and scabbing
- Global response performed using detailed finite element models and NRC specified force-time history
- RXB external walls prevent physical damage from entering RXB
- No internal missiles for secondary impact
- No impact on containment boundary
- Spent fuel pool protected inside RXB below grade
- Reactor Building crane trolley cannot be dislodged 69 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Results (contd)
- Shock damage
- Aircraft impact causes short duration, high acceleration, high frequency vibration
- Core cooling
- At-power and shutdown scenarios considered
- No active equipment required for success
- Adequate heat removal is shown for all strikes
- Spent fuel
- SFP integrity maintained for all strikes 70 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Results (contd)
- Fire damage
- Design and location of 3-hr fire barriers and 3-hr, 5-psid fire barriers prevent propagation of fire into RXB
- Design and location of 5-psid, fast-acting blast dampers at RXB HVAC key design feature
- Concrete shrouds protect exterior wall pipe and HVAC penetrations from physical damage and prevent fire propagation into the RXB
- Fire that enters through external personnel doors at grade level does not propagate beyond stairwells
- All required operator actions occur prior to impact 71 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Assessment Conclusions
- Design and functional capabilities provide adequate protection of public health and safety
- NuScale plant meets 10 CFR 50.150 regulation
- Maintain containment integrity AND core cooling capability (only required to meet one)
- Maintain SFP integrity
- For most postulated aircraft impact strikes, spent fuel pool cooling maintained, meeting all four CFR requirements 72 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Acronyms (1 of 3)
- CRB Control Building
- BDB beyond design basis
- CVCS chemical and volume control system
- CCF common cause failure
- CSDRS certified seismic design response
- CD core damage spectra
- CDF core damage frequency
- CTG combustion turbine generator
- CES containment evacuation system
- D/C demand/capacity
- CFDS containment flooding and drain
- DGN diesel generator system
- DHRS decay heat removal system
- CFR Code of Federal Regulations
- CHF critical heat flux
- EHVS 13.8 kV and switchyard system
- CIV containment isolation valve
- ELVS low voltage AC electrical
- CNV containment vessel distribution system
- CNTS containment system
- ESF earthquake scale factor
- COL combined license 73 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Acronyms (2 of 3)
- FCI fuel-coolant interaction
- LPSD low power and shutdown
- FMEA failure modes and effects analysis
- LR large release
- FP-IE full power, internal event
- LRF large release frequency
- FSAR Final Safety Analysis Report
- mcyr module critical year
- HCLPF high confidence of low probability
- MDP motor driven pump of failure
- MMAF multi-module adjustment factor
- HEP human error probability
- MPS module protection system
- HPME high pressure melt ejection
- NEI Nuclear Energy Institute
- HVAC heating ventilation and air
- NPM NuScale Power Module conditioning
- NR no release
- IE initiating event
- NRC Nuclear Regulatory Commission
- IVR in-vessel retention
- PGA peak ground acceleration
- LOCA loss of coolant accident
- LOOP loss of offsite power 74 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Acronyms (3 of 3)
- RBC reactor building crane
- SAPHIRE Systems Analysis Programs for Hands-on Integrated Reliability
- RCS reactor coolant system Evaluations
- RG Regulatory Guide
- SGTF steam generator tube failure
- SMA seismic margin assessment
- RSV reactor safety valve
- SOARCA State-of-the-Art Reactor
- RTNSS regulatory treatment of nonsafety Consequence Analysis systems
- SSC structures, systems, and
- RVV reactor vent valve components
- RWB Radioactive Waste Building
- SFP spent fuel pool
- RXB Reactor Building
- SRP Standard Review Plan
- SAMDA severe accident mitigation design
- TGB Turbine Generator Building alternative
- UHS ultimate heat sink 75 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Nonproprietary Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 76 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Safety Evaluation with Open Items:
Chapter 19 Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors NuScale Design Certification Application ACRS Subcommittee Meeting May 14-15, 2019
Agenda
- Presentation Topic for May 14, 2019:
- Section 19.1, Probabilistic Risk Assessment
