ML20063J331
| ML20063J331 | |
| Person / Time | |
|---|---|
| Issue date: | 02/19/2020 |
| From: | Charles Brown Advisory Committee on Reactor Safeguards |
| To: | |
| Brown, C, ACRS | |
| References | |
| Download: ML20063J331 (407) | |
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, February 19, 2020 Work Order No.: NRC-0801 Pages 1-276 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 WEDNESDAY 10 FEBRUARY 19, 2020 11 + + + + +
12 ROCKVILLE, MARYLAND 13 + + + + +
14 The Subcommittee met at the Nuclear 15 Regulatory Commission, Two White Flint North, Room 16 T2D10, 11545 Rockville Pike, at 8:30 a.m., Walter L.
17 Kirchner and Jose March-Leuba, Co-Chairs, presiding.
18 19 COMMITTEE MEMBERS:
20 WALTER L. KIRCHNER, Co-Chair 21 JOSE MARCH-LEUBA, Co-Chair 22 RONALD G. BALLINGER, Member 23 DENNIS BLEY, Member 24 CHARLES H. BROWN, JR., Member 25 VESNA B. DIMITRIJEVIC, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 DAVID PETTI, Member 2 JOY L. REMPE, Member 3 PETER RICCARDELLA, Member 4 MATTHEW W. SUNSERI, Member 5
6 ACRS CONSULTANTS:
7 MICHAEL L. CORRADINI 8 STEPHEN SCHULTZ 9
10 DESIGNATED FEDERAL OFFICIAL:
11 CHRISTOPHER BROWN 12 MIKE SNODDERLY 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 AGENDA 2 Opening Remarks, Walter Kirchner . . . . . . . . 4 3 Staff Opening Remarks, Rebecca Patton . . . . . . 10 4 NuScale Opening Remarks, Matthew Presson . . . . 13 5 Rod Ejection Accident Methodology, 6 Kenny Anderson, NuScale . . . . . . . . . . . . 16 7 Loss-of-Coolant Accident Evaluation 8 Model (NuScale) . . . . . . . . . . . . . . 62 9 Non-Loss-of-Coolant Accident Analysis 10 Methodology (NuScale) . . . . . . . . . . 100 11 Rod Ejection Accident Methodology 12 (NuScale) . . . . . . . . . . . . . . . . . .
13 Loss-of-Coolant Accident Evaluation 14 Model (NRR) . . . . . . . . . . . . . . . . .
15 Non-Loss-of-Coolant Accident Analysis 16 Methodology (NRR) 17 Rod Ejection Accident Methodology (NRR) 18 Opportunity for Public Comment . . . . . . . . 275 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 KIRCHNER: This meeting will now 4 come to order. Good morning.
5 This is a meeting of the Advisory 6 Committee on Reactor Safeguards NuScale Subcommittee.
7 I am Walt Kirchner, Chairman of the NuScale 8 Subcommittee. My colleague, Jose March-Leuba, will 9 co-chair this meeting with me today.
10 Members in attendance today are Ron 11 Ballinger, David Petti, Joy Rempe, Jose March-Leuba, 12 Pete Riccardella, Charles Brown, Dennis Bley, Vesna 13 Dimitrijevic. Sorry. And we also have --
14 PARTICIPANT: That's easy for you to say.
15 CO-CHAIR KIRCHNER: Yeah. We have two 16 distinguished consultants with us today, our former 17 chairman, Mike Corradini, and Stephen Schultz. Mike 18 Snodderly and Christopher Brown are the designated 19 federal officials for this meeting.
20 The subcommittee will be briefed on loss-21 of-coolant accident evaluation model, non-loss-of-22 coolant accident analysis methodology, and rod 23 ejection accident methodology. Today we have members 24 of the NRC staff and NuScale to brief the 25 subcommittee.
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5 1 The ACRS was established by statute and is 2 governed by the Federal Advisory Committee Act, FACA.
3 The NRC implements FACA in accordance with its 4 regulations found in the Title 10 of the Code of 5 Federal Regulations Part 7.
6 The committee can only speak through its 7 published letter reports. We hold meetings to gather 8 information and perform preparatory work that will 9 support our deliberations at a full committee meeting.
10 The rules for participation in all ACRS 11 meetings were announced in the Federal Register on 12 June 13, 2019.
13 The ACRS section of the U.S. NRC public 14 website provides our charter, bylaws, agendas, letter 15 reports, and full transcripts of all full and 16 subcommittee meetings, including slides presented 17 there. The meeting notice and agenda for this meeting 18 were posted there.
19 Portions of this meeting can be closed as 20 needed to protect proprietary information pursuant to 21 5 USC 552(b)(c)(4).
22 As stated in the Federal Register notice 23 and in the public meeting notice posted to the 24 website, members of the public who desire to provide 25 written or oral input to this subcommittee may do so NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 and should contact the designated federal official 2 five days prior to the meeting as practicable.
3 We have also set aside ten minutes for 4 comments from members of the public attending or 5 listening to our meetings. We have not received 6 written comments or requests for time to make oral 7 statements for members of the public regarding today's 8 meeting.
9 A transcript of the meeting is being kept 10 and will be made available on the ACRS section of the 11 U.S. NRC public website.
12 We request that participants in this 13 meeting please use the microphones located throughout 14 the meeting room when addressing the subcommittee.
15 Participants should first identify themselves and 16 speak with enough volume and clarity so that they can 17 be readily heard.
18 A telephone bridge line has been 19 established for the public to listen to the meeting.
20 To minimize disturbance, the public line will be kept 21 in a listen-in-only mode.
22 To avoid further disturbances, I request 23 that all attendees put their electronic devices, like 24 cell phones, in the off or noise-free mode.
25 And now we'll proceed with the meeting.
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7 1 I'll just make one quick comment.
2 Given the sensitive nature of a lot of the 3 information, which requires us to access it only 4 through NRC laptops and approved devices, not getting 5 the slide information in advance causes problems, 6 because for many of us if we're traveling, we can't 7 use our NRC laptops while we're traveling. And so I'd 8 just make that observation for the record.
9 And now I'll turn to my co-chair, Jose 10 March-Leuba.
11 CO-CHAIR MARCH-LEUBA: So thanks, Walt.
12 And following up on that, we have a meeting March 2nd.
13 If we could get their slides by Friday instead of 14 Sunday night, it would be greatly appreciated.
15 CO-CHAIR KIRCHNER: Yes.
16 CO-CHAIR MARCH-LEUBA: All right. So, 17 that said, we have an unusual, not unheard of, but 18 unusual situation. We have a subcommittee meeting in 19 which we are reviewing three topical reports. We 20 typically only have one subcommittee for every topical 21 report.
22 So we are going to go through all of the 23 open session for all three topical reports. And then 24 we'll move into closed session for the rest of them.
25 Okay.
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8 1 We know that is in the agenda. I just 2 wanted to make sure you understand that when we finish 3 the review of the first one in the open session, we 4 have not finished the review. We will pick up this 5 afternoon.
6 On that topic, we have a reserved time 7 tomorrow morning to follow up, continue whatever we 8 don't cover in the closed session. By 4:00 or 4:30, 9 we'll see where we are. We may not have to meet 10 tomorrow. It depends on how quiet the members are.
11 And my bet will be we'll be here tomorrow, but just so 12 you know.
13 Another thing that makes this situation 14 unusual is these topical reports describe the 15 methodology that has been used to analyze the NuScale 16 reactor, Chapter 15 and Chapter 19 and all the other.
17 And that was completed, the review from 18 the staff was completed on December of last year. And 19 we are now reviewing these topical reports at the ACRS 20 level. I wanted to make sure that the staff is, to 21 give time to the staff to address this situation, 22 which is clearly not desirable.
23 And my point is we have a good excuse this 24 time. But this should not become a precedent that 25 every single design certification that comes here NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 waits until everything is done to certify, to approve 2 the methodology, because if we find even a minor 3 detail on this modification on your methodology, the 4 only recourse is for us to tell you to rerun all your 5 Chapter 15 analysis again.
6 So you're putting the ACRS in a very 7 difficult situation because it is a big step that --
8 I mean, for any comment that we make on your 9 methodology, it has to raise to a tremendous 10 significance to make you do a lot of work and a big 11 delay. Okay.
12 So right now this is a member's comment.
13 And I hope to convince my colleagues to write in the 14 letter that this, while acceptable this time, if we 15 find it acceptable this time, it should not make a 16 precedence for future certifications. Just for the 17 members keep that in mind, because I will propose that 18 to be in the letter during our full committee meeting.
19 Another question I got from Chris was we 20 will make comments about what we want to hear in the 21 full committee meeting, which is scheduled for March, 22 at the end of all the presentations.
23 So we will have comments on all three 24 topical reports, so take notes if you want. What you 25 want to see in the topical report full committee NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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10 1 meeting next month, take notes and we will cover that 2 at the end.
3 MEMBER KIRCHNER: And with that, we'll 4 turn to Becky.
5 MS. PATTON: Thank you. I'd just like to 6 start by thanking the committee for their time 7 reviewing these topical reports.
8 As you're aware, these are three important 9 methodology topical reports for the NuScale design.
10 These are referenced in the design certification 11 application, FSAR Section 15.0.2.
12 These topical reports were accepted for 13 review by the staff back on April 27, 2017. And the 14 staff had accepted the NuScale DCA for review on March 15 23, 2017.
16 The three methodologies were reviewed in 17 parallel with the staff's Chapter 15 review, although 18 this is not the ideal situation, that being when the 19 methodology is reviewed and approved in advance of the 20 methodology being applied to the design. However, the 21 staff did work with NuScale to accommodate this 22 situation.
23 At the Chapter 15 Phase 2 ACRS meetings 24 last year, the methodologies were open items, although 25 at the time most of the methodology issues had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 1 actually reach resolution. Some, like the update of 2 NRELAP Code to Version 1.4, were ongoing.
3 In the fall of last year, the technical 4 issues with the topical reports were all closed. But 5 the safety evaluation reports were in various stages 6 of editing, writing, and concurrence.
7 So what we decided was to ensure a close 8 focus on closing all of the technical issues with the 9 design, including those on the design certification 10 application, the office prioritized closure of all 11 technical issues above writing, editing, and 12 concurrence.
13 All technical issues were closed prior to 14 the office issuing the Chapter 15 Phase 4 SER in mid-15 December. Since the technical issues with the 16 methodologies were closed, the office strategy was to 17 defer the final writing, editing, and concurrence for 18 the topical report SERs until January.
19 In the Phase 4 SERs, with the, the 20 issuance of the methodology SERs is left as 21 confirmatory items because the topical report SERs 22 hadn't been issued. The approved versions of the 23 topical reports are going to be referenced as part of 24 the Phase 6 SER.
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12 1 of the process and the strategy that we had used and 2 why these are being presented now.
3 CO-CHAIR MARCH-LEUBA: So are you implying 4 that all the technical review had been completed --
5 MS. PATTON: That's correct.
6 CO-CHAIR MARCH-LEUBA: -- prior to the 7 issuing of SER?
8 MS. PATTON: That's correct. Before we 9 issued the Phase 4 SER for Chapter 15, all of the 10 technical issues, including those on the methodology, 11 had been closed.
12 CO-CHAIR MARCH-LEUBA: And the reason for 13 the delay is a technicality and going through the 14 lawyers, all the good that you have to go --
15 MS. PATTON: Right.
16 CO-CHAIR MARCH-LEUBA: -- SER.
17 MS. PATTON: Yeah, and right. And there 18 were some portions where the staff hadn't finished the 19 writing yet, but they had, you know, accepted the RAI 20 responses and the closing of the technical issues.
21 CO-CHAIR MARCH-LEUBA: Okay. So, again, 22 one member's opinion, I think this was an unacceptable 23 approach for this time. I'm just worried that this 24 becomes the precedent, and everybody is going to be 25 doing the same thing over and over again.
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13 1 We should be at least, tell the new 2 submittals, not NuScale, but new reactors that our 3 expectation is to finish the review before we do the 4 review of the Chapter 15, 19, and others.
5 MS. PATTON: We understand your comment.
6 CO-CHAIR KIRCHNER: Okay. Now we'll turn 7 to NuScale. Matthew, do you want to start? Please, 8 go ahead.
9 MR. PRESSON: Thank you very much. I'm 10 Matthew Presson, licensing project manager with 11 NuScale. And do I need to get --
12 CO-CHAIR MARCH-LEUBA: The court reporter 13 can't hear you. Can you? Yeah, so it's -- the room 14 might be a little low.
15 (Off-microphone comments.)
16 MR. PRESSON: Yeah. And as you've 17 discussed, we will be going over several topical 18 reports today. We will be starting with the rod 19 ejection accident methodology. And up here is Kenny 20 Anderson, our nuclear fuels analyst.
21 There's a quick agenda, and there we go.
22 Just to give a quick overview and kind of scope our 23 discussions today, we wanted to provide this slide to 24 kind of show you where it all fits together.
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14 1 from RELAP5-3D. A lot of that discussion of that 2 development is in the LOCA method, at least it's been 3 on more and non-LOCA method. And we also have our 4 control rod ejection methods.
5 So those are the -- up at the top in the 6 image, we have NRELAP5. We have our topical report, 7 LOCA, in blue. In brown beneath that, we have our 8 control rod ejection. And then in purple, we have the 9 non-LOCA methods.
10 So, kind of branching off from that, we 11 have our technical reports, which go into Chapter 15 12 which will be discussed two weeks from now. But this 13 is kind of the basis of --
14 CO-CHAIR MARCH-LEUBA: For those of us 15 that have short memory retention --
16 MR. PRESSON: Yeah.
17 CO-CHAIR MARCH-LEUBA: -- can you describe 18 the difference between a topical and a technical 19 report?
20 MR. PRESSON: Yeah, so the topical reports 21 are stand-alone licensing documents. We'd submit 22 them. They provide a method, for example, for LOCA, 23 how we would handle a loss-of-coolant accident. The 24 way that it is applied within the FSAR, we will have 25 FSAR-specific events.
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15 1 And then for some of our other branching 2 analyses, such as containment response, we have a 3 technical report within the FSAR, as we did not have 4 a separate method going through a separate approval 5 process for it as these topical reports --
6 CO-CHAIR MARCH-LEUBA: The technical 7 report does not have an SER associated with it.
8 MR. PRESSON: Correct.
9 CO-CHAIR MARCH-LEUBA: The SER for Chapter 10 15 or whatever it applies is the one that approves the 11 technical report --
12 MR. PRESSON: Correct.
13 CO-CHAIR MARCH-LEUBA: -- as if it was a 14 chapter of the FSAR.
15 MR. PRESSON: Yeah, the technical reports 16 are essentially part of the chapter.
17 CO-CHAIR MARCH-LEUBA: Call it an appendix 18 of the SER.
19 MR. PRESSON: Yes.
20 CO-CHAIR MARCH-LEUBA: Okay.
21 MR. PRESSON: So, you're going into some 22 of the more technical details. So these topical 23 reports are stand-alone. They are separate approvals.
24 We do use those methods in the FSAR. But we get to, 25 the ACRS gets to review and offer their comments on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 1 those separately here at this meeting. So --
2 CO-CHAIR MARCH-LEUBA: Okay.
3 MR. PRESSON: All right. And with that, 4 I will pass this over to Kenny.
5 MR. ANDERSON: Good morning. Thank you.
6 I'm Kenny Anderson from NuScale. I will be talking a 7 brief overview of the rod ejection event in this open 8 session.
9 So rod ejection is very unique. And 10 that's why it has its own topical. And the rest of 11 the day we'll be talking about the other Chapter 15 12 events.
13 And rod ejection is one that is really 14 dominated by the nuclear physics. And so I have kind 15 of a comparison table to kind of stress how or 16 illustrate how unique it is. And that's why we chose 17 to focus in one topical.
18 The goal of doing the analysis is to 19 satisfy essentially General Design Criteria 28. And 20 we'll talk about that.
21 Yeah, and then the other thing, too, is 22 that it does have very unique acceptance criteria.
23 This is a comparison of the acceptance criteria and 24 whether it's unique to other Chapter 15 events. And 25 that's the last column.
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17 1 And the ones that say no, those are the 2 same as other Chapter 15 events, for example, coolant, 3 high coolant system pressure or critical heat flux.
4 The ones that are yes, those are 5 specifically related to energy deposition in the fuel.
6 And that's to protect the fuel from either dispersal 7 or at core coolability.
8 And then we'll go a little bit more on 9 these, on the different fuel criteria. So, over time, 10 the acceptance criteria for fuel has evolved. And we 11 just wanted to kind of explain where we are with that.
12 And this first figure is from NUREG-800.
13 And that's the criteria we use currently in our 14 topical report. And you can kind of, and the dashed 15 lines, you can kind of get a feel of where NuScale is 16 and related to their, to our corrosion rate. So we 17 fit within that box.
18 In the future, Draft Guide 1327, the 19 acceptance criteria, we do expect to change. And you 20 can see it's, they look pretty similar figures. The 21 key difference is on the X-axis. There's -- the units 22 are different. It's hydrogen uptake. And as that 23 becomes law, we expect to have the methodology and 24 support that as well.
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18 1 figure, the dotted line, the NuScale criteria, is that 2 a fixed criteria or it's a goal that you are --
3 MR. ANDERSON: That's our, that's what we 4 propose to be our criteria in the topical report. So, 5 instead of using the more complicated shape, we say a 6 max of 75 calories per gram. And then that way we 7 don't do an exposure census to ensure that we're 8 within the more complicated shape.
9 CO-CHAIR MARCH-LEUBA: And can you give us 10 a preview of the Chapter 15 results? Where do you 11 fit? Where does the real NuScale fit in that 12 criteria? Can you point with the mouse more or less 13 where --
14 MR. ANDERSON: Sure. Yeah, just in this 15 box here is where we fit. And we'll see --
16 CO-CHAIR MARCH-LEUBA: Yeah, but how close 17 to the, how close are you to limits?
18 MR. ANDERSON: There's a lot of margin.
19 And in the closed session, we have exact numbers and 20 --
21 CO-CHAIR MARCH-LEUBA: Oh, okay.
22 MR. ANDERSON: -- the exact margin.
23 DR. CORRADINI: That's why we don't see it 24 here.
25 MR. ANDERSON: Yes.
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19 1 DR. CORRADINI: Okay. That's what I was 2 waiting for you to say. Thank you.
3 MEMBER PETTI: Just a question, and maybe 4 you'll do it in the closed session. What would the 5 dotted line --
6 CO-CHAIR MARCH-LEUBA: Do you have the 7 microphone?
8 MEMBER PETTI: What would the dotted lines 9 look like on the Draft Reg Guide?
10 MR. ANDERSON: Yeah, they would, we would 11 still have the 75 here. And then M5 is our current 12 fuel. And it has very low corrosion. So I would 13 expect it to be in the less than the 50 over here on 14 DPM.
15 DR. CORRADINI: But I guess since you 16 brought this slide up, I will ask the question and 17 then somebody else who understands materials better 18 than I. But now there's a disconnect. When you 19 showed an M5 reserve load, there is not a one-to-one 20 -- the one-to-one correspondence between Zircaloy 21 uptake of oxygen and hydrogen have -- there is a 22 disconnect.
23 So does hydrogen pick-up now dominate the 24 issue, and that's why you're changing the X-axis on 25 the right-hand side?
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20 1 MR. ANDERSON: Yes. So that was a -- it's 2 more than a unit conversion. It's a --
3 DR. CORRADINI: Yeah, yeah, yeah.
4 MR. ANDERSON: It's the physics behind it.
5 DR. CORRADINI: Yeah, I understand that.
6 But then --
7 MR. ANDERSON: Right. Yeah, but it's 8 generally correlated to exposure.
9 DR. CORRADINI: Okay. And so the exposure 10 that you kind of wiggled around is related to how much 11 burn up?
12 MR. ANDERSON: In our current design 13 roughly --
14 DR. CORRADINI: If you're allowed to say 15 here.
16 MR. ANDERSON: Yeah, that's fine. Roughly 17 30 to 40 of the gigawatt --
18 DR. CORRADINI: Okay.
19 MR. ANDERSON: -- metric ton.
20 DR. CORRADINI: All right. Thank you.
21 DR. SCHULTZ: Kenny, how does that -- on 22 the left-hand graph, you have the maximum ratio per 23 the corrosion limit. How has the corrosion limit been 24 set? And how do you monitor it? How do you propose 25 to monitor it, the corrosion?
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21 1 MR. ANDERSON: Yes. We do have an 2 analysis from our fuel vendor, Framatome, that does 3 perform that calculation. They have a limit of 100 4 microns, thank you. And our analysis shows that, 5 under best estimate normal operating conditions, we're 6 roughly at 30 in those units.
7 DR. SCHULTZ: And you'll monitor that with 8 cycle by cycle inspection of the fuel?
9 MR. ANDERSON: Yes, that's -- I'd have to 10 defer to my colleagues to -- but we can come back with 11 that exact answer.
12 DR. SCHULTZ: That would be great. Thank 13 you.
14 CO-CHAIR MARCH-LEUBA: And one thing I 15 read somewhere, and I'm going to say something that 16 it's not supposed to say. I'm not an expert on this.
17 Because we run in natural circulation and 18 very low flow velocities, the heat transfer 19 coefficient out of the clad is higher than normal in 20 the operating fleet. And therefore, the clad is 21 operating hotter than normal in the operating fleet.
22 Does the higher temperature of the clad 23 affect this? And do we have sufficient data to go 24 over that operating condition?
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22 1 question. I'd want to defer to the --
2 CO-CHAIR MARCH-LEUBA: Please keep us on 3 this.
4 MR. ANDERSON: Yeah.
5 CO-CHAIR MARCH-LEUBA: We'll come, I'll 6 make a note to come back this afternoon.
7 MR. ANDERSON: Okay. Sounds good.
8 DR. SCHULTZ: From the documentation, 9 Kenny, it said that you don't have the information 10 related to hydrogen pick-up at this point in time.
11 Does --
12 MR. ANDERSON: Yes, well, we --
13 DR. SCHULTZ: Does Areva have it? Or --
14 MR. ANDERSON: Well, yeah --
15 DR. SCHULTZ: -- they must be looking at 16 it for their own future work.
17 MR. ANDERSON: Sorry to interrupt you.
18 Yes, we're working on, with Areva or Framatome, our 19 partner, to have an official method for calculating 20 it. So it's, we can do it.
21 DR. SCHULTZ: You haven't gotten there 22 yet.
23 MR. ANDERSON: We just haven't proven it 24 yet. And we're working to do that.
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23 1 criterion box is going to be appropriately set below 2 the curve on the left, right-hand side?
3 MR. ANDERSON: Yes, for many reasons, 4 including the low corrosion performance of M5.
5 DR. SCHULTZ: All right. Thank you.
6 MR. ANDERSON: Okay. So, to analyze the 7 event, unlike the other Chapter 15 events, we have a 8 multi, a unique multi-step process that is a -- we 9 usually rely on the nuclear physics code as the, as 10 that comparison table showed. And this is an event 11 dominated by nuclear physics.
12 So that SIMULATE5 is a steady state 13 nuclear physics analysis that sets the initial 14 conditions up to make sure we have the full spectrum 15 of conditions.
16 We then pass that to the transient 17 kinetics code, SIMULATE-3K. And that calculates the 18 power pulse as a function of time, which is kind of 19 the workhorse or the main item in this event that then 20 feeds the other analysis paths.
21 So, as we'll talk about, the one analysis 22 path is to NRELAP5 to do the system pressurization.
23 Then in the middle path, we pass the pressure pulse or 24 the power pulse to end the conditions calculated by 25 RELAP5 to our subchannel analysis code, VIPRE, to do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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24 1 the official CHF evaluation.
2 And then there's also a path at the bottom 3 where, to calculate the fuel enthalpy criteria. We do 4 essentially a hand calculation that calculates and 5 checks those various energies deposited in the fuel.
6 MEMBER REMPE: Can --
7 MR. ANDERSON: And I'll talk about each 8 one of those.
9 MEMBER REMPE: I'm sorry to interrupt you.
10 But I forgot at the beginning of this session that I 11 was supposed to declare that I might have some 12 conflict of interest. And I'm going to have to limit 13 my discussion on this topic. Okay? Thank you.
14 MEMBER BLEY: Are you going to talk about 15 the actual physical event that starts the rod 16 ejection?
17 MR. ANDERSON: No, unless, not really 18 unless you'd like to talk about it.
19 MEMBER BLEY: Well --
20 MR. ANDERSON: It's generally a 21 postulated.
22 MEMBER BLEY: What I want to ask you is 23 it's postulated to happen like that, right?
24 MR. ANDERSON: Yeah, effectively the -- we 25 would --
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25 1 MEMBER BLEY: What if it's a little 2 slower?
3 MR. ANDERSON: Then it's generally --
4 MEMBER BLEY: Is it better if it's slower?
5 MR. ANDERSON: Yes, yes.
6 MEMBER BLEY: In all cases?
7 MR. ANDERSON: Yes.
8 MEMBER BLEY: You've looked enough to be 9 confident of that.
10 MR. ANDERSON: Yes.
11 MEMBER BLEY: Okay.
12 MR. ANDERSON: What makes this event 13 unique from all the other Chapter 15 events is its, 14 how rapid it is. And so the key is understanding to 15 make sure you have that power pulse, and then is that 16 power pulse high enough to basically create a water 17 hammer, which would break the, or it would violate the 18 criteria of reactor coolant pressurization.
19 And then the other aspects of it is, is 20 that power pulse high enough to deposit a lot of 21 energy in the fuel and have the fuel either fail and 22 have that released radiation that would violate the 23 limit, or would it damage the fuel such that it was no 24 longer coolable.
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26 1 is, is it rapid and qualifying that. And then other 2 events, if it's just slower, we'll talk about where we 3 can see that. It's basically benign.
4 CO-CHAIR KIRCHNER: Follow up on that, so 5 just for the record. When you do the RELAP analysis, 6 you don't take any credit for the rupture of the 7 control rod drive mechanism, right? You just, you 8 essentially are doing the analysis as if the system is 9 intact to get your peak pressure.
10 MR. ANDERSON: Yeah, yes.
11 CO-CHAIR MARCH-LEUBA: And while we have 12 you interrupted, on the non-LOCA methodology, which 13 also uses RELAP on VIPRE, we have some criteria, a 14 screening criteria, for RELAP to determine whether we 15 have to go on VIPRE or not. Here you always run 16 VIPRE, correct? You don't have a screening criteria?
17 MR. ANDERSON: We use the criteria, 18 because we run a full spectrum of cases. So we're 19 looking at hundreds of cases. And only a subset are 20 actually run in VIPRE explicitly, because they're 21 screened by RELAP.
22 CO-CHAIR MARCH-LEUBA: Is that 23 proprietary, or can you tell us what the screening 24 criteria is?
25 MR. ANDERSON: No, it's not proprietary.
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27 1 It's generally we have a correlation in RELAP that we 2 test. And the non-LOCA topical talks about that. But 3 it's basically we benchmark the RELAP evaluations 4 calculation of CHF to VIPRE and then, and proven to 5 ourselves that that's a good indicator. But all the 6 --
7 CO-CHAIR MARCH-LEUBA: I want to keep it 8 at a non-proprietary level. But the RELAP CHF 9 correlation is way off.
10 MR. ANDERSON: Well, there's multiple ones 11 depending on the problem to be solved.
12 CO-CHAIR MARCH-LEUBA: So you're not using 13 the one from the non-LOCA? Are you using a more 14 accurate one here?
15 MR. ANDERSON: Well, there's several CHF 16 correlations in the RELAP model we use. And this one 17 is one that is an approximater for our NSP 18 correlation.
19 DR. CORRADINI: We can wait.
20 MR. ANDERSON: Okay. And --
21 DR. CORRADINI: You can tell us to wait.
22 You don't have to be nice to us. We can wait to the 23 proprietary if you want.
24 MR. ANDERSON: Yeah, well, the non-LOCA 25 topical, in particular, will talk --
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28 1 DR. CORRADINI: Okay.
2 MR. ANDERSON: -- more about that.
3 CO-CHAIR MARCH-LEUBA: Okay. We'll wait 4 for the non-LOCA or the proprietary, and I'll ask you 5 again.
6 MS. McCLOSKEY: Megan McCloskey, NuScale.
7 Yes, we can discuss this in the non-LOCA topical 8 report presentation a little bit later this morning.
9 I would also clarify that the screening 10 that we do with NRELAP5 in the, as part of the non-11 LOCA topical report, is only to indicate the trend of 12 which cases are more limiting for CHF. and the margin 13 assessment is all done with VIPRE and the approved 14 NSP4 CHF correlation.
15 DR. CORRADINI: So, since you're up there, 16 can you clarify? The way I look at this is there's 17 three potential criteria. You then have what I'll 18 call a factor of safety. If you approach your factor 19 of safety, if any one of these three, you tag it as 20 something that's got to be looked at more closely.
21 In other words, if MCFR is --
22 MS. McCLOSKEY: I would defer to Kenny --
23 DR. CORRADINI: Just pick a number. If 24 MCFR is within a factor of, if CHF is within a factor 25 of two of what you see, then you tag it so that you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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29 1 look through all three criteria and only redo them 2 with VIPRE once you get close enough.
3 MR. ANDERSON: Yes, so the, for the -- we 4 kind of analyze each acceptance criteria uniquely.
5 DR. CORRADINI: Oh, okay.
6 MR. ANDERSON: So, for the system 7 response, we calculate, or for the pressurization of 8 the system, we just use RELAP. And we make sure that 9 we've got the limiting case.
10 For the CHF evaluation, we send basically 11 up to hundreds of RELAP cases to pick out a family of 12 what appears to be the limiting.
13 DR. CORRADINI: Okay. Fine.
14 MR. ANDERSON: And we want to be robust.
15 So we don't want to have it too small of a group.
16 DR. CORRADINI: Okay.
17 MR. ANDERSON: And then, in the closed 18 session, you'll see some results of that screening 19 process. There's ten candidates. And then you can 20 see how the VIPRE actually says, oh, this is the one 21 that's the most limiting.
22 DR. CORRADINI: Okay. Thank you.
23 MR. ANDERSON: Yeah.
24 DR. CORRADINI: Thank you very much.
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30 1 to say it in the open session. We need it 21st 2 century. How much does it cost to run VIPRE? I mean, 3 why not run them all properly and know what -- you're 4 supposed to do a whole bunch of scenarios for control 5 rod ejection, I mean, number one, number two, number 6 -- just run them all and you know.
7 I mean, I just don't see the -- in 1970, 8 we had to do it with a slide rule. And yes, 9 definitely you had to do approximations. I don't see 10 the reason for doing approximations now. But anyway, 11 that's just recreational complaining.
12 MR. ANDERSON: So the first main step is 13 this, the steady state initiation of a nuclear physics 14 code. And here, we're focused on setting up the full 15 spectrum and making sure we have all the possible 16 equipment configurations at all the possible times in 17 life. And then we also do with this determine the 18 worst rod stuck out, it's part of the fundamental 19 assumption.
20 Then the workhorse, the star of this whole 21 show is the transient kinetics. We, as part of our 22 work and to prove the methodologies is applicable, 23 this design and the code used is we benchmarked this 24 SPERT-III and the NEACRP computational benchmark.
25 DR. CORRADINI: So let me ask about SPERT-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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31 1 III is back in the 60s, is it not?
2 MR. ANDERSON: I believe the 60s or 70s, 3 I'm --
4 DR. CORRADINI: So those are integral 5 measurements. So you're benchmarking off of integral 6 measurements and you're getting detailed measurements 7 from the computational benchmark? Because SPERT 8 didn't measure anything other than overall power 9 increase based, they didn't do any sort of spatial 10 analysis that you can benchmark.
11 MR. ANDERSON: Correct. To my knowledge, 12 correct, we benchmarked the overall power pulse.
13 DR. CORRADINI: Fine.
14 MR. ANDERSON: And shape and worths.
15 DR. CORRADINI: Thank you.
16 MR. ANDERSON: Yes. Then also in this 17 part of the analysis, we applied uncertainties in the, 18 we do reference a topical, the nuclear analysis 19 topical. Generally that presents our method for 20 calculating the uncertainties. And then each time we 21 apply this method, we do ensure that our uncertainties 22 that we're using are valid.
23 DR. SCHULTZ: Kenny, what you present here 24 are your, what you call your major uncertainties?
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32 1 uncertainties, which are applied.
2 MR. ANDERSON: Correct, these are just the 3 uncertainties for this portion of the analysis.
4 DR. SCHULTZ: I see.
5 MR. ANDERSON: And then at the, in one of 6 the slides there's a stack-up that shows all the 7 uncertainties and where they're treated.
8 CO-CHAIR KIRCHNER: While that's up, which 9 of those uncertainties did you see had the most 10 effect? I have my own idea which one it is, but could 11 you go through those delayed neutron fractions? Rod 12 worth's obvious, Doppler is the, what's terminating 13 the event. And --
14 MR. ANDERSON: Yeah, I don't have a 15 specific quantification. Generally, delayed neutron 16 fraction, the static worth, and the Doppler 17 coefficients are extremely significant. But we could 18 -- I don't have a exact quantification.
19 CO-CHAIR KIRCHNER: Okay.
20 CO-CHAIR MARCH-LEUBA: And this is in the 21 event that you have to evaluate the power, because 22 that typically is the one that becomes worse. And 23 those conditions you can have such high boron 24 concentration that day, the moderator temperature 25 coefficient can be positive. Is that correct? Or do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 1 you always consider it negative?
2 MR. ANDERSON: Well, we, as we'll see in 3 the closed session, we have some example results. And 4 we'll explain why for our design it's actually middle 5 powers is the most limiting. And so I can --
6 CO-CHAIR MARCH-LEUBA: It may not be, if 7 this has serial power with a positive boron 8 coefficient.
9 MR. ANDERSON: I guess I'd have to --
10 CO-CHAIR MARCH-LEUBA: If your boron 11 coefficient were to be positive, very positive, that 12 --
13 MR. ANDERSON: I'll have to check. I 14 don't think that that situation is possible.
15 CO-CHAIR MARCH-LEUBA: We'll wait, I'll 16 wait. I'll make another note for the closed session.
17 MR. ANDERSON: Okay.
18 CO-CHAIR MARCH-LEUBA: I see that I 19 started this morning that we may have many time for 20 tomorrow.
21 MEMBER PETTI: Question, there's burnable 22 poisons in the design?
23 MR. ANDERSON: Yeah, we use gadolinium in 24 the current design.
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34 1 evaluation using the subchannel code VIPRE. And we 2 rely on our subchannel topical, but there are a few 3 deviations from that. This is the unique event, it is 4 very, the power pulse is very quick. It's on the 5 order of milliseconds.
6 And so making sure that the finite 7 difference code is being solved correctly requires 8 smaller time steps. And then with the current number 9 that involves ensuring that the axial nodalization is 10 appropriate. So we do a set of sensitivity cases to 11 make sure that we are using the code appropriately.
12 DR. SCHULTZ: Kenny, in that regard, 13 you've done these things prior to doing the analysis 14 to demonstrate that your analyses is identifying?
15 MR. ANDERSON: Yeah, I would say the 16 answer's both.
17 DR. SCHULTZ: You're not doing these 18 sensitivities for every evaluation that you do.
19 MR. ANDERSON: The answer's both. We do 20 a set of sensitivities to make sure we understand and 21 the methods is good and appropriate. But because of 22 the unique nature of this event, we do specific 23 sensitivities to ensure that what we expect to happen 24 in terms of the codes --
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35 1 in general to understand it, but when you have a 2 limiting case evaluation, you're also going through 3 the sensitivities to demonstrate that you've got, 4 you've nailed it computationally.
5 MR. ANDERSON: Yes.
6 DR. SCHULTZ: Okay, thank you.
7 CO-CHAIR MARCH-LEUBA: Can, remind me 8 again because we saw this two years ago. When you do 9 the RELAP calculation, you do a full core, one, every 10 single channel you calculate the power. When you do 11 the VIPRE calculation with those boundary conditions, 12 how many rods do you model?
13 MR. ANDERSON: In this case, it's 52 rods.
14 CO-CHAIR MARCH-LEUBA: So you do only a 15 fraction of each bundle.
16 MR. ANDERSON: Yeah, the bundle of, or the 17 assembly where the actual, we postulate the rod to be 18 ejected, every rod in that assembly is fully detailed.
19 And then as we get further away, there's progressive 20 lumping, such that the furthest away is --
21 CO-CHAIR MARCH-LEUBA: I remember the maps 22 when we saw. Roughly there are 52 representative 23 regions, and one to one for the rods that are closer 24 to the control rod? Or do you use pin power 25 reconstruction to identify three rods in the fuel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 bundle is the worst, or something like that?
2 MR. ANDERSON: The transient kinetics does 3 do pin power reconstruction, and we do use that to 4 inform the power distribution that we use --
5 CO-CHAIR MARCH-LEUBA: So you run RELAP, 6 you get the average power of the bundle. Using pin 7 power reconstruction, you get the power in every rod.
8 And that, then you confirm that the distribution 9 around the control is adequate. Is that more or less 10 what you're doing?
11 MR. ANDERSON: More or less. But yeah 12 then, we get the full power distribution from 13 SIMULATE-3K, and then we, yeah, then like you said we 14 apply that in the VIPRE.
15 CO-CHAIR MARCH-LEUBA: Thank you.
16 MR. ANDERSON: As we talked about, one of 17 the unique features of this event is the fuel 18 criteria. And all -- most of them use an indicator of 19 calories per gram to help indicate either fuel failure 20 or core coolability.
21 And to satisfy that, we do a very 22 conservative adiabatic heat-up hand calculation, where 23 we essentially put all the energy into the pallet and 24 calculate the calories per gram as a result of the 25 power pulse calculated from SIMULATE-3k.
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37 1 DR. SCHULTZ: This is the calories per 2 gram as well as fuel melting is from this? And the 3 fuel melt evaluation is also with this analysis?
4 MR. ANDERSON: Yes, we do both the 5 calories per gram and the fuel centerline melt.
6 DR. SCHULTZ: These are the conservatisms 7 that are applied to the evaluation, I'll call it, of 8 fuel melt.
9 MR. ANDERSON: Correct.
10 DR. SCHULTZ: Conservative in that regard.
11 Thank you.
12 MR. ANDERSON: Okay, the next part we can 13 talk about is the system model. And there's two 14 parts, so this is part one. This is the part where we 15 evaluate CHF, and as we talked about previously, this 16 is primarily to, well, to screen the cases for VIPRE.
17 And as part of that it also looks at the 18 boundary conditions that's happening in the whole, the 19 whole module, the inlet flow and the exit pressure to 20 pass those to VIPRE to ensure VIPRE has all that 21 information.
22 The SIMULATE-3K doesn't do the, doesn't do 23 a boundary. It assumes boundary conditions, it 24 doesn't calculate that. But the event is quite rapid, 25 and that's so it justified that.
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38 1 The other part, there's another aspect 2 that RELAP is used for, and that's for the system 3 pressurization. For us, the, part of the intent of 4 GDC 28 is to prevent the very high integral energy 5 that the, vaporizes the steam and can create a water 6 hammer. But with our very low integrated powers, what 7 actually happens is the system in the classical peak, 8 the power pulse type event, the energies are just too 9 low.
10 And so you don't see a system 11 pressurization response. So we do have a more 12 traditional non-LOCA type analysis to have a pulse 13 that's just high enough to be below, to increase the 14 power, but it's below the high power trip set point.
15 And then eventually the system will heat up and 16 pressurize, and then the system will trip on the high 17 pressure. And then we show that that does stay below 18 the high pressure acceptance criteria.
19 DR. CORRADINI: You're going to say that 20 all over again, you've lost me. I thought I'd got 21 you, but then as you started proceeding you lost me.
22 Can you try that one more time?
23 MR. ANDERSON: Yes, there's the classical 24 rod ejection, which, you know, GDC 28 is really 25 designed to protect, is there's this extreme power NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 1 pulse. And that extreme power pulse can vaporize the 2 coolant.
3 DR. CORRADINI: Right.
4 MR. ANDERSON: And we've, we want to 5 protect against. But if the power pulse is so low 6 that there's not that much energy to increase the 7 system coolant, then that's not actually the limiting 8 pressurization.
9 The limiting pressurization looks like a 10 traditional non-LOCA event, where you increase the 11 power just enough to be below the set point and have 12 it sit there and slowly pressurize the system. And 13 then eventually the system pressure is high enough to 14 hit the system pressure set point.
15 MEMBER PETTI: So it's like a transient 16 overpower.
17 MR. ANDERSON: Yes.
18 MEMBER PETTI: In some of the other 19 reactors.
20 MR. ANDERSON: Yes.
21 MEMBER PETTI: And let me just, so I think 22 I know, why is the worth so low?
23 MR. ANDERSON: It's the static, it's part 24 of the design, the static worths are just low.
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40 1 you're running at power and the rods are pretty much 2 all out. So when you eject a rod that's already well 3 withdrawn, you're just not putting that much worth 4 into the system. And so you just get a pressurization 5 as a result.
6 MEMBER PETTI: I figured it's the low 7 initial stored energy of this. These are half the 8 linear power of a large light-water reactor. And all 9 this burnable poison is holding down any excess of 10 water in it, right?