- Presentation Topics for May 15, 2019:
- Section 19.2, Severe Accident Evaluation
- Section 19.3, Regulatory Treatment of Nonsafety Systems for Passive Advanced Light Water Reactors
- Section 19.5, Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts
- Presentation Topic Included in Chapter 20 (for future discussion):
- Section 19.4, Strategies and Guidance To Address Loss of Large Areas of the Plant Because of Explosions and Fires May 15, 2019 Non-Proprietary 2
Section 19.2 Severe Accident Evaluation
Staffs Review - Overview
- Staff conducted two regulatory audits
- Staff conducted independent confirmatory analyses NuScale Chapter 19 - PRA and Severe May 15, 2019 4 Accident Evaluation
Overview of Review Guidance
- SECY-90-016 and SECY-93-087, and the associated SRMs - containment performance goals NuScale Chapter 19 - PRA and Severe May 15, 2019 5 Accident Evaluation
Severe Accident Mitigation
- Containment performance goals for new reactors
- Probabilistic: CCFP < 0.1
- Deterministic: stay intact for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provide barrier against uncontrolled release
- Uncertainties associated with in-vessel retention prevent confirming the NuScale design meets the Commission containment performance goals NuScale Chapter 19 - PRA and Severe May 15, 2019 6 Accident Evaluation
Severe Accident Mitigation Alternative criteria
- Probabilistic: CCFP < 0.1
- Applied to in-vessel steam explosion and containment bypass
- Deterministic: no large release
- Applied to accidents for which demonstration of in-vessel retention is inconclusive NuScale Chapter 19 - PRA and Severe May 15, 2019 7 Accident Evaluation
Severe Accident Mitigation
- Containment lower head failure
- No large release, due to aerosol deposition in reactor and containment and reactor pool scrubbing
- Steam explosion in containment
- No large release, due to aerosol deposition in reactor and containment NuScale Chapter 19 - PRA and Severe May 15, 2019 8 Accident Evaluation
Severe Accident Mitigation
- High pressure melt ejection
- NuScales unique design results in depressurization before lower head challenged
- Steam explosion in the reactor vessel
- Insufficient energy release to fail containment upper head
- Staff performed independent confirmatory analysis NuScale Chapter 19 - PRA and Severe May 15, 2019 9 Accident Evaluation
Severe Accident Mitigation
- Severe accident-induced steam generator tube failure
- Staff did not evaluate applicants assumption that tubes are less susceptible to failure under compression than under tension
- Low rate of creep under postulated accident conditions
- Small impact on large release frequency
- Steam generator relief valve relieves into containment
- Tube failure leads to release only if the main steam or feedwater isolation valves fail to close NuScale Chapter 19 - PRA and Severe May 15, 2019 10 Accident Evaluation
Severe Accident Mitigation
- NuScale used MACCS to show the dose at the site boundary (0.167 miles) is less than NuScales large release definition
- Benchmarked MACCS plume modeling against ARCON which is valid for close-in distances
- Large margins between the large release definition and the doses calculated using MACCS NuScale Chapter 19 - PRA and Severe May 15, 2019 11 Accident Evaluation
Severe Accident Mitigation
- Reactor pool scrubbing for module drop accidents
- Pool scrubbing reduces the release to a fraction of a large release
- Supplemental analysis showed that NuScales pool scrubbing model is conservative NuScale Chapter 19 - PRA and Severe May 15, 2019 12 Accident Evaluation
Severe Accident Mitigation
- Staffs independent confirmatory analysis
- Staff compared its independent MELCOR simulation results to the applicants
- Staffs simulations generally reproduced the applicants simulations but usually with timing shifts
- Differences unlikely to affect the applicants analysis of severe accident mitigation NuScale Chapter 19 - PRA and Severe May 15, 2019 13 Accident Evaluation
Multi-module Severe Accident Risk
- Hydrogen generation and control
- Core damage sequence of an un-isolated CVCS line break outside containment could lead to a hydrogen combustion event under the bio-shield
- Staff is reviewing the redesigned bio-shield
- Staff is evaluating the potential impact on the multi-module severe accident risk
- Staff is tracking this as an open item (RAI 9447, Question 03.11-19)
NuScale Chapter 19 - PRA and Severe May 15, 2019 14 Accident Evaluation
Summary Severe Accident Evaluation