11 DR. CORRADINI: So you're searching for --
12 I guess I was -- the way Dave explained it, is that 13 correct? Because that's how I was guessing is the 14 reason.
15 MR. ANDERSON: Yes, that's --
16 MEMBER PETTI: From a non-physicist view 17 --
18 MR. ANDERSON: No, yeah, that's exactly 19 right. Yeah, if you look at a reactor design very 20 different from ours, like SL1, it had a very --
21 DR. CORRADINI: Bad example.
22 MR. ANDERSON: Well, it had a very high, 23 it had very high static worths, and that's basically 24 the opposite of us. So we have very low static 25 worths, and a key input that is where we have the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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41 1 power and searching limits.
2 DR. CORRADINI: And we'll see some 3 representative numbers in closed session.
4 MR. ANDERSON: Yes.
5 DR. CORRADINI: Okay.
6 DR. SCHULTZ: Kenny, in the draft Reg 7 Guide 1327, the guide sets limits also for the 8 potential for pellet clad interaction looking at 9 differential pressure between system and internal 10 pressure of the fuel. Do you do evaluations of that, 11 of the pressure versus the internal pressure of the 12 fuel to see whether you match up with the assumptions 13 that are in the reg guide -- draft reg guide, excuse 14 me?
15 MR. ANDERSON: Oh, I was going to say we, 16 as we saw in that first table, we do, those are some 17 of the criteria, the differential pressure criteria, 18 where we used the calories per gram as an indicator.
19 DR. SCHULTZ: So you're looking at it to 20 match up with the calorie-per-gram levels that are 21 chosen for that draft reg guide. You're not doing the 22 calculation of differential pressure to determine 23 whether you --
24 MR. ANDERSON: Correct, yeah, we --
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42 1 problem there.
2 MR. ANDERSON: We use, we calculated the 3 calories per gram, and we used that as an indicator 4 for the internal pressure and the differential 5 pressure.
6 DR. SCHULTZ: Yeah, I'm just curious as to 7 whether, when you do all these calculations, whether 8 you see situations where that differential pressure 9 criteria would in fact come into play.
10 MR. ANDERSON: And generally no. What 11 limits, as we'll see in the closed session, what 12 limits us is the CHF evaluation.
13 DR. SCHULTZ: Okay, that's what I was 14 looking for. Thank you.
15 MEMBER PETTI: So a typical -- I'm sorry.
16 That's all right, no, you go ahead.
17 MEMBER BLEY: Okay, go back two, one or 18 two slides where you have the picture. There, that 19 one. When we make these things up and then analyze in 20 great detail the neutronics and other parts of what's 21 going on, what about the kind of larger, gross things?
22 Exactly where do we assume the break occurs that 23 allows the rod to be ejected?
24 MR. ANDERSON: Will it be at the --
25 MEMBER BLEY: Can you show it on there?
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43 1 MR. ANDERSON: Oh, sure. So the, this is 2 the reactor vessel.
3 MEMBER BLEY: And those are the mechanisms 4 sticking out --
5 MR. ANDERSON: Yeah, and then so then 6 there would be a failure --
7 MEMBER BLEY: Right around there.
8 MR. ANDERSON: Yeah, and then a --
9 MEMBER BLEY: Okay. So somehow that 10 happens and it falls off.
11 MR. ANDERSON: It was postulated --
12 MEMBER BLEY: And where's the rod go?
13 MR. ANDERSON: For this analysis, it --
14 MEMBER BLEY: Disappears.
15 MR. ANDERSON: It disappears, yes, thank 16 you.
17 MEMBER BLEY: So where's the rod go, and 18 can it -- what happens when it hits the top of the 19 reactor vessel and maybe the containment? I mean, 20 we're doing all this fancy analysis on a really 21 hypothetical event, and yet the --
22 MEMBER BALLINGER: The rod is about 30 or 23 40 feet long.
24 MEMBER BROWN: Like a spear.
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44 1 boundary is way up there.
2 MEMBER BLEY: It ain't way up in this one.
3 I mean, it's way up in other plants.
4 MR. ANDERSON: Yeah, as part of the, and 5 we can bring in our ASME guys, well we --
6 MEMBER BLEY: It just goes away.
7 MR. ANDERSON: Well, in general it's not 8 an ASME event that the, we don't use j-groove welds or 9 anything like that. And so they, we don't analyze 10 turbine missiles -- or I mean, not turbine, missiles.
11 MEMBER BLEY: No, you don't.
12 MR. ANDERSON: Yeah, and the reason we 13 don't do that is because the ASME, you know, rules and 14 regulations for that. But we can get more detail on 15 that.
16 MEMBER RICCARDELLA: You say you don't use 17 j-groove welds in the top bay?
18 MR. ANDERSON: Correct.
19 MEMBER RICCARDELLA: How are those nozzles 20 attached, full pen loop?
21 MR. ANDERSON: I'd have to -- yeah, it's 22 part of the integral to the -- well, I'd have to get 23 back on that, the exact.
24 CO-CHAIR KIRCHNER: But, or the ends are 25 bolted flange connections.
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45 1 PARTICIPANT: That's what it looks like.
2 MEMBER RICCARDELLA: But the attachment, 3 yeah, but it's, that's bolted at the other end of the 4 penetration. There's got to be an attachment to the 5 vessel head, right. We --
6 MR. ANDERSON: Yeah, I'm just going to 7 make sure I bring back the best people to talk about 8 that.
9 MEMBER RICCARDELLA: Yeah, but I have a 10 more general question, is aren't there other ways to 11 eject a rod other than a pressure boundary failure?
12 I mean, I thought like in a regular PWR, you've got 13 these mechanisms that can fail, or in a BWR you can 14 drop the rod without necessarily failing the pressure 15 boundary.
16 MR. ANDERSON: Yeah, in --
17 MEMBER BALLINGER: These are fine motion 18 control rods, I think. And so the failure, when you 19 lose power it clamps them on. They can't move. So 20 it's sort of like the navy.
21 MEMBER RICCARDELLA: Yeah, but if a --
22 MR. ANDERSON: One of the things to 23 remember, too, is it's not necessarily how much the 24 rod moves from beginning to end, it's how quickly the 25 rod moves. And it's the quickness of the rod moving NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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46 1 when you involve the delayed neutron fraction that 2 makes the dynamic worth so important.
3 MEMBER RICCARDELLA: But that being the 4 case then, doesn't this involve a LOCA? If you break 5 the penetration --
6 MR. ANDERSON: Yes, it does.
7 MEMBER RICCARDELLA: Oh, so your analysis 8 does assume that a LOCA, that there is a LOCA.
9 CO-CHAIR KIRCHNER: But they don't take 10 credit for the leakage.
11 PARTICIPANT: It evaporates. The control 12 rod disappears --
13 PARTICIPANT: It disappears.
14 PARTICIPANT: In a certain amount of time.
15 DR. CORRADINI: It's a bounding 16 calculation.
17 MEMBER PETTI: Yeah, it's a bounding 18 calculation.
19 MEMBER RICCARDELLA: I understand, but it 20 seems to me there might be ways that the teeth on the 21 mechanism break or something, and you eject a rod 22 without necessarily failing the penetration.
23 MEMBER PETTI: We could ask our risk 24 experts. I don't think this is a risk-dominant event, 25 you just don't see it.
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47 1 CO-CHAIR KIRCHNER: No, and in NuScale's 2 favor in this design, the extension is quite 3 substantial and the weight. So there's no way for the 4 rod to disappear, short of having something like a 5 failure of the control rod drive.
6 MEMBER BROWN: Then why couldn't, I mean, 7 you've got a vacuum in the containment vessel or 8 partial vacuum. You've got lots of pressure in the 9 reactor pressure vessel. You eject a rod that seems 10 I've got considerable forces driving it up. Why 11 couldn't it penetrate the containment vessel?
12 It's not even addressed. In fact, as the 13 SER from the staff says --
14 DR. CORRADINI: I wouldn't want to address 15 it.
16 MEMBER BROWN: How would you want it 17 addressed?
18 DR. CORRADINI: Why would you want to 19 address it?
20 MEMBER BROWN: I don't know.
21 DR. CORRADINI: It's a bounding 22 calculation to show that from a reactivity and 23 insertion standpoint, you have no problem.
24 MEMBER BROWN: I'm talking about the LOCA 25 --
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48 1 DR. CORRADINI: If you get mechanistic, 2 you're never going to get close to this.
3 MEMBER BROWN: I'm talking about the LOCA 4 situation that this SER even says you could cause a 5 LOCA with this. And you're, all the, how do you cool 6 it after that? All the coolant, just said that 7 disappears from the ability to cool the core, if you 8 look at their pictures. And it's not covered in the 9 LOCA, it's not covered in the LOCA TR.
10 MR. ANDERSON: Well, I think it is as part 11 of the LOCA methodology.
12 MEMBER BROWN: I just looked, I looked up 13 rod ejection and it's not in there.
14 MR. ANDERSON: And we --
15 MEMBER BROWN: Doesn't mention it.
16 MR. ANDERSON: Yeah. We could have the, 17 as part of the, or we could do it anytime you want, 18 but we could have the LOCA people speak to that.
19 MS. McCLOSKEY: Megan McCloskey, NuScale 20 again. The control rod ejection's not part of the 21 LOCA topical report, as you identified. It was 22 considered with respect to the containment pressure 23 response analysis. But in the NuScale design, the 24 liquid space breaks are much more limiting compared to 25 the vapor space breaks, which would be from the top of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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49 1 the reactor pressure vessel.
2 And the amount of energy that's added due 3 to the power pulse is small compared to the thermal 4 energy of the fluid initially in the RCS that drives 5 the containment pressurization response. So it's also 6 non-limiting for peak pressure temperature.
7 MEMBER BROWN: That part I understand, I 8 guess my question is really related to the fact why 9 isn't the velocity of this ejection that it's being 10 driven from the core, 2000, whatever the pressure is 11 down in the reactor pressure vessel, the vacuum up 12 there, but why can't it drive itself right through the 13 containment vessel. And now I've got a complete 14 breach of the containment --
15 MR. ANDERSON: But we can provide a --
16 MEMBER BROWN: And that's not discussed in 17 any of the documents that I could find.
18 MR. ANDERSON: Yeah, we can provide an 19 answer for the missile, the missile question.
20 MEMBER RICCARDELLA: That's what it is, 21 it's a missile question.
22 CO-CHAIR MARCH-LEUBA: You can and -- you 23 can or you will?
24 MR. ANDERSON: We will, sorry. I can't at 25 this moment articulate it.
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50 1 DR. CORRADINI: I just think we're going 2 down a rabbit hole that we don't want to go down.
3 MEMBER BROWN: I don't agree with that, 4 Mike.
5 DR. CORRADINI: Well, I do --
6 MEMBER BROWN: I mean, you've got the CHF 7 --
8 DR. CORRADINI: That's just the 9 consultant's opinion, it's a rabbit hole.
10 MEMBER BROWN: But then you've got, can 11 you breach, actually breach the containment vessel.
12 Why isn't that something to think about?
13 DR. CORRADINI: Then you'd have to do a 14 mechanistic acceleration, which is going to be small 15 because the depth, the distance between the time where 16 it is and to where it's going to hit is so small you 17 probably wouldn't hit it for, with a few meters a 18 second. Therefore, it probably won't attack, it won't 19 hurt the containment at all.
20 MEMBER BROWN: Huge differential pressure.
21 DR. CORRADINI: But if you make it 22 mechanistic, Charlie, then the power pulse itself will 23 be much smaller than what they're assuming. It's a no 24 nevermind in either case, but if you want to make a 25 consistent calculation with what Dennis I think is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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51 1 driving towards, Dennis started it, which is --
2 MEMBER BROWN: If he hadn't I would have, 3 so.
4 DR. CORRADINI: Is that you would probably 5 have a very small power rise, very small power rise.
6 MEMBER RICCARDELLA: Yeah, but forget 7 about the power run, just hole in the containment.
8 Just having a LOCA that -- and then having something 9 that pierces the containment is a concern, independent 10 of the power surge.
11 MEMBER BROWN: Exactly. To me it looks 12 like you're looking at the trees.
13 MEMBER RICCARDELLA: No, I would suspect 14 you could do the analysis and you show, you could show 15 that that's not going to penetrate.
16 MEMBER BROWN: Well, that would be fine.
17 I just --
18 MR. ANDERSON: We have considered it and 19 we do have an answer, I just, I don't know, and I'll 20 have to work on my colleagues to get that answer for 21 you.
22 CO-CHAIR MARCH-LEUBA: So let me go back 23 to my question. When will we see that analysis?
24 MR. ANDERSON: Oh, we have an answer, 25 we'll have an answer for you today.
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52 1 CO-CHAIR MARCH-LEUBA: So you have 2 somebody in Corvallis working on it? My people will 3 work with your people.
4 MEMBER RICCARDELLA: But you're saying 5 there's no other way -- I guess it doesn't really 6 matter. You're saying there's no other way to eject 7 the control rod other than to violate the pressure 8 boundary. And the fact that you didn't assume you 9 violate the pressure boundary is conservative for the 10 purposes of this analysis, is that what you're saying?
11 MR. ANDERSON: Yes, generally, we 12 postulate this as a bounding method for evaluating the 13 GDC 28 criteria.
14 CO-CHAIR MARCH-LEUBA: So what I hear the 15 members say is that from the point of view of 16 neutronics and the power peak, you're likely 17 conservative. I mean, we cannot, we don't see where 18 you are non-conservative. However, there is a 19 possibility of a missile inside the containment that 20 you are going to tell us this afternoon why it's not 21 a problem.
22 MR. ANDERSON: Correct.
23 DR. CORRADINI: For this one person, if 24 you're going to make them do this, then you're going 25 to have to ask the question is does it leak, or does NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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53 1 it actually sever. Because it strikes me if I start 2 worrying about a control rod drive mechanism opening 3 up, I would expect it to leak, not sever and shoot 4 out.
5 MEMBER RICCARDELLA: Well, we came pretty 6 close to severing it at Davis-Besse, so I mean.
7 MEMBER BALLINGER: Except for the fact 8 that now we have all new materials and a lower 9 temperature.
10 MEMBER RICCARDELLA: I understand all 11 that, but --
12 MEMBER BALLINGER: And a few other minor 13 details.
14 MEMBER RICCARDELLA: But the other 15 question I'd like answered is what are those 16 attachments of the control rod drive nozzles to the 17 head if they're not j-groove welds?
18 MR. PRESSON: Those are full penetration 19 welded.
20 MEMBER RICCARDELLA: Full penetration 21 welded, okay, thank you. That should solve the 22 problem.
23 MEMBER BALLINGER: That's another minor 24 detail.
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54 1 understand from them is they have evaluated two 2 limiting cases. One is the rod moves at speed and the 3 issue of power pulse. The other one is the rod heats 4 the containments. We're going to see that this 5 afternoon.
6 MR. ANDERSON: We'll provide the 7 evaluation for the missile.
8 CO-CHAIR MARCH-LEUBA: So you've done two 9 complete independent bounding analysis and both of 10 them are good.
11 MR. ANDERSON: Correct.
12 CO-CHAIR MARCH-LEUBA: And none of the two 13 are credible within some --
14 MEMBER BROWN: I'm not arguing 15 credibility, it's just a rod ejection, we're required 16 to evaluate it, it has a certain end result, can that 17 end result be a problem? And that's all.
18 CO-CHAIR MARCH-LEUBA: I'm going to write 19 it down for my closed session. Ask Charlie to ask the 20 question again.
21 MR. ANDERSON: So in summary this is the 22 overall approach to analyzing the event, and which we 23 then provide input to the Chapter 15. And in the 24 closed session there'll be example inputs and output 25 values and show how the uncertainties stack up.
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55 1 CO-CHAIR KIRCHNER: Okay, any other 2 questions on rod ejection in the open session? We've 3 got notes for the closed session.
4 MR. BRISTOL: This is Ben Bristol with 5 NuScale. The RAI 9647, we specifically address the 6 difference between the postulated event for GDC 28 and 7 the CRDM, the postulated CRDM failure, which is not a 8 design-basis failure.
9 So the dynamic effects of that event are 10 not part of the design basis, up to and including 11 considerations of if the shaft were to accelerate to 12 the containment head.
13 PARTICIPANT: What does that mean, it's 14 not part of the design basis?
15 MR. BRISTOL: It is not part of the design 16 basis. So the, a failure of the CRDM pressure 17 boundary is part of the RPV, is considered part of the 18 RPV pressure boundary per the ASME specifications.
19 MEMBER RICCARDELLA: But you're saying you 20 don't postulate that as a potential LOCA.
21 MR. BRISTOL: That's correct. The dynamic 22 effects of the failure of the CRDM is not considered 23 as part of the design, design basis.
24 MEMBER BROWN: You don't have to worry 25 about it penetrating the containment, in other words.
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56 1 MR. BRISTOL: That's correct. It is 2 designed to ensure that it will not fail as part of 3 the reactor pressure boundary.
4 CO-CHAIR MARCH-LEUBA: Wait, wait, wait, 5 how is it designed and you haven't analyzed it? I 6 mean, you're saying it's not a design-basis event.
7 I'm not going to bother with analysis of this missile 8 inside my container.
9 MR. BRISTOL: So the components are 10 designed per the ASME specification so that they do 11 not need to be considered as postulated failures.
12 MR. PRESSON: So that's part of the full 13 weld --
14 MEMBER RICCARDELLA: Are you saying 15 because it's a vessel, not a pipe? I mean we --
16 MR. BRISTOL: That's correct.
17 MEMBER RICCARDELLA: Design piping to ASME 18 code as well. We postulate failures of piping. This 19 mechanism up to that flange that attaches the, it 20 seems more like a pipe to me than a vessel. But.
21 DR. CORRADINI: My suggestion is wait till 22 closed session.
23 CO-CHAIR MARCH-LEUBA: I think the thing 24 is wait until Chapter 15 when they talk about LOCAs.
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57 1 LOCAs, I think they'll be able to demonstrate that a 2 failure of that part of the pressure boundary is not 3 the limiting case for them.
4 CO-CHAIR MARCH-LEUBA: Yeah, but that, 5 there's a difference between local depressurization 6 concerns and the concern is the missile.
7 PARTICIPANT: And the missile's a separate 8 topic, yeah.
9 MEMBER BROWN: This is a totally different 10 design than all the rest of them, the containment 11 that's just sitting right there adjacent with 12 everything else. It's a very closely coupled system, 13 so I don't understand why you make the comment that 14 it's not part, you don't have to consider it as part 15 of the design.
16 It's like telling me I don't have to worry 17 my reactor protection system because I design it not 18 to fail. Jesus, you know, when am I going to go work 19 on that one?
20 CO-CHAIR MARCH-LEUBA: That's what they 21 tell me when I complain about ATWS. It's not going to 22 happen.
23 MEMBER RICCARDELLA: Historically, in 24 nuclear design, we've made the assumption that pipes 25 can fail, and we design for LOCAs that are failures of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 1 pipes. And, but the reactor vessel itself, we draw 2 the line there and we say that, we're not going to 3 postulate failure of the reactor vessel.
4 But to me, it's skinning the cat pretty 5 thin to say well, I got this long tube that sticks out 6 of the vessel, but we don't have to assume that fails 7 because it's part of the vessel. I mean, that's a 8 very legalistic interpretation of something that I 9 think was beyond the intent of that. But we can talk 10 about it more later.
11 CO-CHAIR MARCH-LEUBA: But let me be --
12 but the summary of the rod ejection accident, rod 13 ejection is considered here as a neutronic power 14 peaking effect. So we have to take a note to address 15 this issue. And just because we're talking about the 16 methodology for LOCA, that doesn't include missile 17 this afternoon. We need to address it for March 2 18 when we're talking Chapter 15 about --
19 MS. PATTON: I'd just like to point out 20 there's -- I can give you guys a couple of references 21 where this is addressed in the DCA SER. So you'll 22 find the missile section with a discussion of this in 23 Section 3.5.1.1 and then a discussion with respect to 24 return to power.
25 This is also addressed. It was a UOI.
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59 1 That's in 15.0.6.4.3. But you're right. It's a 2 appropriate for the DCA section, not the 3 methodologies.
4 CO-CHAIR MARCH-LEUBA: Do you understand 5 the issue that the members are having?
6 MS. PATTON: Yeah.
7 CO-CHAIR MARCH-LEUBA: We are worried that 8 because there's an initiation then in which the they 9 are all fails and that causes a failure of the 10 containment.
11 MS. PATTON: That's correct. What I am 12 telling you is what NuScale said is correct and you'll 13 find sections in the SE where it writes up --
14 CO-CHAIR MARCH-LEUBA: Well, 3.5.1.1 15 should tell me something about --
16 MS. PATTON: 3.5.1.1 will give you the 17 missile information and then there's also --
18 CO-CHAIR MARCH-LEUBA: It's not going to 19 tell me what they told me, that I don't want to 20 consider it because it's --
21 MS. PATTON: It's going to tell you that 22 it's -- that it's not considered as a -- as a failure.
23 CO-CHAIR MARCH-LEUBA: So we may have to 24 revisit our Chapter 3 letter. You disagree?
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60 1 write-up and it's in there from what you guys saw 2 previously. So --
3 MR. PRESSON: And the RAI 9647 does ask 4 you pretty much this exact question. So having that 5 question and that response.
6 CO-CHAIR MARCH-LEUBA: Yeah. The question 7 -- can anybody --
8 PARTICIPANT: What's the RAI?
9 MR. PRESSON: 9647.
10 CO-CHAIR MARCH-LEUBA: Chris, can you get 11 us a copy of the RAI 9647?
12 Has anybody analyzed this missile? I 13 mean, the way that Dr. Rena (phonetic) says there is 14 no way this missile has enough time to accelerate to 15 break anything.
16 DR. CORRADINI: Well, why don't you ask 17 the question of current light water reactors required 18 to analyze what you're asking? I think the answer --
19 the answer is no.
20 CO-CHAIR MARCH-LEUBA: I am asking is this 21 reactor safe or not?
22 MEMBER RICCARDELLA: Containment is 50 23 feet away.
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61 1 could be 50 feet away or five feet away. It's going 2 to be the same impact in terms of kinetic energy.
3 CO-CHAIR MARCH-LEUBA: Well, my point is 4 Mike's hand calculation says that it's not going to 5 fail anything. But have you guys done anything to it 6 or you just said we don't have to do it? It seems to 7 me that you don't have to do it.
8 MR. PRESSON: I'll have to get back. So 9 --
10 CO-CHAIR MARCH-LEUBA: If you can think 11 about it for the closed session and tell us.
12 MR. PRESSON: Yeah.
13 CO-CHAIR KIRCHNER: Okay. At this point 14 then, Matthew, then are you and Kenny going to do the 15 next presentation or you have to change out people?
16 MR. PRESSON: I have to change out some 17 people so --
18 CO-CHAIR KIRCHNER: Okay. Well, why don't 19 you go ahead and do that and --
20 CO-CHAIR MARCH-LEUBA: Let me remind you 21 that you guys are responsible for your own name tags.
22 Are you going to have one tomorrow?
23 CO-CHAIR KIRCHNER: Okay. Thank you, 24 Kenny.
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62 1 reminding me that we are a little behind so -- and 2 it's our fault. So it was our fault.
3 MR. PRESSON: All right. So we are now 4 here to discuss our loss of coolant accidents 5 evaluation model methodology. I am sensing today will 6 be Dr. Pravin Sawant, supervisor of code validation 7 methods, Dr. Selim Kuran, who is our thermal hydraulic 8 analyst, and we left off Ben Bristol, our supervisor 9 within hydraulic methods. So but he will presenting 10 as well.
11 Start and pass it off to Pravin.
12 DR. SAWANT: So in this open session on 13 LOCA evaluation model topical report we will just try 14 to provide a overview of the methodology. Basically 15 our objective will be to show, like, how methodology 16 follows Reg. Guide 1.203 guidance.
17 We will visit each element of Reg. Guide 18 1.203 and provide a overview of how methodology is 19 addressing those elements and more specificities will 20 be eventually discussed in closed session.
21 So next slide. Yeah. So progression of 22 LOCA in our design is simpler in part because of 23 innate features of the design. With all the RCS 24 complements contained within RPV the potential for 25 number of breaks and the size of breaks is limited, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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63 1 and the coolant is simply captured in containment, 2 cooled, and returned back to the RPV, and the 3 phenomena -- familiar phenomena that are in all to 4 model LOCA progression are well known and that makes 5 the modeling LOCA straightforward.
6 We are following, as I said, Reg. Guide 7 1.203 guidance and the methodology is in compliance 8 with the applicable continuities in 10 CFR 5046 9 Appendix K. So --
10 CO-CHAIR KIRCHNER: Pravin, can I ask you 11 to speak out a little louder and pull your microphone 12 closer. Yeah.
13 DR. SAWANT: Okay.
14 CO-CHAIR KIRCHNER: You've made an 15 assumption that --
16 MEMBER BLEY: That microphone is not 17 turned on. You can turn it on by pushing -- nearer to 18 you. There you go.
19 CO-CHAIR KIRCHNER: It should have a green 20 light.
21 DR. SAWANT: Okay.
22 CO-CHAIR KIRCHNER: Can we go back one 23 slide? You made a -- you postulated something.
24 You're correct. Your design has less extended piping 25 runs and such. But you went -- you kind of said it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 1 simpler to analyze but yet the third sub-bullet, the 2 CNB heat transfer, that actually is an area that's not 3 simpler. So we will come back and visit -- revisit 4 that with you.
5 DR. SAWANT: Yes. So yeah, it's -- we are 6 calling it simpler because what we saw was, like, the 7 heat transfer mechanisms and all are well understood 8 -- that it's been condensation, conduction, and 9 convection to be -- convection to cool on external 10 surface. These are the -- these mechanisms are 11 understood from first principles.
12 CO-CHAIR KIRCHNER: Yes, but let me say it 13 this way. Taking NRELAP5 and ringing it up to do that 14 analysis is not straightforward and the code wasn't 15 designed for that. So it does actually challenge you 16 in the application of the code to analyzing these 17 phenomena. I just want to make that observation.
18 DR. SAWANT: And we have --
19 CO-CHAIR MARCH-LEUBA: Let me help you a 20 moment. We have -- you have performed a number of 21 verification tests that has a benchmark in RELAP and 22 that confirmed the condensation correlation, which I 23 think is proprietary -- the name of it. But there is 24 a condensation correlation in RELAP that has been 25 verified against data. Is that correct?
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65 1 DR. SAWANT: Yes. So we have NuScale's 2 experimental data to confirm that. We have to be 3 careful. We have to model the same heat transfers.
4 We will talk about this.
5 CO-CHAIR MARCH-LEUBA: And so they agree 6 with you all that that was a difficult problem and 7 that's why they spend a lot of money on tests.
8 CO-CHAIR KIRCHNER: Yeah. Yeah. You're 9 making something simpler than it actually was for you.
10 You had to put a fair amount of effort into both your 11 methodology and your benchmark experiments that you 12 ran out in Corvallis.
13 DR. SAWANT: We do have extensive database 14 that we will talk about and when you look at the 15 report and this difference of the phenomena we will 16 see that these are not new and in time I will brief 17 you. So we will talk about that in more detail in 18 closed sessions.
19 So, yeah, 10 CFR 50.46 provides the 20 acceptance criteria policies as designed in light 21 water reactors and we are all aware of these 22 acceptance criteria.
23 What we saw that in our design peak clad 24 temperature is this will be the normal operating clad 25 temperature on the steady state and core during LOCA NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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66 1 doesn't experience ACHIP (phonetic) and it remains 2 covered and we don't see clad heat up.
3 So we ended up using the much conservative 4 acceptance criteria for LOCA, which are listed there 5 as the sub-bullets for the last bullet that core 6 remains covered and we show that by comparing 7 collapsed level.
8 We show that by comparing collapsed level 9 show that -- showing that collapsed level we miss our 10 protective wheel, minimum critical heat flux ratio 11 remain several or -- stage (phonetic) of our limit and 12 we also had additional acceptance criteria that 13 continued pressure and temperature in its developed 14 design limit because containment is an integral part 15 of ECCS, although I want to mention that there is a 16 separate methodology to calculate peak containment 17 pressure and temperature.
18 CO-CHAIR MARCH-LEUBA: Yeah. I asked you 19 cover, and because we got all these slides yesterday 20 I read it overnight. I asked you to cover the 21 difference between the different CHF correlations that 22 we are using all over different methodologies.
23 You're referring here to the CHFR limit or 24 the RELAP LOCA analysis.
25 DR. SAWANT: LOCA analysis.
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67 1 CO-CHAIR MARCH-LEUBA: Yes. Will we --
2 will you guys be covering that or is it the staff that 3 is going to be covering the difference in CHF?
4 DR. SAWANT: Part of LOCA closed session 5 presentation we will talk about the correlation that 6 is in the LOCA methodology.
7 CO-CHAIR MARCH-LEUBA: So let me to do it 8 myself, okay, because if anybody is reading the record 9 and goes back in time to July or June of 2019, we left 10 this rule with me saying that there is no way that 11 these CHF correlations are possibly correct because 12 there is a significant difference between the ones 13 predicted by the non-LOCA, ones predicted by LOCA and 14 then finally ones predicted by rolodiction (phonetic) 15 are also different, and when we went to Corvallis we 16 had a presentation and we -- told that indeed they are 17 different, and for LOCA there is not a CHF correlation 18 but there's a critical power correlation, and that's 19 a difference in numbers and that's why they gave a 20 factor of two difference between one and the other.
21 So I just wanted to put in the record that 22 as far as I am concerned I am satisfied on whatever we 23 left on back in June has been -- has been resolved and 24 needs to be on the record that we don't have that 25 discrepancy any longer.
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68 1 DR. SAWANT: Understood.
2 CO-CHAIR MARCH-LEUBA: So you are talking 3 about the CHFR -- that you are -- the CHF correlation 4 disagrees with the power correlation and using the 5 LOCA specific for the NuScale fuel and applicable to 6 the RELAP average bundle conditions.
7 DR. SAWANT: Yes, for LOCA -- and this 8 slide shows a roadmap of the methodology. Again -- we 9 do satisfy conservative requirements of Appendix K and 10 these elements are basically from Reg. Guide 1.203 11 with Element 1 that talks about establishing the 12 requirements for the evaluation model which is 13 generally done by developing for it.
14 So we present are working LOCA LTR Chapter 15 4. We will talk about the phenomena in a little more 16 detail in closed session.
17 Element 2 is, basically, a development of 18 assessment base that involves development of your 19 separated and integrated case assessment in this and 20 it's been documented in LTR Chapter 7, and Element 3 21 is a development of actual mathematical model.
22 We talk about that in several chapters in 23 LTR we would like to visit this mathematical model in 24 more detail in closed session.
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69 1 evaluation of evaluation model, which is there we go 2 through top-down and bottom-up evaluation to assure 3 the applicability of the evaluation model to predict 4 these high-ranked phenomena.
5 So, again, like in open session I recall 6 the overview but there is some more details about 7 scaling in distortion analysis in closed session.
8 DR. SCHULTZ: Pravin, you draw this as if 9 it's direct from top to bottom, that is, in terms of 10 information flow. Was it not iterative? In other 11 words, you did the PIRT but I presume you came back to 12 that as you did Elements 2, 3 and 4.
13 DR. SAWANT: Yes -- like, we were talking 14 in closed session. Like, there is some phenomena that 15 we had a different understanding when we started 16 developing methodology and when you finish the 17 analysis. So we will talk about those kind of things.
18 DR. SCHULTZ: Thank you.
19 DR. SAWANT: This slide just gives very 20 intense safety systems. I think almost everybody 21 probably is aware of the -- how safety systems --
22 NuScale safety systems work.
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70 1 part of the RPV on lower relation and when ECCS 2 activates all these valves open and the primary 3 function of these ECCS is to depressurize the RPV so 4 that we can inject the coolant that is collected in 5 the containment back into the RPV and develop it.
6 And the actuation signal for ECCS is a 7 high containment level. It also gets actuated on loss 8 of AC power. But there is a timer, like, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
9 And there is a feature called inadvertent actuation 10 block which is there to prevent the inadvertent 11 opening of any of these releases as well when the 12 reactor pressure is very high, and these blocks they 13 release the ECCS once the differential pressure 14 between RPV and CNV goes below some certain set point.
15 And model prediction system function is, 16 like, basically, to make sure that reactor scrams and 17 steam generator containments are isolated and safety 18 systems are activated. This can be ECCS and DHRS.
19 Decay heat removal system consists of 20 blueprints each designed to remove 100 percent decay 21 heat and just wanted to note that we are not crediting 22 this additional system in our LOCA analysis -- LOCA 23 evaluation model and we will show that that -- that 24 that approach is conservative for LOCA analysis.
25 CO-CHAIR MARCH-LEUBA: You go back --
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71 1 probably the question is for Ben, or maybe you know.
2 I was asked before earlier and I didn't remember.
3 What action of the model protection system triggers 4 the DHRS passive cooling -- turning it on?
5 I mean, the -- I know loss of power will 6 do -- will be one. But if not loss of power, what 7 triggers it? Is it high containment level or --
8 MR. BRISTOL: So the containment will 9 isolate on high containment pressure after the break 10 occurs and then DHR.
11 CO-CHAIR MARCH-LEUBA: High pressure?
12 MR. BRISTOL: High containment --
13 CO-CHAIR MARCH-LEUBA: Oh, high 14 containment pressure.
15 MR. BRISTOL: Yeah, high containment 16 pressure so which is still well below atmospheric 17 conditions. So we reached that actuation very quickly 18 under --
19 (Simultaneous speaking.)
20 CO-CHAIR MARCH-LEUBA: Under the DHRS 21 automatic critical -- I mean, DHR is activated by high 22 pressure containment?
23 MR. BRISTOL: So the secondary isolates on 24 high containment pressure and then the DHR will 25 actuate on high secondary pressure after the isolation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 1 of the secondary.
2 CO-CHAIR MARCH-LEUBA: I knew that. I 3 just -- somebody asked me and I couldn't remember.
4 It's convoluted but we will include this.
5 DR. SAWANT: So Element 1 of the Reg.
6 Guide 1.203 which is establishing the requirements for 7 the evaluation model as a first step you basically 8 develop the PIRT and PIRT processes are quite well 9 established.
10 Basically, we divide the NPM into 11 different systems and complements so that we can 12 identify high-ranked phenomena in each phase of the 13 LOCA progression and this PIRT development was done by 14 organizing the panel of experts -- known experts in 15 thermal -- area.
16 DR. CORRADINI: Can you remind us who 17 those are?
18 DR. SAWANT: Analysis presentation we will 19 a list of those experts.
20 DR. CORRADINI: Oh, okay. Fine.
21 DR. SAWANT: So I will defer that.
22 DR. CORRADINI: Okay. Thank you.
23 DR. SAWANT: There is a long list so --
24 MEMBER BLEY: Page 42.
25 DR. SAWANT: Yeah.
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73 1 MEMBER BLEY: Page 42 of the topical gives 2 a list of 16 of them. Yeah.
3 DR. SAWANT: So because summaries with our 4 acceptance criteria for the LOCA methodology and we 5 identified the phenomena and ranked them for their 6 importance based on their impact on the PR submerator 7 (phonetic) into three different importance levels and 8 we also identified knowledge level, like, based on the 9 -- how well they are understood and what kind of 10 assessments they have at the level.
11 So this next slide basically shows the --
12 how we break down our system into different components 13 for development of PIRT and then also a typical 14 progression of LOCA in this design.
15 This is -- break progression where we show 16 the depressurization of RPV and pressurized from the 17 CNV Basically, on the left side you can see that RPV 18 is divided into low plenum core, a riser for plenum 19 and a downcomer and pressurizer to identify the 20 phenomena. And on the right side you can see that 21 Phase 1 is basically the blowdown phase where RPV 22 depressurizes and CNV pressurizes.
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74 1 of CNV.
2 That is the initiation of Phase 1(b) which 3 ends when there is a recirculation established from 4 containment to -- back to the RPV because initially 5 when ECCS opens there is a flow from RPV to CNV 6 through all the walls, like RVV -- RRVs, but once 7 pressure is equalized and there is enough high-risk 8 straticate (phonetic) in the containment flow in the 9 ECCS and that's the end of our Phase 1(b).
10 So we considered, like, a Phase 1(a) and 11 1(b) as a part of LOCA analysis and Phase 2, which is 12 the long-term cooling analysis, which is -- covers the 13 long-term cooling analysis methodology.
14 Element 2 of the Reg. Guide 1.203 is like 15 establishing the assessment base. As I said earlier, 16 this involves, like, developing separate and 17 integrative assessments. But before we talk about the 18 assessment base, I'd just like to review some 19 background on our mathematical model that we have 20 selected.
21 Our thermal -- system thermal hydraulic 22 code that we have selected for LOCA analysis is 23 NRELAP5. It's developed from the RELAP5-3D as a 24 baseline for it and I have some sub-bullets there that 25 gives basic information of RELAP5-3D, which is I think NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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75 1 well known thermal-hydraulic code.
2 We have established a code configuration 3 control as it relates all the development to convert 4 RELAP5-3D. The NRELAP was done under NuScale's NQA-1 5 program and these -- we'll talk about some of these 6 modifications in closed session.
7 But these modifications that we did in 8 RELAP-3D are mainly related to model NuScale-specific 9 components and some modifications so that we can 10 implement Appendix K models and corrections that were 11 reported as part of RELAP5-3D program.
12 So the next slide shows just overview of 13 our assessment base -- basically, database. We have 14 -- as we will see in closed session we have extensive 15 coverage of all of our high-ranked phenomena.
16 There are six NuScale-specific integral 17 effect tests that were performed in NIST facility.
18 There are two separate effect tests that also were 19 performed in a NIST facility. Four of the separate 20 effect tests were -- these are also NuScale-specific.
21 Facilities includes helical coil simulator 22 test facilities, SIET, and CHF test facilities, KATHY 23 and Stern. So there is an extensive set of NuScale-24 specific data for evaluation of evaluation models.
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76 1 legacy separate effect test data. These are the datas 2 that we are offering from the existing literature 3 because we found that they are applicable for our 4 design. So our database includes those nine tests 5 also.
6 Next slide. Since this NIST facility is 7 our -- one of the primary source of separate effect as 8 relating to -- effect data there's just a little bit 9 more datas of the design features of the facility.
10 This facility has all the important 11 components of the NPM are reprinted here. RPV has 12 basically a lower plenum core riser, downcomer 13 pressurizer, represented as a integral part of 14 reactive pressure vessel.
15 RPV is located outside of the containment 16 for scaling reasons and it's connected to the 17 containment by ECCS walls and break -- by ECCS lines 18 and break lines.
19 Containment -- that you see that new 20 vessel is connected to the reactor pool, which is 21 represented by green vessel there by a rectangular 22 heat transfer plate, which it scales the NPM active 23 containment heat transfer area, and there is this 24 shell -- for the containment that is there to contain 25 the pressure. So I just want to highlight that that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 heat transfer plate is the scale of the area that 2 connects the containment to the pool.
3 And, yeah, the scaling -- a little bit 4 scaling information, scale it's a power rolling 5 (phonetic) scaling used for designing the facility.
6 It's various height, various volume scale. Pressure 7 temperatures are one-to-one scale and time scale is 8 also the same, one to one.
9 CO-CHAIR KIRCHNER: Could you, with the 10 mouse, point, Matthew, perhaps to where the 11 containment is on that -- on that figure. On the 12 right. Yeah, it's there. But now go over to your 13 diagram. Yeah, there.
14 DR. SAWANT: Okay. This is -- yeah.
15 CO-CHAIR KIRCHNER: So that's the 16 containment there. So is the gap in the containment 17 scaled or is that prototypical in terms of -- where I 18 am getting at is the distance between the reactor 19 vessel and the containment wall. Is that prototypical 20 or scaled?
21 DR. SAWANT: The volume is scaled.
22 CO-CHAIR KIRCHNER: No, I understand the 23 volume is scaled. But is the gap -- the width, is 24 that prototypical or is it scaled as well?
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78 1 heat transfer plate that connects the --
2 CO-CHAIR KIRCHNER: No. No.
3 DR. CORRADINI: He's asking -- he's asking 4 -- look at the blue. He's asking you've got a 5 semicircle -- a semi-hemisphere or a semi-cylinder 6 with the blue. He's asking the width of the blue, is 7 it scaled or not. The blue which is the containment 8 -- is the width of it scaled? The top -- the height 9 I understand is --
10 DR. SAWANT: Yeah. Width of the heat 11 transfer plate that connects the --
12 DR. CORRADINI: Not the plate.
13 CO-CHAIR KIRCHNER: Not the plate, but the 14 actual --
15 DR. CORRADINI: Volume.
16 CO-CHAIR KIRCHNER: The annulus. Yeah, 17 what simulates the annulus. What I am getting at is 18 that you would have a turbulent mixing as you have a 19 blowdown.
20 DR. SAWANT: Okay.
21 CO-CHAIR KIRCHNER: But that mixing, in 22 effect, on the condensation and heat transfer is going 23 to be to a certain extent impacted by how wide or that 24 annulus -- you know, in the real -- in your actual 25 NuScale power module that gap is like -- like this and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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79 1 something -- I am just guesstimating. But how did you 2 scale that? Is that kind of the same width or --
3 DR. KURAN: Well, this is Selim Kuran from 4 NuScale.
5 The issue is that maintaining the required 6 volume scale along with the heat transfer area as well 7 as internal length scale, which is where we do 8 hydrotherameter (phonetic) is very difficult issue at 9 the same time.
10 CO-CHAIR KIRCHNER: Yes.
11 DR. KURAN: So you prioritize what you are 12 scaling for. Obviously, volume comes first, and the 13 same importance heat transfer area. So we are not 14 scaling the internal length scale that would define 15 multi-dimensional phenomena in this facility.