- Due to the open items, the staff is unable to make a finding on the applicants severe accident evaluation in DCA Part 2, Tier 2, Section 19.2
- RAI 9447
- Accident Source Term Topical Report Review NuScale Chapter 19 - PRA and Severe May 15, 2019 15 Accident Evaluation
Section 19.3 Regulatory Treatment of Non-safety Systems
Regulatory Treatment of Non-safety Systems (RTNSS)
- Review evaluated NuScales determination that there are no RTNSS SSCs
- Staff found the evaluation reasonable and acceptable NuScale Chapter 19 - PRA and Severe May 15, 2019 17 Accident Evaluation
Section 19.5 Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts
Technical Topics Overview
- Staff Verified:
Adequacy of Key Design Feature Descriptions
- Fire Protection
- Structures
- Systems How Key Design Features met 10 CFR Part 50.150(a)(1)
(i) The reactor cores remain cooled and containment remains intact, (ii) The spent fuel pool integrity is maintained Shock Damage Description Reasonably Formulated Assessment of Aircraft Impact Assessment Performed by Qualified Individuals May 15, 2019 Section 19.5: Aircraft Impact Assessment 19
One Confirmatory Item RAI 9241, Question 19.5-23: related to the availability of monitoring functions to ensure that the identified design features are performing as expected following the impact of a large commercial aircraft.
May 15, 2019 Section 19.5: Aircraft Impact Assessment 20
Technical Topics
==
Conclusions:==
- Section 19.5 meets 10 CFR Part 50.150 because the applicant:
Adequately identified and described the key design features and functional capabilities credited to meet 10 CFR Part 50.150 Adequately described how the key design features meet the assessment requirements in 10 CFR Part 50.150(a)(1)
Developed a reasonably formulated assessment performed by qualified individuals May 15, 2019 Section 19.5: Aircraft Impact Assessment 21
Abbreviations And Acronyms
- CCFP - conditional containment failure probability
- CVCS - chemical and volume control system
- DCA - design certification application
- NRC - Nuclear Regulatory Commission
- RAI - request for additional information
- RTNSS - regulatory treatment of nonsafety systems
- SSCs - structures, systems, and components NuScale Chapter 19 - PRA and Severe May 15, 2019 22 Accident Evaluation
Questions/comments from members of the public before the closed session starts?
23 Non-Proprietary
LO-0519-65534 May 9, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee meeting on May 15, 2019. The materials support NuScales presentation of Chapter 21, Multi-Module Design Considerations, of the NuScale Design Certification Application. is the nonproprietary presentation entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Nadja Joergensen at 541-452-7338 or at njoergensen@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-7H4 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Prosanta Chowdhury, NRC, OWFN-8H12 : ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0519-65534 :
ACRS Subcommittee Presentation: NuScale FSAR Chapter 21, Multi-Module Design Considerations, PM-0519-65533, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Nonproprietary ACRS Subcommittee Presentation NuScale FSAR Chapter 21 Multi-Module Design Considerations 1
May 15, 2019 PM-0519-65533 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Presenters Chris Maxwell Senior Reactor Operator 4 Sarah Bristol Supervisor, Probabilistic Risk Assessment Nadja Joergensen Licensing Specialist 3 2
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Modular Designs
- Per 10 CFR 52.1, modular design means A nuclear power station that consists of two or more essentially identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated independent of the state of completion or operating condition of any other module co-located on the same site, even though the nuclear power station may have some shared or common systems.
- The NuScale Power Plant design is consistent with this definition.
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Modular Designs
- 10 CFR 52.47(c)(3) requires evaluation of module operating configurations, considering:
- Common (shared) systems
- Interface requirements
- System interactions
- Restrictions during construction and startup 4
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Shared Systems
- The NuScale Power Plant is designed such that each NPM can be safely operated independent of other NPMs.
- With the exception of the ultimate heat sink (UHS),
safety-related systems are module-specific and functionally independent of shared systems and other NPMs.
- Emergency core cooling system
- Containment system
- Decay heat removal system
- Module protection system
- Demineralized water isolation valves 5
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Shared Systems
- The UHS has sufficient capacity to remove heat from one NPM experiencing a DBA, while simultaneously removing heat from 11 NPMs in shutdown and cooldown.
- Long-term heat removal is provided by the UHS and the module-specific safety-related systems without reliance on other shared systems or operator action.
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Shared Systems
- Module heatup system
- Fire protection system
- Boron addition system
- Fire detection system
- Containment flooding and drain
- Fuel handling equipment system
- Module assembly equipment
- Reactor component cooling
- Instrument air system water system
- Gaseous radioactive waste
- Process sampling system system
- Liquid radioactive waste system
- Auxiliary boiler system
- Normal control room HVAC
- Site cooling water system system
- Nitrogen distribution system
- Reactor building HVAC system
- Demineralized water system
- Control room habitability system 7
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Shared Systems
- Chilled water system
- Highly reliable DC power system
- Reactor pool cooling system
- Normal DC power system
- Safety display and indication
- Pool surge control system system
- Pool cleanup system
- Plant protection system
- Pool leak detection system
- Plant control system
- Spent fuel pool cooling system
- 13.8 kV and switchyard system
- Medium voltage AC electrical distribution system
- Low voltage AC electrical distribution system 8
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System Interactions
- The shared systems that have potential for an adverse system interaction or an undesirable multi-module interaction were evaluated.
- The evaluations demonstrate that shared system operation does not result in adverse system interactions, such as
- a loss of a safety-related function,
- a DBE and a simultaneous degradation of a safety-related function,
- a DBE and simultaneous degradation of critical operator information, or
- a DBE and a requirement for operator actions outside the control room.
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Interface Requirements
- Shared systems that serve one NPM at a time are equipped with isolation features that prevent a direct module-to-module interface during normal operation.
- For an adverse multi-module interaction to occur as a result of a failure associated with a shared system that serves one NPM at a time:
- Abnormal lineups
- Multiple concurrent failures 10 PM-0519-65533 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Interface Requirements
- The reactor component cooling water system (RCCWS) is the only shared system that directly interfaces with multiple NPMs and is also designed to simultaneously support more than one NPM at a time.
- The RCCWS is designed such that no single failure can cause the loss of RCCWS heat removal from more than one NPM.
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Construction and Startup
- During the construction phase that occurs prior to the initial NPM fuel load, the shared and module-specific systems within the Reactor Building (RXB), Control Building (CRB), and Radioactive Waste Building (RWB) are substantially completed with the exception of the installation of additional NPMs.
- The construction method and the phased expansion of NPMs provide assurance that
- the operating configuration is not materially different than that assumed in the safety analysis and
- that the independence of NPM safety-related systems is maintained.
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Construction and Startup
- In addition, the analysis of shared system interactions continues to apply to the operating NPMs during installation of subsequent NPMs.
- Consequently, restrictions in operating configurations or interface requirements are not necessary to ensure the safe operation of operating NPMs during installation, testing, or startup of subsequent NPMs.