16 CO-CHAIR KIRCHNER: So it was a leading 17 question because I am thinking of the impact on the 18 condensation heat transfer based on turbulence during 19 the blowdown and subsequent phases in that gap, and 20 whether that is characteristic in terms of the 21 phenomena with the actual NPM. You see where I am 22 going with this?
23 DR. SAWANT: Yes, I think we can address 24 that in closed session.
25 CO-CHAIR KIRCHNER: Okay.
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80 1 DR. KURAN: Well, given the curvature of 2 the surface in a prototypical geometry and how that is 3 similar to, you know, flat plate, I think for that 4 film condensation on a flat surface we are not off.
5 CO-CHAIR KIRCHNER: Yeah. No, I 6 understand that, you know, because the diameter of the 7 actual module containment vessel is so large you can 8 effectively treat it as a flat plate, in effect.
9 But the gap becomes a factor in how the 10 turbulence, how the phenomena, plays out during the 11 blowdown and that would affect, you know, the 12 condensation rate and that would affect how the pool 13 that accumulates is heated or sub-cooled. You see 14 where -- that is what I am asking.
15 DR. SAWANT: Yes, and then this 16 distributed distribution of mixture in containment.
17 While it is not high-ranked phenomena, we will talk 18 about that in closed session.
19 CO-CHAIR KIRCHNER: In the closed session.
20 Okay.
21 MR. BRISTOL: And in the assessments in 22 the closed session you'll see that the containment 23 level increases characteristic within time to the NPM.
24 CO-CHAIR KIRCHNER: So they are -- that's 25 the volume part is --
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81 1 MR. BRISTOL: That's right.
2 CO-CHAIR KIRCHNER: -- is consistent.
3 Yeah.
4 MR. BRISTOL: Right. So the accumulation 5 and the removal of that surface area is 6 characteristically scaled.
7 CO-CHAIR KIRCHNER: Okay.
8 MEMBER BLEY: I have a much simpler 9 question. When you did the PIRT you then went after 10 the things that were of high importance and limited 11 knowledge, I guess. You had four knowledge scales --
12 1, 2, 3, 4, with 4 being we know this real well and 13 you had a lot of 2 and 3 level knowledge, which is 14 partial uncertainty, and no 1's that I saw, after 15 you've done your experiments you have all of those up 16 to 4's? Did you evaluate where you stand after you 17 did all the experiments?
18 DR. SAWANT: So we -- so based on our 19 assessment base and assessment of correlations and 20 models, we can finally conclude that we have enough 21 information to say that our evaluation model is 22 applicable to model those phenomena 23 So yeah, we had final conclusions showing 24 that's applicable.
25 MEMBER BLEY: You're well-bounded now.
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82 1 Your knowledge is good in all of these areas that were 2 --
3 DR. SAWANT: Yes.
4 MEMBER BLEY: -- questionable before.
5 Okay.
6 DR. CORRADINI: So can I follow on 7 Dennis's question?
8 You had three PIRT panels over six years.
9 DR. SAWANT: Yes.
10 DR. CORRADINI: A couple of them were the 11 same.
12 DR. SAWANT: They were --
13 DR. CORRADINI: Of people. So did 14 something drastically change after the three times 15 that you did it? Or, to put it differently, why did 16 you do it three times?
17 DR. SAWANT: So there were evaluation of 18 the design --
19 DR. CORRADINI: So these are different 20 designs they were looking at?
21 DR. SAWANT: Some minor changes to the 22 design like how ECCS is functioning or it was changed 23 from 2008 to 2015 and we basically captured those 24 design evaluations.
25 DR. CORRADINI: Okay. By these --
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83 1 DR. SAWANT: Yes, on the PIRT panels. We 2 heard three PIRT panels over this period of 2008 to 3 2015 and each time we went through the same process to 4 make sure that PIRT is consistent with the design.
5 DR. CORRADINI: But these experiments were 6 done after the 2015 PIRT panel.
7 DR. SAWANT: Yes. Mm-hmm.
8 DR. CORRADINI: So the relevant one, if I 9 were to look at one versus the other, would be the 10 2015 because the design was more stable by that time 11 and the knowledge gaps -- the places where you needed 12 to do something were based off of that.
13 DR. SAWANT: Yes. All the experiments 14 that we -- we are going to talk about NIST --
15 NuScale's experiments that were conducted after 2005.
16 DR. CORRADINI: Okay. Fine.
17 DR. SAWANT: Except, like Stern CHF tests, 18 they were performed earlier. But we will talk about 19 those tests.
20 DR. CORRADINI: That's fine. I just want 21 to make sure I understood the logic.
22 MEMBER BALLINGER: So the PIRT ranking 23 table that Dennis has talked about you got heavy duty 24 instrumentation in your experiment. After the 25 experiments were there any surprises? Because NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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84 1 surprise sometimes is actually more beneficial than 2 confirmation bias, if you want to call it.
3 DR. SAWANT: Mm-hmm. Not huge surprises.
4 But there were some lessons learned that we 5 implemented eventually in the evolution model 6 development.
7 MEMBER BALLINGER: You're going to tell us 8 about those?
9 DR. SAWANT: Yes. Yes.
10 CO-CHAIR MARCH-LEUBA: While we are 11 wasting time, can you go back to slide seven?
12 MEMBER BALLINGER: While we are what?
13 CO-CHAIR MARCH-LEUBA: Wasting time. We 14 are behind schedule. But we have all of tomorrow 15 morning --
16 MEMBER BLEY: Well, you're not focused on 17 angels on a pin.
18 CO-CHAIR MARCH-LEUBA: Okay. I am looking 19 at the PIRT and I don't see any -- that the 16 experts 20 identified volume distribution in the vessel and 21 containment as one of the important characteristics of 22 this. Is that correct?
23 DR. SAWANT: The PIRT was specific for 24 LOCA, thermal hydraulic --
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85 1 I asked you to go to level -- to seven you can see 2 here. If you see under this condition where the water 3 levels are, the core is boiling off concentrating 4 boron inside.
5 The stimulator is condensing the still 6 water and the still -- the steam is going into 7 containment. So you have a high concentration of 8 boron in the red region and a very low concentration 9 of boron in the downcomer.
10 Eventually, you are going to flood the 11 containment to recover your levels so you can refuel 12 and you are going to restart the circulation and flush 13 that cold unborated water from the downcomer. Was 14 that identified in the -- in your PIRT?
15 DR. SAWANT: Not in the LOCA PIRT. It's 16 part of --
17 CO-CHAIR MARCH-LEUBA: This is my favorite 18 topic, right.
19 DR. SAWANT: It's part of long-term 20 cooling analysis and I think --
21 CO-CHAIR MARCH-LEUBA: It was not 22 identified in the long-term cooling analysis here 23 because I have that data -- important because I asked 24 for it. It was not. So somebody is going to address 25 my favorite topic, right. Not here, not now, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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86 1 eventually.
2 MR. SCHMIDT: This is Jeff Schmidt from 3 Reactor Systems. Yes, that'll be in the Chapter 15.
4 CO-CHAIR MARCH-LEUBA: I want to say in 5 the -- I am going to suggest that the PIRT was 6 slightly defective in the sense that it did not 7 identify a very significant physical mechanism that's 8 happening here.
9 And the problem I see is that both NuScale 10 engineers and the staff reviewers we are putting our 11 sight shields from -- like a donkey and looking 12 specifically at very specific problems and are not 13 seeing the whole thing.
14 And we are taking advantage of things that 15 happened in operating reactors and we have -- where we 16 have a lot of experience and saying that could not 17 possibly happen in my reactor.
18 Therefore, I don't have to analyze it 19 when, obviously, this reactor has particular behavior 20 that does not occur in operating reactors like the rod 21 ejection could create a missile that actively heats 22 containment.
23 And I've been reviewing the references 24 that the staff gave us and Section 351.1 says, no, it 25 can't happen, and the response to RAI 92 -- 9647, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 which Chris has been nice enough to give me, says it 2 cannot happen so we are not going to analyze it.
3 Okay. I see a problem in the -- both the 4 design and the review of another reactor in which we 5 take all the advantages of the existing reactors, 6 because if they don't do it we don't have to do it, 7 and then all these advantages we say that doesn't 8 apply to me either.
9 So I am complaining about it and I see two 10 examples that came out this morning, and inside the 11 vessel missiles that can be generated in a heated 12 containment, which I agree with Mike, it is not going 13 to do anything but you didn't analyze it. You said 14 it's not possible.
15 I see the boron dilution event, the PIRT.
16 They didn't even identify it. So I see deficiencies.
17 Okay.
18 My five cents.
19 CO-CHAIR KIRCHNER: And just as a 20 reminder, we are in a subcommittee meeting and you're 21 hearing individual opinions, and we will speak to you 22 through a letter report after a full committee 23 meeting. So just keep these in context. Thank you.
24 Thank you, Jose.
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88 1 bit of context. There's several things that we have 2 considered and licensing strategy comes into play when 3 we are considering how we want to submit a particular 4 safety evaluation. The dynamic effects of CRDM 5 failure was considered within NuScale and it was 6 determined the most appropriate strategy for 7 addressing those dynamic effects was the strategy that 8 you see in the topical or in the FSAR itself and, I 9 mean, the -- I think a very simple evaluation of the 10 momentum versus the pressure and the length to the top 11 of the CRDM or to the top of containment would lead us 12 to think that that's probably a fairly benign 13 transience.
14 CO-CHAIR MARCH-LEUBA: It's not that I 15 agree with you but a back-of-the-envelope calculation 16 that is placed in the record or at least kept as an 17 internal document in Corvallis would help your 18 argument that yes, we did consider it. We looked at 19 it and the maximum velocity was only so many meters 20 per second that cannot do anything.
21 MR. BRISTOL: Sure. And, again, not all 22 of that analysis that we have considered throughout 23 ends up as part of our licensing submittal basis.
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89 1 possibly happen.
2 MR. BRISTOL: That's correct.
3 MR. PRESSON: Well, at least, in terms of 4 the -7th. But yes.
5 MEMBER REMPE: Your response says a simple 6 calculation could show. It would provide more 7 confidence if you'd said we did a simple calculation 8 and found. Is that -- which happened?
9 MR. BRISTOL: We had considered a variety 10 of different strategies and the strategy with all of 11 the rest of the considerations specific to the dynamic 12 effects of the CRDM failure was to go the route that 13 we have submitted as part of the FSAR.
14 MEMBER REMPE: Well, that's one thing what 15 you submit for your licensing strategy. But what I am 16 asking did you do a calculation -- has it been done?
17 MR. BRISTOL: We have -- we have 18 considered the surface areas in the DPs. I cannot say 19 that we have specifically documented the acceleration 20 effects and the -- and those dynamic effects because 21 of the analysis strategy that we selected.
22 CO-CHAIR MARCH-LEUBA: So that sounds to 23 me like no, we did not do our calculations. Let me 24 --
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90 1 long-term, the long-term cooling does consider the 2 effects of boron transport. We identified within 3 NuScale the limitation there was the accumulation of 4 boron.
5 Now, the recovery actions we will discuss 6 in two weeks.
7 CO-CHAIR MARCH-LEUBA: I cannot wait to 8 see what the recovery actions are going to say because 9 I don't see how you can recover.
10 MR. BRISTOL: Sure.
11 CO-CHAIR MARCH-LEUBA: And we will talk 12 about it, but you're going to hear -- next month 13 you're going to hear a lot about a two-space bin 14 getting out of the tube. Once you separate the boron 15 you cannot put it back in the tube. Okay. Let's keep 16 going.
17 DR. SAWANT: So Element 3 of Reg. Guide 18 1.203 builds the development of a mathematical model.
19 So, again, in closed session we will talk about the 20 analysis that we did and justifications of how they 21 selected nodalization, time slips, also biases that we 22 are implementing or initial and boundary conditions 23 and treatment of set points and trips.
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91 1 presents this local break spectrum calculations where 2 we look at the -- several break sizes and locations, 3 probability and single failures, and we will talk 4 about those results in closed session.
5 And we also did several sensitivity 6 calculations that we are required by Appendix K but 7 also the sensitivity calculations that -- phenomena-8 specific to guide us on particular aspects of the 9 methodology, and also to establish our conservative 10 biases. So we will, again, like, introduce those 11 calculations in closed session.
12 Then, finally, this Element 4, which is 13 the applicability evaluation where we assess the 14 accuracy of evaluation model with the bottom-up and 15 top-down analysis. So, basically, in bottom-up 16 analysis we look at every high-ranked phenomena and 17 models and correlations needed to simulate those high-18 ranked phenomena.
19 So we identify important parameters for 20 those models and correlations and what is the required 21 range for those parameters in NPM local application so 22 that we can evaluate the models and correlations for 23 their applicability to all those range. So we look at 24 the validity of the models and correlations, their 25 assessments against external data that is reported as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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92 1 well as any specific assessment that we have 2 performed.
3 We also look at scalability of those 4 models and correlations to our NPM design. So all of 5 this information is summarized in LOCA LTR Table 8-1 6 and we will revisit some of these aspects in closed 7 sessions.
8 For the top-down evaluation, we basically 9 look at the field equations and numerics and its 10 applicability to model LOCA operation. But a major 11 part of top-down evaluation is to show the 12 applicability of NIST integral effect test data 13 through a simulator LOCA in NPM through scaling and 14 distortion analysis. So we will talk about that 15 aspect a little bit more later in closed session.
16 And, basically, we show that -- how the 17 data -- the IET data is applicable and why NRELAP is 18 still able to model the LOCA in the NPMs.
19 So, again, these are the conclusions. We 20 will expand on these conclusions in closed session or 21 we -- so the evaluation model has multiple layers of 22 conservatisms in the addition to Appendix K, so we 23 will talk about those conservatisms in closed session.
24 LOCA methodology that we are presenting we have 25 developed here is cycle independent. Again, like, we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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93 1 will talk about that in closed session. It's 2 supported by extensive assessment database and so 3 sensitive to calculations and follow Reg. Guide 1.203 4 guidance and what we have showed is that core doesn't 5 experience CHF and it remains covered and there is no 6 clad heat-up and continuing pressure, again, 7 preserving its design limit. So that ends, like, my 8 open statement.
9 CO-CHAIR KIRCHNER: Okay.
10 DR. SAWANT: But we helped, like, this --
11 opening up a reactor vent valve is also presented as 12 part of LOCA LTR Appendix B, and extension of LOCA 13 evaluation model to analyze that event. You will 14 recall that event. So --
15 MR. BRISTOL: Thanks, Pravin.
16 So, just quickly, the purpose of the 17 Appendix B is to describe the evaluation or the 18 extension of the LOCA evaluation model for what's 19 traditionally analyzed more like a non-LOCA event, 20 i.e., an inadvertent opening of a reactor valve. In 21 the NuScale design that is not an event that is re-22 isolated. Therefore, phenomenologically, it's more 23 consistent with the LOCA event progression.
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94 1 we do the evaluation. So as part of the appendix 2 there's a description as to how -- the events 3 including the applicability of the LOCA PIRT and 4 assessment basis, and then the applicability 5 evaluation that Pravin just covered. However, the 6 evaluation model is slightly different.
7 Next slide. So those differences are in 8 order to be a little bit more consistent with our AOO 9 evaluation techniques and specifically that's related 10 to the data set that was used to assess the CHF 11 correlation limit as well as the way that the power 12 profiles are applied. Otherwise, the evaluation model 13 nodalization containment modeling, et cetera, are 14 consistent with the LOCA EM.
15 Next slide. So in conclusion, the IORV 16 extension of the LOCA EM is justified because of the 17 similarity of the transient production and the 18 phenomena.
19 Again, there's some minor differences in 20 the way that CHF is evaluated that's more consistent 21 with an AOO event. The transient conclusions, the CHF 22 occurs very quickly. It's driven by flow and pressure 23 transient effects.
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95 1 and the collapse level remains well above the top of 2 active fuel for the event.
3 MEMBER PETTI: I have a question. Why is 4 it classified as an AOO? Is it based on probability?
5 Do you think that once in the light fuel plant you're 6 going to inadvertently open this valve?
7 MR. BRISTOL: We do not expect to 8 inadvertently open one of the valves in the life of 9 the plant. It was a conservative characterization and 10 that's consistent with what we evaluate as being 11 industry precedent for a 15661 type of event.
12 MEMBER PETTI: Okay. A 1561 --
13 MR. BRISTOL: So that's the inadvertent 14 opening of a reactor safety valve. In this case 15 because the reactor safety valves and reactor vent 16 valves are fairly similar, even though the ECCS 17 analyses, as we know, have a very low likelihood of 18 spuriously opening. It's a postulated event and we 19 characterize it similarly.
20 MEMBER PETTI: Thank you.
21 CO-CHAIR MARCH-LEUBA: We have heard --
22 let me use two minutes. We have heard that the IEV or 23 the RVVs or RRVs especially have some tolerances on 24 their block/no block activation pressures and, 25 therefore, it is very important to use the proper NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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96 1 staggering of opening of valves in the analysis.
2 Is that -- it becomes part of your 3 methodology that we are approving today in this report 4 or is that something that you consider once?
5 MR. BRISTOL: That would be the 6 implementation of the SR. So the implementations 7 calculations as part of the EM development we evaluate 8 sensitivities to a range.
9 But the actual specifics of the range and 10 the change that was made to the range is part of the 11 Chapter 15 analysis and not part of the top core.
12 CO-CHAIR MARCH-LEUBA: So that would a 13 question for the staff, which I'll ask you later on is 14 if we found that the staggering -- the time of opening 15 of the valves made a significant difference why wasn't 16 it a limitation or a condition that for implementation 17 of this EM the valve shall use the most conservative 18 valve opening times according to specifications?
19 I mean, if you can do it between -- let me 20 use some number parameters -- 500 and 700 psi. You 21 should use either 500 or 700 on these -- and in the 22 proper order.
23 MR. BRISTOL: So the methodology evaluates 24 the IAV and identifies it as being important for 25 particularly the minimum level scenarios, and then NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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97 1 specifies that as a sensitivity that's performed on a 2 regular basis as part of the Chapter 15 event 3 analysis.
4 CO-CHAIR MARCH-LEUBA: If we already know 5 that opening -- if you have an IORV of an RRV, opening 6 the second IORV first and then the RVV is worse. So 7 the methodology should specify that that's what you 8 need to do.
9 MR. BRISTOL: So that's worse in respect 10 to peak containment pressure as we identified. So 11 that would be covered as part of the technical report 12 for containment pressure analysis.
13 CO-CHAIR MARCH-LEUBA: And for the next 14 six months if you run the analysis you are going to do 15 it that way, correct?
16 MR. BRISTOL: For the containment peak 17 pressure analysis, yes. Yeah.
18 CO-CHAIR MARCH-LEUBA: But we are 19 approving this topical report for the next 60 years.
20 When you guys are working somewhere else and the new 21 engineer comes in, if that is not specified in the 22 methodology he may forget.
23 So, I mean, this is something that -- it 24 should be a condition. If we know that it affects the 25 results, it should be a condition in the SER. It NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 1 should have been in your topical report as your 2 methodology and -- okay, I am just putting it in here 3 that there might be a discussion of it as a condition 4 to your SER.
5 CO-CHAIR KIRCHNER: I would disagree with 6 you, Jose. If the methodology is robust enough to 7 look at different valve sequences and their opening, 8 then the limitation comes based on the containment 9 pressure analysis or other places like Chapter 15, not 10 in the methodology report.
11 CO-CHAIR MARCH-LEUBA: You know, it comes 12 on the methodology. These type of limitations come 13 always in the topical report.
14 It tells you whenever you use this 15 methodology this is the order of opening valves that 16 you have to use. You know, otherwise, if you -- if 17 you use a different opening order you get no 18 conservative results.
19 CO-CHAIR KIRCHNER: Well, you look at the 20 various sequences of valves opening and you then come 21 up with your peak containment pressure.
22 CO-CHAIR MARCH-LEUBA: Or the -- or the 23 methodology. Look at the spectrum of opening times, 24 which would be the best thing to do. But it has to be 25 in the methodology. Otherwise, I want to run one.
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99 1 And my concern -- I mean, these guys are going to do 2 it right.
3 My concern is 60 years from now when the 4 new guy comes in what is he going to do and that's why 5 we do approved methods so you follow them and nobody 6 has to question.
7 Yes. Okay. That will be for the staff.
8 DR. SCHULTZ: And what is the technical 9 basis for removing the 15 percent stored energy bias 10 that -- in this particular evaluation?
11 MR. BRISTOL: So we apply more of a cycle-12 specific fuel condition and so the LOCA methodology 13 was developed considering a broad array of potential 14 stored energy conditions based off of the FSAR core 15 design.
16 DR. SCHULTZ: And the methodology is to be 17 cycle specific in this particular evaluation?
18 MR. BRISTOL: And the IORV would be more 19 cycle-specific. Correct.
20 DR. SCHULTZ: Thank you.
21 CO-CHAIR KIRCHNER: Any more questions 22 from the members? Okay. I am going to propose that 23 we take a break and return at quarter to 11:00 on the 24 clock on the wall. And so we are recessed.
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100 1 off the record at 11:28 a.m. and resumed at 11:43 2 a.m.)
3 CO-CHAIR KIRCHNER: We're back in session.
4 We're going to turn back to NuScale staff 5 again. And now we're going on to the non-Loss of 6 Coolant Accident methodology.
7 So, please, Matthew.
8 MR. PRESSON: Thank you.
9 CO-CHAIR KIRCHNER: And introduce your 10 colleagues.
11 MR. PRESSON: Yes. Again, I'm Matthew 12 Presson, Licensing Project Manager for NuScale. We'll 13 be discussing the non-Loss of Coolant Accident in this 14 presentation.
15 And presenting this will be Ben Bristol, 16 Supervisor of System Thermal Hydraulics; Megan 17 McCloskey, our thermal hydraulics analyst; myself, and 18 Paul Infanger will be available if there are any 19 specific questions that come up, so.
20 I will pass it over to you all.
21 MS. McCLOSKEY: Thanks, Matthew.
22 Today in the open, in the open session 23 we'll cover the scope of the topical report, starting 24 with the non-LOCA events and the acceptance criteria 25 and which of the acceptance criteria that are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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101 1 addressed as part of the non-LOCA system analysis in 2 the topical report, as well as those that are 3 addressed by other methodologies that are part of 4 separate review and approval processes. The interface 5 with those methodologies is part of the non-LOCA 6 topical report.
7 I have one slide on the factors of the 8 NuScale power module design that really control the 9 margins to the acceptance criteria for the non-LOCA 10 event.
11 And then we'll move into an overview of 12 the development of the non-LOCA EM starting with the 13 PIRT and the GAP analysis done to identify phenomena 14 that were specific to or more important to the non-15 LOCA system analysis compared to the LOCA EM and the 16 RELAP, NRELAP5 assessment plan for non-LOCA. And 17 then, finally, review the general event analysis 18 methodology and the scope of the event-specific 19 methodology that's identified in the topical report.
20 So first, what in-scope versus out-of-21 scope of this topical report were focused on the 22 system transient analysis of non-LOCA events using 23 NRELAP5. And the EM is applicable for the MTM plant 24 design during the short-term transient progressing --
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102 1 compression vessel remains above the top of the riser 2 and natural circulation is maintained.
3 We describe the interface between the 4 system transient analysis and the downstream sub-5 channel and accident radiological analysis, but the 6 details of those EMs are part of separate topical 7 reports.
8 And as we had already started to discuss 9 today, we have NuScale topical report for control rod 10 ejection and LOCA and valve opening events. So, those 11 are not here.
12 We discussed the containment pressure 13 temperature analysis technical report in as part of 14 the Chapter 15 review with this subcommittee back in 15 June of last year.
16 And, as I mentioned, the EM applies to the 17 short-term transient progression where the riser 18 remains covered. And that longer-term transient 19 progression with DHRS that may result in a return-to-20 power event or riser uncovering is excluded from the 21 scope.
22 In terms of the applicable initiating 23 events they are summarized here. The events are 24 classified as anticipated operational occurrences, 25 unless they're identified as infrequent events for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 1 postulated accidents. They're categorized into one of 2 five categories, depending on the impact to the 3 reactor module from the specific -- from the 4 initiating event. So, what we have here is a subset 5 of the Chapter 15 event that we have had some previous 6 discussions on.
7 There are two NuScale-specific events that 8 are covered here. One of those is a cool-down event, 9 the loss of containment vacuum or containment 10 flooding. And then one heat-up event is inadvertent 11 operation of the DHRS.
12 MEMBER MARCH-LEUBA: And this is one of 13 the good points of the standard review plan. I mean, 14 they approach Chapter 15 real well by dividing into 15 these five categories. Makes you think about it. And 16 it's pretty comprehensive.
17 Really it's one of the, I will call it one 18 of the few places where the SRPs will be helpful 19 because it tells you what to do.
20 But do you think of any specific picture?
21 Because here's what developed based on the experience 22 with operating reactors. We have 50 years' 23 experience, 60 years' experience with operating 24 reactors and we know what can happen to them. Do you 25 consider anything that is not an operating reactor?
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104 1 NuScale has special control rods or special valves or 2 special --
3 MS. McCLOSKEY: When we went through this, 4 the identification of design basis events for the 5 NuScale power module design we considered the systems 6 that are specific, we considered the NuScale system 7 and components and what effects those failures could 8 have on the modules. And what we found through that 9 systematic review process was that our scope of 10 initiating events look a lot like a typical PWR.
11 And then we have event progressions that 12 maybe that are somewhat un -- that are unique to the 13 NuScale design.
14 The long-term cooling with ECCS event 15 progression is a little bit unique because we, with 16 the boiling condenser condensing design of the ECCS in 17 containment will tend to redistribute flowing into the 18 core. And we don't have a mechanism as part of the 19 safety system design to require that operators take 20 action to prevent accumulation, further accumulation 21 of water in the core, while we treat it as part of the 22 recovery processes but we don't have a hot rod 23 switchover time.
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105 1 what we found there was that with respect to events 2 that have -- with respect to margin to acceptance 3 criteria our limiting conditions are in the short-term 4 part of the event progression. And so that's part of 5 why the focus of this topical report is on the short-6 term event progression.
7 MEMBER MARCH-LEUBA: Okay, thanks.
8 DR. CORRADINI: So, but I guess Jose's 9 question I thought you were going to answer it by 10 saying the two italicized things are the two things 11 that make this unique, everything else is pretty much 12 -- So, are you coming back to those specifically and 13 we're going to talk about those later? Or should I 14 ask now?
15 MS. McCLOSKEY: Go ahead and ask now 16 because I'm --
17 DR. CORRADINI: Well, I'm kind of curious 18 about how the loss of containment vacuum and/or 19 containment flooding would be initiated.
20 MS. McCLOSKEY: Well, loss of containment 21 vacuum since during normal operation the containment 22 operates in sub-atmospheric conditions --
23 DR. CORRADINI: Right.
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106 1 evacuation system. But maintaining that sub-2 atmospheric condition during normal operation, 3 containment flooding is postulated.
4 DR. CORRADINI: But they're separate, 5 they're separate events. It's a common system.
6 But what I was trying to get at is I 7 wouldn't have one necessarily concurrent with the 8 other.
9 MS. McCLOSKEY: Correct.
10 DR. CORRADINI: Okay.
11 MS. McCLOSKEY: Correct.
12 DR. CORRADINI: All right.
13 MS. McCLOSKEY: We kind of lumped them 14 into the same event analysis. But the causes are 15 different.
16 DR. CORRADINI: Okay.
17 MS. McCLOSKEY: And there's specific 18 calculations that are addressed are different RELAPS.
19 DR. CORRADINI: Okay. Right. And the 20 inadvertent operation of DHRS is just early, early 21 switchover to DHRS?
22 MS. McCLOSKEY: There is a couple, there's 23 different permutations of that event. It may be 24 inadvertent opening of one of the DHRS actuation 25 valves. It may be inadvertent actuation of one train NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 or two trains.
2 And we've also considered inadvertent 3 secondary side isolation of one train or two trains as 4 part of that event.
5 DR. CORRADINI: Okay, thank you.
6 MS. McCLOSKEY: Then this slide shows the 7 non-LOCA event acceptance criteria. And these are 8 really by just the range of acceptance criteria that 9 are defined in the SRP and in the NuScale DSRS for 10 AOOs and postulated accidents. We've also included 11 the infrequent events acceptance criteria, which is 12 consistent with the radiological acceptance criteria 13 that's discussed in the SRP.
14 The criteria in bold are those, the 15 maximum RCS pressure and the maximum steam generator 16 pressure that are addressed by system analysis.
17 And we also consider, we also evaluate the 18 escalation criteria. We demonstrate that an AOO 19 doesn't postulate an accident or constitute a loss of 20 functionality.
21 DR. CORRADINI: So, so I know that you 22 guys have told us this, but I forget. Infrequent 23 event versus accident, I'm looking at all the 24 criteria. They're essentially the same. So, what am 25 I missing as to why the differentiation?
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108 1 MS. McCLOSKEY: Well, radiological dose 2 criteria are different --
3 DR. CORRADINI: Ah, okay. Excuse me.
4 MS. McCLOSKEY: -- in a small fraction of 5 the accidents. That's critical.
6 DR. CORRADINI: Okay, thank you.
7 MS. McCLOSKEY: So this slide, again, 8 shows the general approach of the NuScale 9 methodologies to demonstrate that the range of 10 acceptance criteria for the non-LOCA analyses are met.
11 We take the plant design in combination with some 12 performance into the NRELAP5 model of the plant and 13 evaluate the pressure and primary and secondary 14 pressure limits, and demonstrate that we face 15 stabilized condition.
16 Boundary conditions are provided to the 17 downstream VIPRE analysis or for the radiological dose 18 analysis, which are, and the methodologies for those 19 are in separate topical reports.
20 And this slide gives a very high-level 21 overview of the characteristics of the design that 22 governed margin to acceptance criteria for the non-23 LOCA transient. For minimum critical heat flux ratio 24 there are key features during the design where we have 25 an overall low power density. And as a result of that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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109 1 general operation it's the combination of high 2 temperature, high power, high pressure conditions that 3 result in the limiting NCHFR conditions during 4 reactivity insertion events as compared to more rapid 5 power overshoot reactivity insertion events.
6 Our primary pressure is protected by the 7 reactor safety valve lift. There are two safety 8 valves on the reactor pressure vessel, and they're 9 non-LOCA analysis for the maximum lift pressure and 10 the minimum capacity of the first valve to maximize 11 the transient pressure calculation.
12 For secondary site pressure the design 13 pressure is equal to the primary site pressure. And 14 the secondary pressure keeps after DHRS actuation of 15 the boiling condenser mode coolant for the DHRS is 16 established. So, it's really limited by the primary 17 site temperature conditions.
18 And then in terms of radiological release, 19 the module protection system is the one to rapidly 20 automatically isolate the source of the release based 21 on the measured conditions.
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110 1 operation and support due to an event.
2 MEMBER MARCH-LEUBA: Some are complaining 3 it doesn't affect the non-LOCA methodology, but on 4 radiological releases we still want to hear more about 5 the procedure to de-isolate containment at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 6 make sure of the hot oxygen content. So, we are 7 expecting to hear more about that beginning in the 8 full committee next month.
9 And then the last one, establishing a 10 safe, stable condition, I keep hearing from people 11 that a return to power is not a safe, stable 12 condition. Maybe safe but certainly not stable.
13 So, sure, I mean, we're using here 14 nomenclature. And I'm complaining about semantics.
15 But that makes it different than other reactors to 16 have a weakened return to power. And that was just 17 justification for complaining.
18 MS. McCLOSKEY: The non-LOCA evaluation 19 model was developed to perform a conservative analysis 20 following the intent of Reg Guide 1.203, and applying 21 a graded approach to the evaluation model development 22 and assessment process.
23 As the first element of the EMDAP includes 24 establishing the applicable transience and the 25 acceptance criteria, which we've just discussed. And NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 1 then what the transience and acceptance criteria 2 identify we still develop the non-LOCA PIRT. This was 3 done towards the end of 2015, beginning of 20 --
4 sorry, end of 2014, beginning of 2015 timeframe, so 5 the design had progressed fairly far along in the 6 development. And we were able to take advantage of 7 that when evaluating the PIRT.
8 Then we did a GAP analysis that examined 9 the high-ranked phenomena identified in the non-LOCA 10 PIRT as compared to the LOCA PIRT and the assessment 11 plan for the NRELAP5 assessment plan as part of LOCA.
12 Considering that overlap we determined that a graded 13 approach to the EMDAP was reasonable, specifically in 14 that the LOCA topical report describes the NRELAP5 15 code and the validation basis for LOCA since the 16 premise of the graded approach was that that topical 17 report would be approved. And the non-LOCA topical 18 report leverages that code description and much of the 19 validation basis described without repeating that 20 information.
21 And so that the results of the GAP 22 analysis allowed us to focus on the additional NRELAP5 23 assessment and sensitivity calculations to focus on 24 non-LOCA phenomena events predominantly in the steam 25 generator and the DHRS validation.
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112 1 I'll have some additional about that in 2 the closed session tomorrow morning.
3 Finally, in the non-LOCA EM suitably 4 conservative emission boundary conditions are applied 5 for several parameters to bound the plant response.
6 And we'll talk as well about sensitivity 7 calculations to demonstrate the fact there's control 8 of margins to acceptance.
9 This slide summarizes the PIRT 10 developments, since the non-LOCA events are 11 categorized into different event types. We have 12 several of them to consider.
13 What we did for the PIRT development was 14 to fairly generically characterize the phases of the 15 non-LOCA transient progression.
16 Phase 1 is the pre-tripped transient where 17 the RCS was still near full power conditions or higher 18 power conditions, RCS flows are high. The DHRS hasn't 19 been actuated yet. The figures of merit during this 20 phase are critical heat flux ratio and RCS pressure.
21 Phase 2 is the post-trip transition from 22 reactor trip and DHRS actuation while the DHRS is 23 started but cooling is not fully effective. So, the 24 module is in a transitional state of flow and power 25 level is reactor trips.
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113 1 MEMBER PETTI: He's having trouble hearing 2 you I think, our transcriber. So, make sure you're --
3 MS. McCLOSKEY: Okay. I'll try to speak 4 a little more directly into the microphone here.
5 Thank you.
6 And then you'll see CHF, critical heat 7 flux ratio was a figure of merit throughout all of the 8 phases. RCS pressure and secondary pressure were also 9 figures of merit here as the DHRS gets actuated and 10 starts to pressurize the secondary system. For events 11 that results in breaks inside the containment on the 12 main steam line and the feedwater line. We also 13 considered containment pressure.
14 And then those three reflect stable, 15 natural circulation under with DHRS cooling fluid only 16 effective, and we're at low flow levels, and decay 17 heat power levels.
18 Under this condition, at least in this 19 phase the RCS mixture level was considered important 20 because with sufficient DHRS cooling and the right 21 initial mass inventory in the RCS, the DHRS coolant 22 can result in riser uncovery and interruption of 23 natural circulation. And at the end of Phase 3, 24 that's where we find the end of Phase 3 in the PIRT, 25 and the cooling could result in reactivity insertion, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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114 1 and therefore we also considered some sub-criticality 2 during this phase.
3 MEMBER BLEY: In your report I don't 4 recall that you gave the names of the people who 5 participated in this PIRT. Did you?
6 MS. McCLOSKEY: We did not include that in 7 the --
8 MEMBER BLEY: You did in the other report, 9 in the LOCA report.
10 MS. McCLOSKEY: -- in the non-LOCA topical 11 report, right.
12 MEMBER BLEY: Yes. Any reason? Same 13 people or was it always your people?
14 MS. McCLOSKEY: Different people.
15 Predominantly NuScale internal experts familiar with 16 the plant design and the transient response. It was 17 also reviewed by some folks with broader industry 18 experience outside of Nu -- from outside of NuScale.
19 MEMBER BLEY: Any reason you didn't 20 identify them this time? It seems odd to me, that's 21 all.
22 MS. McCLOSKEY: Didn't consider it 23 necessary to include in the topical report in this 24 space.
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115 1 RAI. Well, we cannot ask for it. But we can ask for 2 a supplement to this presentation --
3 MEMBER BLEY: Right.
4 MEMBER MARCH-LEUBA: -- to give us a list 5 if you feel it's important.
6 MEMBER BLEY: It just feels funny. You're 7 counting on experts and you don't say who they are.
8 MS. McCLOSKEY: Yes.
9 MEMBER MARCH-LEUBA: In your opinion, 10 Megan, was this intentional or was this an oversight 11 that you didn't think it was important?
12 MS. McCLOSKEY: I think that the PIRT 13 development was quite thorough and detailed. And that 14 that's evidenced in the summary of the PIRT that's 15 described in the topical report, and didn't rely 16 necessarily on the specific individuals involved.
17 MEMBER MARCH-LEUBA: What I was telling 18 Matt during the break is it's easier to get a scoping 19 calculation done than justify why you didn't do it.
20 MS. McCLOSKEY: Uh-huh.
21 MEMBER MARCH-LEUBA: How much work will it 22 take to provide us, send us a list this afternoon and 23 you give it to us?
24 MS. McCLOSKEY: I have the, I have the 25 PIRT and I can show you the, the committee and the who NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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116 1 reviewed it.
2 MEMBER MARCH-LEUBA: It's easier to get it 3 done than to justify why you didn't do it. I mean, it 4 will take you, what, 10 minutes? So, and you can put 5 it on the record this afternoon when we're in the 6 closed, closed session.
7 MS. McCLOSKEY: Right.
8 So, after the PIRT was developed, and 9 again this was a fairly comprehensive and integrated 10 PIRT that covered the different ranges of event types 11 and the phases, and it was -- and comprehensive in 12 terms of the range of phenomena addressed that were 13 focused on the non-LOCA events themselves and not 14 specifically for the system thermal hydraulic 15 response.
16 So, after the PIRT was developed we did a 17 GAP analysis to determine how we would address the 18 high-ranked phenomena identified. And that can be --
19 the way we addressed it falls very broadly into four 20 categories. There are phenomena where the validation 21 performed as part of the NRELAP5 assessments for LOCA 22 is considered efficient -- sufficient and applicable.
23 We identified where additional validation 24 or benchmarks for non-LOCA were warranted, primarily 25 in the steam generator and DHRS regions.
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117 1 Several phenomena were focused on or are 2 addressed by the downstream sub-channel analysis.
3 And then other areas we bounded the issues 4 of conservative input.
5 So, here it is identified from the GAP 6 analysis, again related to DHRS modeling of heat 7 transfer. The topical report describes validation 8 against the KAIST test and the separate effects test 9 done at NIST for full DHRS heat removal, as well as 10 plant sensitivity calculations we'll discuss further 11 in the closed session some more for steam generator 12 modeling and heat transfer, the validation based on 13 the large-scale testing done at the CX facility for 14 the secondary side. Heat transfer and pressure drop 15 in the helical coil steam generators. And the fluid 16 heated test as well.
17 We did a reactivity event responses, more 18 important, can be more important in some of the non-19 LOCA events. We also did a benchmark against the 20 RETRAN-3D code.
21 And, finally, we looked at the integral 22 response with testing done at the NIST facility.
23 MEMBER MARCH-LEUBA: On the steam 24 generator heat transfer coefficient, DS-2 -- you know, 25 my favorite topic is on the secondaries, so I could NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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118 1 spend a whole week running TRACE calculations on 2 reviewed and verified on containment body only.
3 But what I'm seeing is that the power 4 transfer from primary to secondary installations 5 oscillates a lot in a single -- on each two. Right?
6 And RELAPS 1300 might be something different.
7 But the other it wasn't seemed to change 8 much, if anything it increases. Did you guys even try 9 to -- because now the design we have on the table says 10 we don't care if the secondary oscillates. Indeed, we 11 expect it to oscillate. However, did you keep this 12 coefficient like you mentioned in DS-2 agrees with 13 what I'm calculating with TRACE? Or did it change a 14 lot during oscillations? Did you even attempt to 15 measure it?
16 I'm sure you measured it in the steady 17 state.
18 MS. McCLOSKEY: My recollection of what we 19 looked at as part of the stability topical report was 20 that there's a relatively small change in the overall 21 heat transfer coefficient that we observed based on 22 the DS-2 analysis results.
23 DR. CORRADINI: That was the Appendix A?
24 MS. McCLOSKEY: I believe so.
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119 1 that the flow oscillations are uncorrelated two to 2 two, when you do the other, you get the square root of 3 N, and then when it is, well, don't assume it isn't, 4 and it does not transfer to the core.
5 MS. McCLOSKEY: Right.
6 MEMBER MARCH-LEUBA: But the heat transfer 7 itself, if you put it to oscillate, it goes like that.
8 You can see it. You can plug the power in TRACE and 9 it oscillates just as much as the flow does. Behind 10 that, so the whole oscillation doesn't change that 11 much.
12 MS. McCLOSKEY: And we have some 13 additional discussion in the closed session about the 14 effect of the primary to secondary side heat transfer 15 in the steam generator and the effect on the initial 16 safety conditions as well as the non-LOCA transient 17 progression. And what we observed in the range of 18 sensitivity calculations is that it affects where your 19 steam generator is operating initially, but doesn't 20 have a significant impact on your margins to 21 acceptance criteria.
22 MEMBER MARCH-LEUBA: My point is if the 23 design says it is going to be unstable and going to be 24 oscillating like crazy, your methodology should assume 25 that. And I don't think it does.