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Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 14 PM-0519-65533 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Safety Evaluation with Open Items:
Chapter 21, Multi-Module Design Considerations NuScale Design Certification Application Review ACRS Subcommittee Meeting May 15, 2019
NRC Staff Review Team
- Technical Staff Hanry Wagage - FSAR Tier 2, Chapters 6 and 9 Joseph Ashcraft - FSAR Tier 2, Chapter 7 Nan Chien, Chang Li - FSAR Tier 2, Chapter 9 Sheila Ray - FSAR Tier 2, Chapter 8 Angelo Stubbs - FSAR Tier 2, Chapters 9 and 10 Ryan Nolan, Michelle Hart - FSAR Tier 2, Chapter 15 Marie Pohida - FSAR Tier 2, Chapter 19
- Project Managers Gregory Cranston - Lead Project Manager Prosanta Chowdhury - Chapter Project Manager May 15, 2019 Chapter 21 Multi-Module Design Considerations 2
NuScale Standard Design Shared Systems Evaluated by NRC Staff (SE Section) (No. NPMs Supported)
- 1. RWB heating, ventilation, and air conditioning (HVAC) system (9.4.3) (12 NPMs)
- 2. Diesel generator building HVAC system (9.4.4) (12 NPMs)
- 3. Turbine building HVAC system (9.4.4) (6 NPMs)
- 4. Annex building HVAC system (9.4.2) (12 NPMs)
- 5. Containment flooding and drain system (9.3.6) (6 each for two independent subsystems)
- 6. Normal control room HVAC system (9.4.1) (12 NPMs)
- 7. Reactor Building (RXB) HVAC system (9.4.2) (12 NPMs)
- 8. Control room habitability system (6.4) (12 NPMs)
- 9. Boron addition system (9.3.4) (12 NPMs)
- 10. Reactor component cooling water system (9.2.2) (6 each for two independent subsystems)
- 11. Circulating water system (10.4.5) (6 each for two independent subsystems)
- 12. Site cooling water system (9.2.7) (12 NPMs)
May 15, 2019 Chapter 21 Multi-Module Design Considerations 3
NuScale Standard Design Shared Systems Evaluated by NRC Staff (SE Section) (No. NPMs Supported)
- 13. Demineralized water system (9.2.3) (12 NPMs)
- 14. Auxiliary boiler system (10.4.10) (12 NPMs)
- 15. Potable water systems (9.2.4) (12 NPMs)
- 16. Ultimate Heat Sink (9.2.5) (12 NPMs)
- 17. 13.8 KV and switchyard system, medium voltage AC electrical distribution system (EMVS), low voltage AC electrical distribution system (ELVS) (8.3.1)
(12 NPMs)
- 18. Highly reliable DC power system (EDSS) common (EDSS-C) (8.3.2) (12 NPMs)
- 19. Normal DC power system (EDNS) (8.3.2) (12 NPMs)
- 20. Safety display and indication system (SDIS) (7.2.11) (12 NPMs)
- 21. Plant Protection System (PPS) (7.2.11) (12 NPMs)
- 22. Plant Control System (PCS) (7.2.11) (12 NPMs)
- 23. Utility water system (9.2.9) (12 NPMs)
May 15, 2019 Chapter 21 Multi-Module Design Considerations 4
NuScale Standard Design Shared Systems Evaluated by NRC Staff
- The failure of shared systems that are not safety-related is considered within the NuScale transient and accident analyses and is evaluated in SER Chapter 15, Transient and Accident Analysis.
- SER Section 15.0.0, Classification and Key Assumptions, contains the staffs review of the categorization and classification of Design Basis Events (DBEs).
- SER Section 15.0.3, Radiological Consequences of Design Basis Accidents, contains staffs review of the radiological consequences of DBEs.
- The staff discussed multi-module risk including internal and external events in SER Section 19.1.4.9, Evaluation of Multimodule Risk.
May 15, 2019 Chapter 21 Multi-Module Design Considerations 5
Chapter 21 SE Conclusion
- Consistent with the SRP and, DSRS if applicable, the staff reviewed information presented in the DCA on the design and operation of the aforementioned shared systems.
- As applicable to NuScale multi-module design considerations and assertions, the staff has documented findings and conclusions in the SER sections cited in the previous slides.
May 15, 2019 Chapter 21 Multi-Module Design Considerations 6
Acronyms DBE: Design Basis Event DCA: Design Certification Application DSRS: Design Specific Review Standards HVAC: Heating, Ventilation, and Air Conditioning ITAAC: Inspections, Tests, Analyses, and Acceptance Criteria NPM: NuScale Power Module NRO: US NRC Office of New Reactors NRR: US NRC Office of Nuclear Reactor Regulation PCS: Plant Control System PPS: Plant Protection System RWB: Radioactive Waste Building RXB: Reactor Building SDIS: Safety Display and Indication System SER: Safety Evaluation Report SRP: Standard Review Plan May 15, 2019 Chapter 21 Multi-Module Design Considerations 7