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120 1 MR. PRESSON: And I would clarify that we 2 don't -- our design doesn't present itself as 3 oscillating on the secondary side like crazy. But we 4 do join that to a DWO. It's part of our assessment to 5 protect against that if it occurs, but the intent is 6 not for that to happen.
7 MEMBER MARCH-LEUBA: Yeah, but I haven't 8 seen any evidence that that's --
9 MR. PRESSON: Yeah. To be hopefully 10 resolved.
11 CO-CHAIR KIRCHNER: I think what works to 12 their favor for most of these transients is that, 13 you're right, even though there is a lot of 14 oscillation it's masked because you've got so much 15 thermal inertia on that side.
16 MEMBER MARCH-LEUBA: On -- yeah.
17 CO-CHAIR KIRCHNER: So, individual tubes 18 having oscillating heat transfer properties is a 19 concern from fatigue, and wear, and vibration, and 20 other matters but probably for these analyses the 21 large volume in the primary system just will mask it.
22 And it won't be a critical factor.
23 MEMBER MARCH-LEUBA: The first thing they 24 do is trip the reactor and isolate the secondary. So, 25 it really doesn't have an impact on the AOOs. And we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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121 1 have so much margin.
2 But, again, it's easier to have done it 3 and say, see, it doesn't have it, than say, no, I 4 don't want to do it. Again, that's philosophy.
5 And again, with the DHRS same problem.
6 You did a lot of analysis with non-condense -- the 7 biggest problem with DHRS, there are two problems with 8 DHRS operating: non-condensables that block the steam 9 from going from one sector to the other; or over 10 filling.
11 Did any of the tests that you benchmarked 12 against have non-condensables in them? Because over 13 filling is something very different.
14 MS. McCLOSKEY: Non-condensables were not 15 specifically introduced into the tests as an objective 16 of the tests that we validated against.
17 In the DHRS design there is a level switch 18 that is -- that will indicate what if there is 19 sufficient stimulation of non-condensables towards the 20 top of the DHRS that it would expect to be inoperable.
21 So that will be recognized --
22 MEMBER MARCH-LEUBA: The key purpose of 23 the DHRS you assume your non-condensables are to limit 24 to each level, in that you have non-condensables just 25 above the trip. You should assume that; right?
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122 1 MS. McCLOSKEY: What we, what we did --
2 MEMBER MARCH-LEUBA: I know I've seen some 3 sensitivity to condensables.
4 MS. McCLOSKEY: That was done as part of 5 the DHRS thermal hydraulic analysis and as part of the 6 design development. What we see in non-LOCA transient 7 events is you buy that DHRS heat transfer plus or 8 minus 30 percent is not having a significant impact on 9 our margins to acceptance criteria. So, we haven't 10 specifically initialized with non-condensables in the 11 calculations.
12 MEMBER MARCH-LEUBA: Just put in my 13 concerns. It can fail, too. It's just highly 14 unlikely to.
15 MR. BRISTOL: Yeah. And just to build off 16 of the thermal hydraulic analysis demonstrated that 17 for the period of time of the non-LOCA short-term 18 transient a\progression the system pressures are much 19 too high that that mass concentration is negligible.
20 Therefore, it is not specifically added as part of the 21 modeling techniques in the non-LOCA EM.
22 MEMBER MARCH-LEUBA: Well, the AOOs do not 23 assume the worst expected conditions for non-24 condensables?
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123 1 non-condensables in that loop for the AOO event 2 analysis.
3 MEMBER MARCH-LEUBA: Yeah, I'm not saying 4 you model there. But assume the heat transfer 5 calculation of DHRS assuming the worst tech spec non-6 condensables possible. That's what tech specs are 7 for. I mean, you have to demonstrate that you have --
8 I can only have this much non-condensables, and that's 9 what I'm assuming on my, on my analysis.
10 MR. BRISTOL: Yeah. I think I'd answer 11 that with we've done some pretty extensive 12 sensitivities looking at the influence of pool 13 temperature and the initial steam generator mass on 14 the DHR performance, and generally concluded that the 15 DHR performance isn't very sensitive for figures of 16 merit.
17 So, we stopped. We stopped once we'd 18 reached that conclusion.
19 MEMBER MARCH-LEUBA: Okay.
20 MS. McCLOSKEY: So, our overall conclusion 21 with the work done to support the applicability is 22 that the code with the MTM system model is applicable 23 for calculations of the MTM non-LOCA system response.
24 The overall non-LOCA transient analysis 25 process is described in the topical report Section 4.
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124 1 And this is fairly straightforward and typical.
2 We developed a plant-based model in the 3 NRELAP5 that includes geometry, the control systems, 4 protection systems. That base model is adapted as 5 necessary for the specific event analysis model, the 6 malfunction that's causing the initiating event, and 7 the desired initial conditions survives the transient 8 response.
9 We've taken, we perform our steady state 10 and transient analysis calculations in the steady 11 state calculations. The topical report identifies a 12 list of various steady state conditions. And those 13 are prioritized depending on the transients of 14 interest. And the results are compared to the 15 intended design conditions, and the transients are 16 performed as restarts.
17 We evaluate the results of the transient 18 analysis calculations to confirm our margins to 19 acceptance criteria for the pressure, primary and 20 secondary pressure. And confirm we have an 21 appropriate run time to demonstrate a safe, stabilized 22 condition is achieved.
23 And then we identify cases, as applicable, 24 for downstream sub-channel or radiological analysis.
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125 1 directions are to maximize reactor power as well as 2 core exit pressure and core inlet temperature. And 3 that's based on the conservative directions for the 4 NuScale-specific CHF correlation as well as minimum 5 RCS flow rate, which is typical.
6 As was mentioned earlier in the discussion 7 about the topical report, we can use the NRELAP5 CHF 8 calculation from a dummy hot rod model of a screening 9 tool to reevaluate it to identify there weren't any 10 cases, to evaluate downstream analysis.
11 And for events such as the reactivity 12 event, uncontrolled rod withdrawal, single rod 13 withdrawal, where our margin is more limited in CHF, 14 we'll analyze more sub-channel analysis cases. And 15 typically all of the sub-channel, all of the cases get 16 sent to the sub-channel analysis to confirm we've 17 captured the limiting CHF result.
18 Similarly for radiological analysis, we 19 identified cases for maximum release and iodine 20 spiking time.
21 MEMBER MARCH-LEUBA: Okay. Point five, 22 second bullet, the word "may" is giving me a 23 heartache. I mean, when you have a topical report to 24 approve, the purpose was that you were going to place 25 it in the licensing basis as a reference that can be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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126 1 used for future design licensing safety analysis 2 reports, so for cycle 21 they want to use this Dash-A 3 methodology and submit it to the staff for review or 4 for the approval.
5 And all the staff has to do is ask 6 themselves did they use the Dash-A methodology to 7 calculate the set points? That's how the system is.
8 I'm complaining that the methodology is 9 not descriptive enough when you say it may or may not 10 be used. It is used.
11 MS. McCLOSKEY: Yes. What we intended to 12 indicate there was that for some events like a, like 13 the increase in inventory events or CBCS malfunction 14 that has a very limited impact on the overall module 15 core condition, we may use -- we'll use the screening 16 tool to identify one, maybe one limiting case against 17 them for sub-channel analysis.
18 And that's sufficient to demonstrate for 19 that event that it's not limiting with respect to CHF 20 margins.
21 For other events, like reactivity events, 22 we'll do more analysis into the sub-channels.
23 MEMBER MARCH-LEUBA: The problem I have is 24 that the CHF correlation that you guys chose to 25 implement for the non-LOCA calculation, it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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127 1 incredibly biased, I mean, it's grossly non-2 conservative. And we're not talking 10 percent here.
3 Right, when we want to do assessment I can talk the 4 number, but it is bad, bad, bad.
5 The only thing that saves you is that it 6 seems to follow the same phase as the good 7 correlations. But it's incredibly non-conservative.
8 MS. McCLOSKEY: And we're not using it to 9 demonstrate margin to the CHF acceptance criteria.
10 It's only to identify the trend.
11 MEMBER MARCH-LEUBA: So what I would like 12 to see as a condition on the staff will see, it 13 specifies you identify the top N by transience. And 14 those N's you analyze VIPRE where N can be 1, 2, 5, 15 12. And we agree on what is reasonable. Or one from 16 each of the categories or something like that.
17 To be an approved methodology it has to be 18 well defined, and I don't think it is well defined.
19 The way you have it is trust that it will work for 20 Cycle 21.
21 And when the staff is seeing that section 22 of this report all they need to do is check -- they 23 tell me they use the Dash A approved methodology, I 24 don't have to review it. And if the Dash A approved 25 methodology is wishy-washy then I cannot do that.
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128 1 And the staff doesn't want to be reviewing 2 your analysis 21 years from now.
3 So that, I will bring that to the piece up 4 here and we'll have to suffer with it. But I think 5 that prescriptive it's easier for you, it's easier for 6 them, it's easier for everybody instead of 7 flexibility. Flexibility is bad.
8 And, of course, myself, I would have run 9 every single VIPRE case because CPUs are cheap. And 10 you should have it to alternators, so all you have to 11 do is push a button and everything comes out. But 12 that's in my -- I give my thoughts.
13 Please continue.
14 CO-CHAIR KIRCHNER: Let me ask one 15 question since on that same bullet how much 16 conservatism do you pile into this hot rod, this 17 "dummy hot rod"? Do you load up all the biases, 18 whether it's stored energy or initial conditions or 19 whatever, and then use it, may use it, or use it as a 20 screening tool? Or is it an average rod?
21 Hot rod leads me to believe that you put 22 probably all of the EM biases into that particular 23 analysis before you use it as a screening tool.
24 MS. McCLOSKEY: The hot rod is separate, 25 a separate part of the RELAP model that has no NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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129 1 feedback to the system thermal hydraulic response. I 2 would have to look up the specific biases that go into 3 that.
4 MR. BRISTOL: Yeah, so, we applied the 5 radial peaking typically, but we didn't necessarily go 6 through an effort of tuning it so that it matched the 7 actual margin calculations better because once we'd 8 demonstrated that the trends were sufficient then 9 that's where we, that's where we stopped in terms of 10 the development of that model.
11 In the cases like Megan was alluding to, 12 the reactivity events, the power changes as a function 13 of the -- or the power distribution changes as a 14 function of the initial condition. And there's some 15 specific sub-channel techniques for how they address 16 that. We don't try to adopt any of that and, 17 therefore, that's a place where there's a lot more 18 cases that are sent for sub-channel analysis.
19 MS. McCLOSKEY: All right. In the topical 20 report, Section 7.1 discusses methodology that's 21 generally applicable to all of the events in terms of 22 basic conditions that are considered and prioritized 23 and met. The treatments of plant controls, that if a 24 non-safety related plant control system is a benefit 25 to the transients but not credited in the non-LOCA NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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130 1 analysis, loss of power, the analyses consider whether 2 power available or loss of AC, or loss of AC and DC 3 power is more limiting for the acceptance criteria.
4 Or the single failure of a safety-related component in 5 the analyses. This is the section that discussed the 6 founding reactivity input, predominantly the moderator 7 temperature coefficient, the Doppler temperature 8 coefficient, as well as the decay heat pump 9 contribution and SCRAM worth.
10 And it discusses other parameters such as 11 the valve characteristics and the modular protection 12 system and the limits and response times.
13 No operator actions are required to 14 achieve the safety functions, so that's identified in 15 this section as well.
16 And then in the event-specific methodology 17 subsections of 7.2 for each event, there's a 18 description of the initiation and progression. We have 19 put those criteria of interest which identify which 20 criteria would be more challenged in this event and, 21 therefore, more sensitivity calculations are done for 22 that criteria compared to other events.
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131 1 pressure and secondary pressure response.
2 The limiting single failures and lower 3 loss of power scenarios for the events are identified, 4 or the need to submit a calculation is identified for 5 the event as well. And similarly, in the initial 6 condition biases and conservatism there's a table for 7 each event of whether the initial condition is biased 8 or if it's sampled in the analysis.
9 And examples are presented in Section 8.
10 So, overall the evaluation model was 11 developed on a graded approach following the guidance 12 in Reg Guide 1.203. It is applicable only to the MTM 13 type plant design with natural percolation and helical 14 coil steam generator and coil pressurizer meets the 15 NRELAP5 code as we've been discussing to demonstrate 16 primary and secondary pressure criteria are met, and 17 that a safe, stabilized condition is achieved, and 18 provides boundary conditions to downstream sub-channel 19 and radiological analysis.
20 MEMBER MARCH-LEUBA: Thank you very much.
21 I think we have reached the end of the open 22 presentation from NuScale. You can go and enjoy being 23 part of the audience in the back while we make the 24 staff --
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132 1 at 11:30. We're scheduled to have lunch at 11:30.
2 Let's start with the staff after lunch.
3 So, we'll take a recess until 12:30 on 4 this clock.
5 MEMBER MARCH-LEUBA: Which is now correct.
6 CO-CHAIR KIRCHNER: Yeah. And we're, just 7 to calibrate everyone, we're about one presentation 8 behind in terms of time.
9 So, with that, we're recessed.
10 (Whereupon, the above-entitled matter went 11 off the record at 11:26 a.m. and resumed at 12:30 12 p.m.)
13 CO-CHAIR KIRCHNER: We're going to 14 recommence our meeting and turn to Chris Van Wert from 15 the staff. Please go ahead, Chris.
16 MR. VAN WERT: All right. Thank you very 17 much. Good, I guess, afternoon now. So, yes, this is 18 the staff's presentation. It's an open session 19 presentation. Hopefully it will answer a few of the 20 questions that you brought up this morning. And with 21 that I will go ahead and get started.
22 Again, I'm Chris Van Wert. The reviewers 23 were myself and Jeff Schmidt. Becky is our branch 24 chief.
25 Just as a quick overview Rev. 1 of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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133 1 topical report was submitted just a few months ago in 2 November. We do have a full committee scheduled for 3 March 5th. And if there are no changes, we will be 4 working on completing the dash A version of this 5 topical report soon, but in the very least it will be 6 done by Phase 6 of the DCA.
7 All right. As a high overview of what the 8 staff's review included, we looked at -- we evaluated 9 the analysis criteria, also the code suite that was 10 used within the analysis methodology, the evaluation 11 of the plant and cycle assumptions that were used, and 12 also the rod ejection accident analysis methodology 13 itself.
14 I do think it's important to just remind 15 everyone that this does not include the actual 16 licensing basis analysis that is contained in the DCA 17 itself in Section 15.4.8, but they do include an 18 extremely similar example calculation in Section 6 of 19 the topical report, but it is not the licensing basis 20 itself. In addition of course to the review of the 21 materials provided we did audit some of the 22 calculations that were used to support their 23 application.
24 To talk a little bit about the analysis 25 criteria, I want to give you a brief overview of what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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134 1 they did and our evaluation of it.
2 For the most part they followed the SRP 3 Appendix B -- sorry, SRP Section 4.2, Appendix B 4 Guidelines guidance. The first item, reactor coolant 5 system pressure, they kept with that guidance and used 6 maximum RCS pressure must be less than or equal to 120 7 percent design pressure.
8 For fuel cladding failure they did make a 9 couple of modifications, but they were in a 10 conservative direction. The first criteria is in 11 regards to high temperature at zero power conditions.
12 They set their limit to 120 calories per gram for all 13 cladding differential pressures whereas the SRP allows 14 for some increased enthalpy values for smaller 15 differential pressures. So it took the most limiting 16 one and stuck with that for everything.
17 For intermediate and full power conditions 18 they assumed failure for heat flux -- or local heat 19 flux exceeds the CHF limits. It's also worth pointing 20 out that whereas the SRP would allow you to fail that 21 rod and use that as part of your fission gas inventory 22 they do not allow it to move forward in the core 23 design if they have a failure there.
24 CO-CHAIR MARCH-LEUBA: Hey, Chris --
25 MR. VAN WERT: Yes?
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135 1 CO-CHAIR MARCH-LEUBA: -- the figures --
2 the one -- the slide they showed us this morning, the 3 one that has two figures --
4 MR. VAN WERT: I'm coming to it.
5 CO-CHAIR MARCH-LEUBA: All right.
6 (Laughter.)
7 MR. VAN WERT: Two more slides and I'll --
8 CO-CHAIR MARCH-LEUBA: It's inconsistent 9 with your numbers there, you know that?
10 MR. VAN WERT: Yes, and we were discussing 11 that right before lunch. And I'll explain --
12 CO-CHAIR MARCH-LEUBA: Did somebody change 13 their mind?
14 MR. VAN WERT: What was that?
15 CO-CHAIR MARCH-LEUBA: Did somebody change 16 their mind?
17 MR. VAN WERT: There was an inconsistency 18 which I will explain --
19 CO-CHAIR MARCH-LEUBA: Okay.
20 MR. VAN WERT: -- two slides from now.
21 CO-CHAIR MARCH-LEUBA: I'll patiently 22 await --
23 (Laughter.)
24 MEMBER BLEY: We're still in open session.
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136 1 session.
2 MR. VAN WERT: Yes. Yes. Well, yes.
3 MEMBER BLEY: Yes.
4 MR. VAN WERT: But I did confirm with 5 NuScale that I can mention certain numbers that are 6 listed as non-probable within the report.
7 The PCMI failure threshold, which is what 8 you're referring to, in Section 2.3.2 of their topical 9 report they request the threshold as the figure B-1 10 from the SRP. That's what we initially did our review 11 towards and that's going to lead to the clarification 12 here in a second. And also failure is assumed at the 13 fuel temperature anywhere, and the fuel pellet exceeds 14 incipient fuel melting conditions.
15 To finish off some of the criteria, core 16 coolability, they stuck with the same as the SRP. You 17 must remain below 230 cals per gram maximum. And then 18 fuel melt is not allowed.
19 For fission product, since they do not 20 allow a core design that results in clad failure due 21 to melting enthalpy increase or exceeding MCHFR 22 limits, and also by incorporating some of the more 23 limiting differential pressure-based limits, they 24 preclude the need to follow the guidance that's 25 provided in the SRP 4.2, Appendix B regarding fission NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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137 1 gas release. So they're a little bit more 2 conservative in that perspective.
3 CO-CHAIR MARCH-LEUBA: This methodology, 4 once approved, is going to be applicable to power 5 uprates --
6 MR. VAN WERT: Correct.
7 CO-CHAIR MARCH-LEUBA: -- which may or may 8 not come sooner or later.
9 MR. VAN WERT: Right.
10 CO-CHAIR MARCH-LEUBA: They will be frozen 11 to this -- I mean these are more restrictive criteria 12 than you would need.
13 MR. VAN WERT: Right. Right.
14 CO-CHAIR MARCH-LEUBA: It's easier to 15 prove.
16 MR. VAN WERT: Yes, so the criteria --
17 we're approving these criteria. They're consistent 18 with our guidance or more restrictive.
19 CO-CHAIR MARCH-LEUBA: Yes.
20 MR. VAN WERT: They do these analyses on 21 a cycle-by-cycle basis, and we'll get a little bit 22 more into this discrepancy here and you can see --
23 CO-CHAIR MARCH-LEUBA: I'm not talking the 24 discrepancy.
25 MR. VAN WERT: Oh, okay. About the no-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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138 1 fuel failures?
2 CO-CHAIR MARCH-LEUBA: Yes.
3 MR. VAN WERT: Yes, that's their criteria.
4 So they cannot fail it and then --
5 CO-CHAIR MARCH-LEUBA: Are they married to 6 this criteria for power uprates?
7 MR. VAN WERT: Until Rev. 1 comes in of 8 the -- or Rev. 2 of the topical report comes in --
9 CO-CHAIR MARCH-LEUBA: Unless --
10 MR. VAN WERT: -- addressing that issue, 11 they are married to it.
12 CO-CHAIR KIRCHNER: And by then the DG 13 should be --
14 MR. VAN WERT: Pardon?
15 CO-CHAIR KIRCHNER: If they come back for 16 power uprates, by then I expect the Draft Guide, your 17 last bullet --
18 MR. VAN WERT: Correct. Correct.
19 CO-CHAIR KIRCHNER: -- will have been 20 released.
21 MR. VAN WERT: Yes. So that leads into 22 the discussion that we are aware that there is a draft 23 guidance that's been in the works for a number of 24 years now. It has been out in the public for comment 25 multiple times, however, at the time the submittal of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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139 1 the application the topical report, the guidance in-2 house in play was the SRP 4.2, Appendix B.
3 With that being said, we knew that there 4 were going to be some questions. So there was a 5 question this morning regarding kind of comparing 6 apples to oranges when you have a corrosion-based 7 limit figure and then a hydrogen-based figure.
8 Luckily I was able to talk with Tim 9 Drzewiecki who is a colleague of mine. I don't see 10 him in the audience right now, but he was willing to 11 make a MATLAB script and using some oxidation limits 12 that are calculated for NuScale and then the freed 13 hydrogen from the zirc water reaction and then the 14 hydrogen pickup information developed in -- or 15 presented in -- I think it's also a Draft Guide, 4.224 16 or 6. One of those two. You kind of do a translation 17 and put the SRP figure B-1 into the newer parameters.
18 I also want to highlight again that the 19 blue figure here, which is from the draft guidance, 20 that is literally draft and I do not want to stand 21 behind it in saying that this is what would eventually 22 become the final Reg Guide, but it just give you a 23 flavor for trying to get as close to an apples-to-24 apples comparison as we could.
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140 1 question.
2 MR. VAN WERT: Yes.
3 MEMBER BLEY: Once the Draft Guide is a 4 Reg Guide, the SRP's guidance to the staff on how to 5 review the Reg Guide is guidance to applicants on how 6 to kind of meet the regulation. Do you immediately 7 make a change to the SRP or does the staff guy have to 8 try to figure out which one to apply?
9 MR. VAN WERT: I mean immediate as 10 immediate as the NRC moves, but yes, it would be 11 something that would have to be updated as part of SRP 12 4.2. Yes.
13 MEMBER BLEY: Okay. But that might take 14 years, yes.
15 MR. VAN WERT: Yes. I'm not saying the 16 timeline, but it would be eventually done, yes. I 17 think it would be done relatively quickly because that 18 would be a pretty glaring discrepancy.
19 So with that in mind, if we were to look 20 at this and -- yes, if you happen to have the topical 21 report in front of you, you can look at tables 6-2 and 22 6-3, and that provides some of the analysis results 23 for the sample equilibrium cycle analysis. Again, 24 that's not the analysis of record, but that will give 25 you an idea of where the numbers kind of fall. And I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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141 1 think the maximum change in enthalpy was around 29 2 cals per gram. So I think there was a question 3 earlier about how far within that dotted box they 4 were. They're well within it.
5 Now as far as the question goes why was 6 that dotted box on there versus me talking about the 7 SRP line on here, the topical report in Section 2.3.2 8 requests the figure. They basically copied the figure 9 from B-1, put it in their report, but it's the same 10 figure. So that's I think figure 5-2. One or two?
11 PARTICIPANT: Two.
12 MR. VAN WERT: Two? Okay. So they 13 request that, but then when you go to Section 5.3, 14 they discuss how their maximum oxide calculated for 15 their plant is significantly lower, and they kind of 16 put a little bit of margin on top of that. And then 17 they ran it up and came out to be around 75 cals per 18 gram and said, okay, we're going to just live within 19 this box so that it's very clear. It doesn't matter 20 if there's a revision to the Reg Guide that comes out 21 later. They will be in good standing for meeting all 22 the future requirements.
23 I would not be surprised if later if they 24 come in with a revision. They could change things, 25 but as long as it lived within our guidance we are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 fine. We can't impose a stricter limit without some 2 heavy justification.
3 Any further questions on this figure?
4 (No response.)
5 MR. VAN WERT: Okay.
6 MEMBER BLEY: Just for the record we are 7 aware of the status of DG-1327, so we will probably 8 after this summer timeframe we look to schedule Mr.
9 Clifford to come before us and present that.
10 MR. VAN WERT: Right.
11 PARTICIPANT: If he's still employed.
12 MEMBER BLEY: If he's still -- yes, well, 13 part of that, we want to catch him before he goes with 14 his body of knowledge, so to speak. Thank you.
15 MR. VAN WERT: All right. So a lot of the 16 codes were discussed this morning that were used in 17 the SPERT. In general as far as the valuation goes 18 the analysis is based on mostly the Studsvik code 19 suite, CASMOS5, SIMULATE5 for the core physics and the 20 initialization, SIMULATE-3K for the 3D nodal kinetics, 21 NRELAP5 for the transient system response, and VIPRE 22 for the sub-channel analysis.
23 Most of these: CASMOS5, SIMULATE5, NRELAP5 24 and VIPRE, have been reviewed and approved in a 25 separate topical report, however SIMULATE-3K, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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143 1 validation for that code was supplied within this 2 topical report. So our review did include that.
3 We looked at their validation which was 4 basically two parts. They validated against external 5 data, the SPERT III experiments. They also validated 6 against a benchmark from the NEACRP control rod 7 injection problem, and the staff did it acceptable.
8 DR. SCHULTZ: Chris?
9 MR. VAN WERT: Yes.
10 DR. SCHULTZ: This is not the first time 11 that the staff's reviewed SIMULATE-3K, correct? And 12 other licensees --
13 MR. VAN WERT: Correct, but it was not --
14 DR. SCHULTZ: -- have utilized it?
15 MR. VAN WERT: Yes, other licensees. It 16 was not --
17 DR. SCHULTZ: Right.
18 MR. VAN WERT: -- a stand-alone Studsvik 19 product, so they couldn't just refer to an approved 20 topical.
21 DR. SCHULTZ: Got you. I see. So it came 22 in on --
23 MR. VAN WERT: Yes.
24 DR. SCHULTZ: -- a separate document, a 25 different document?
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144 1 MR. VAN WERT: Right, there's an operating 2 plant or owner that --
3 DR. SCHULTZ: Yes.
4 MR. VAN WERT: -- for their fleet. The 5 have seen that.
6 DR. SCHULTZ: I recall similar approaches 7 in terms of the validation in the --
8 MR. VAN WERT: Right.
9 DR. SCHULTZ: -- application.
10 CO-CHAIR KIRCHNER: So, Chris, just as a 11 follow-up on Steve's point, as a process question. So 12 you accepted SIMULATE-3K as part of this TR. Does 13 that mean SIMULATE-3K is approved as --
14 MR. VAN WERT: It's approved for use in 15 NuScale's rod injection accident methodology.
16 CO-CHAIR KIRCHNER: And that's the limit 17 of the approval?
18 MR. VAN WERT: Yes, and it --
19 CO-CHAIR KIRCHNER: So it's not broader?
20 They couldn't come back say and say we want to use 21 SIMULATE-3K for some other purposes or --
22 MR. VAN WERT: Correct. Correct.
23 CO-CHAIR KIRCHNER: Okay.
24 MR. VAN WERT: Now I don't want to --
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145 1 asking for approval for --
2 (Simultaneous speaking.)
3 MR. VAN WERT: They could always come back 4 and request it --
5 CO-CHAIR KIRCHNER: Yes.
6 (Simultaneous speaking.)
7 MR. SCHMIDT: -- as a new application, but 8 this is only applicable to NuScale rod ejection.
9 CO-CHAIR KIRCHNER: But I guess SIMULATE-10 3K does not have generic approval?
11 MR. VAN WERT: Correct.
12 CO-CHAIR KIRCHNER: All right.
13 MR. VAN WERT: Yes, I would normally 14 expect Studsvik as a vendor to come in with the 15 topical report and then try to sell that off to 16 potential customers, but I have not heard of any 17 potential plans of that coming in anywhere in the near 18 future.
19 CO-CHAIR MARCH-LEUBA: And while we're 20 interrupting you, this applicant talked about 21 uncertainties in parameters.
22 MR. VAN WERT: Yes.
23 CO-CHAIR MARCH-LEUBA: And we have the 24 opportunity of as part of the ACRS research quality 25 review process -- which we review a couple of NUREGs NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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146 1 from research. We reviewed one on uncertainty of 2 nuclear parameters. And the conclusion from the NUREG 3 is that there is a tremendously high and unexpected 4 uncertainty on beta effective. And the numbers 5 published there based on the scale calculations are 20 6 to 100 percent error, which even the authors, when we 7 confronted them, said that was probably an error on 8 the -- in the ENDRB data.
9 PARTICIPANT: Or processing error.
10 CO-CHAIR MARCH-LEUBA: Or processing. I 11 mean how confident are we that the beta effective that 12 CASMOS produces is not 25 percent off?
13 MR. SCHMIDT: I mean we have been using 14 that beta effective like from CASMOS3 to 5 for a long 15 period of time. I don't know if I know the 16 uncertainty associated with that parameter. I mean 17 usually you look at beginning cycle and end of cycle 18 type values, but I don't recall the uncertainty 19 applied to beta effective here. This is what, 20 table --
21 CO-CHAIR MARCH-LEUBA: Yes, back when I 22 worked in ORNL they paid me to do an analysis of 23 uncertainty, of how to generate cross-section for 24 TRACE-PARCS. And I wasn't qualified to do the job, so 25 I just passed the money to somebody else. And in that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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147 1 NUREG, which I can give you a copy of, we found out if 2 you do the weighting of it affected with fission power 3 versus weighting with the flux, there was like a 15-20 4 percent error.
5 So the uncertainty was the effect issue.
6 We should maybe -- I mean it's not 100 percent because 7 we'll never be able to match a SPERT.
8 MR. SCHMIDT: Right. Yes.
9 MR. VAN WERT: And that's probably --
10 CO-CHAIR MARCH-LEUBA: But this is large.
11 MR. VAN WERT: I would expect that that 12 would be in the Nuclear Analysis Codes and Methods 13 Qualification Topical.
14 CO-CHAIR MARCH-LEUBA: Yes.
15 MR. SCHMIDT: The actual beta effective?
16 MR. VAN WERT: Yes.
17 MR. SCHMIDT: That's in the --
18 (Simultaneous speaking.)
19 MR. VAN WERT: Yes, I was trying to --
20 MR. SCHMIDT: Yes.
21 MR. VAN WERT: -- quickly look it up and 22 I didn't see it. But since CASMOS5 I think was 23 covered under that separate topical report, I would 24 imagine that's where it is.
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148 1 parameters where it cannot possibly be that large, as 2 EMDAP says, but other evaluations kind of point in the 3 line that maybe. And it has a tremendous impact on 4 this answer. I mean, you know what I mean? The 5 direct impact, real fast. So I'm just putting it out 6 there.
7 MR. VAN WERT: All right. Thank you.
8 CO-CHAIR MARCH-LEUBA: And I can -- we can 9 find NUREG numbers to put them on the record. Let me 10 look for them and I'll --
11 PARTICIPANT: I have it.
12 CO-CHAIR MARCH-LEUBA: Oh, you have them?
13 PARTICIPANT: Yes, they're within my 14 notes.
15 MR. VAN WERT: You want me to continue?
16 Are we looking up the --
17 CO-CHAIR MARCH-LEUBA: Continue. We'll 18 wait. Get a break and put it on the record.
19 MR. VAN WERT: Okay. So the staff also 20 reviewed the plant and cycle assumptions used in the 21 analysis are within the methodology. We determined 22 that the methodology included ranges in power, time of 23 cycle, core power that averaged a wide range of 24 operating conditions, and we captured the most 25 limiting condition for NuScale. The staff also agreed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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149 1 with the assumptions associated with the automatic 2 system response of non-safety systems and that they 3 were conservative and that the methodology for the 4 timing of loss of AC power conservatively biases the 5 RCS pressure evaluation.
6 And as far as the REA methodology, the 7 staff reviewed the analysis methodology including 8 steady-state initialization, the dynamic core 9 response, the dynamic system response, sub-channel 10 heat flux and the adiabatic heat-up fuel response. In 11 that review the staff's review included the 12 methodology by which information is passed between the 13 codes, the application of the uncertainties and the 14 modeling assumptions used for the inputs and handling 15 of the reactor trips. We concluded that the 16 methodology for calculating the system response, sub-17 channel and fuel response analyses were conservative 18 and acceptable for demonstrating compliance with the 19 acceptance criteria.
20 And in conclusion, overall for the staff's 21 review we agreed with them that the REA methodology 22 either follows or is more conservative than the staff 23 guidance and that the methodology accounts for various 24 potential operating conditions in time and life and 25 conservatively addresses the uncertainties and plant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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150 1 conditions. Therefore, the staff finds that the use 2 of this topical report is acceptable for evaluating 3 REA for the NuScale plant design.
4 CO-CHAIR MARCH-LEUBA: Is there something 5 you can say in open session about the time in cycle 6 where this becomes worse? Because we mentioned in 7 open session this morning it wasn't at zero power, but 8 it was more -- something else.
9 MR. VAN WERT: But which part becomes 10 worse?
11 MR. SCHMIDT: Like what time in life?
12 Like conditions?
13 CO-CHAIR MARCH-LEUBA: Yes. We can wait 14 for closed.
15 MR. SCHMIDT: Yes, why don't we wait for 16 NuScale? Yes, because I think they're going to 17 present --
18 CO-CHAIR MARCH-LEUBA: Yes.
19 MR. SCHMIDT: -- analysis results.
20 MR. VAN WERT: They're presenting numbers 21 later, so --
22 CO-CHAIR MARCH-LEUBA: But you have 23 reviewed that and you --
24 MR. SCHMIDT: Yes, there was a whole --
25 there's a whole table on the top of the report that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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151 1 looks at powers, times in lifes, EOC, MOC, EOC.
2 CO-CHAIR MARCH-LEUBA: And you're happy 3 with their --
4 MR. SCHMIDT: Yes.
5 MR. VAN WERT: Yes.
6 MR. SCHMIDT: Yes, it was a full spectrum.
7 CO-CHAIR MARCH-LEUBA: Yes, I wanted to 8 make sure in the public record.
9 MR. SCHMIDT: Yes.
10 MR. VAN WERT: Yes.
11 CO-CHAIR MARCH-LEUBA: You're happy with 12 it?
13 MR. VAN WERT: Yes. Yes, we --
14 MR. SCHMIDT: That was the main focus.
15 CO-CHAIR MARCH-LEUBA: And in the closed 16 record we'll have more fun with the numbers.
17 MR. SCHMIDT: Okay.
18 MR. VAN WERT: All right. Sounds great.
19 All right. Any other questions?
20 CO-CHAIR MARCH-LEUBA: Just wanted to put 21 on the record since I was talking --
22 MR. VAN WERT: Yes.
23 CO-CHAIR MARCH-LEUBA: -- I found the 24 NUREG report. There are two: NUREG-CR-7249 and 25 NUREG-CR-7164. Those two address the beta effective NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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152 1 uncertainty and are coming from different points of 2 view, but it's something that for -- maybe not now, 3 but for future reviews you guys need to consider 4 because it has the potential of being bad.
5 MR. VAN WERT: Yes, thank you.
6 CO-CHAIR KIRCHNER: Well, maybe a 7 different way to say it without being pejorative is 8 that certain concepts will be much more dependent on 9 beta effective for their safe operation and shutdown.
10 And so that's why we flagged it as a concern. And 11 that will be reflected in our report to RES as part of 12 our biennial review.
13 MEMBER REMPE: It's not voted on yet, 14 so --
15 CO-CHAIR KIRCHNER: Not voted on, so --
16 (Laughter.)
17 CO-CHAIR KIRCHNER: We expect will be in 18 -- as part of our letter report.
19 DR. SCHULTZ: Chris?
20 MR. VAN WERT: Yes?
21 DR. SCHULTZ: Just to go back to the basis 22 of the staff approval, the discussion or the approach 23 that NuScale has taken to assign uncertainties 24 throughout their methodology and then the way in which 25 they combine those uncertainties in order to create a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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153 1 bounding evaluation, is that -- are those all 2 specifically identified as part of their methodology 3 that you have approved?
4 MR. VAN WERT: So some of this is actually 5 coming from the Nuclear Methods Topical Report. So 6 you're right that it is approved as part of REA, but 7 that's because it's also referencing an approved 8 topical report.
9 DR. SCHULTZ: It even goes further than 10 that in terms of their overall methodological 11 approach?
12 MR. VAN WERT: Correct.
13 DR. SCHULTZ: Thank you. That helps.
14 MR. VAN WERT: Thank you. I did my best 15 to try to put you on schedule, or at least a little 16 closer to schedule, but --
17 CO-CHAIR KIRCHNER: Thank you, Chris.
18 Yes.
19 Members, any further questions of Chris or 20 Jeff at this point? Because if not, we will move onto 21 the LOCA evaluation model and the staff's review of 22 the applicant's TR.
23 (No response.)
24 CO-CHAIR MARCH-LEUBA: And the same goes 25 for you guys. Chris doesn't want to hold onto your NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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154 1 name tags, so save them for tomorrow.
2 MEMBER REMPE: So, Walt and Jose, while 3 they're coming up I'm not sure if other members 4 noticed it, but yesterday while we were getting all 5 the new slides I noticed that the staff SE had 6 changed. And I had earlier versions, and so I had 7 reviewed those. But like the LOCA one had an increase 8 of three pages. The non-LOCA had an increase of seven 9 pages.
10 PARTICIPANT: Yes, I can explain.
11 MEMBER REMPE: Yes, I'd like some 12 explanation on the record, because that seems a bit 13 for editorial changes and I -- did anyone have a 14 chance to really carefully compare what caused that 15 increase in three pages and seven pages?
16 CO-CHAIR KIRCHNER: Well, maybe it's a 17 question that we can direct to the staff. I can't 18 answer that, but the agreement that we had to prepare 19 for this meeting was that the staff would give us as 20 near a final draft as possible mid-last month. And as 21 to the actual extent of the changes --
22 MEMBER REMPE: I'd like to know what they 23 are.
24 CO-CHAIR KIRCHNER: -- we can ask the 25 staff to address that.
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155 1 MEMBER REMPE: Yes, because I --
2 CO-CHAIR KIRCHNER: And maybe Becky can 3 help us here.
4 MS. PATTON: Yes, right. So I will tell 5 you that a couple of the changes that were made added 6 an amazing number of pages when you did it. We had to 7 add -- anywhere where we were referencing a figure 8 number they wanted the full title of the figure in 9 every single circumstance. So when you actually run 10 a compare; and I have some file compare versions on my 11 computer if anybody wants to see them, that added a 12 huge amount. Every time we said we audited anything, 13 they wanted the full ADAMS reference in there.
14 Other than that there were a few times 15 where maybe the lawyers wanted a little bit more of a 16 basis explanation, and so there was -- there were 17 maybe a couple of times where there were a few 18 sentences added, but there weren't like substantive 19 technical changes in any of these. It was a large 20 amount of changes due to editorial formatting things.
21 MEMBER REMPE: Okay. So I'm a trusting 22 soul on this. The fact that you're saying on the 23 record that there were no substantial changes I'll buy 24 off on it, but I just was surprised at the length of 25 the --
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156 1 (Simultaneous speaking.)
2 MS. PATTON: Yes, right, that's correct.
3 And, I mean, there are some explanatory type of basis 4 statements that were added in various places. Like I 5 said, I can -- we can make a compare file version 6 available if anybody wants it.
7 MEMBER REMPE: I doubt any of the other 8 members had a chance to go through the updated 9 versions and so I just wanted to make sure we 10 discussed it here today. So thank you.
11 CO-CHAIR MARCH-LEUBA: Since you offer, 12 send us a compare.
13 MS. PATTON: Okay. Sure.
14 CO-CHAIR KIRCHNER: All right. So we're 15 going to turn to the staff's review of the LOCA 16 evaluation method.
17 And I'll turn to you, Carl, right? Are 18 you heading up or --
19 DR. LU: It will be Shanlai Lu from NRR 20 staff, Division of Safety Systems, and Carl is going 21 to run the slides for me. Actually he can also answer 22 questions. We have been working together.
23 This is one of the most heavily invested 24 -- of the topical report from both NuScale and the 25 staff side. So we are dividing this presentation to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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157 1 two parts: The first part will be high-level summary 2 in -- for the non-public -- the public session. And 3 then we are going to have also proprietary sessions 4 later, so in details.
5 So for this particular part -- and then 6 I'm going to cover the -- the review team is here.
7 And you can see that because it's a -- really a lot of 8 effort from NuScale side, we started from -- almost 15 9 years ago started having the concept, developed this 10 idea of the reactor design. So we had a lot of folks 11 join together to form a team and then at different 12 time people spent time on that.
13 So from NRC side we have Carl Thurston and 14 Dr. Peter Lien from Office of Research, Dr. Weidong 15 Wang actually from ACRS staff running the TRACE code 16 in support of Office of Research, Ron Harrington from 17 Office of Research, myself, Mr. Antonio Barrett. He 18 worked on the CHF. Tim worked also on CHF. Syed 19 Haider, because it's really related to the 20 containment, the ECCS system. So he worked on this 21 one, too. And then we have our expert supposed to be 22 on the public line.
23 Do we have a bridge line open now? I 24 don't know. Is bridge here?
25 CO-CHAIR KIRCHNER: Well, it's muted.
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158 1 DR. LU: Okay. Yes, oh, it's muted?
2 Okay.
3 All right. So we have Dr. Leonard Ward.
4 He retired from the NRC and he has been a LOCA expert.
5 And Dr. Don Rowe, a thermal hydraulic expert from IEEO 6 previously. Now they all work for NuMark. And then 7 Bert Dunn from Areva. He was the lead principal 8 engineer on LOCA before he retired and joined the 9 NuMark. And then we also have Upendra from Brookhaven 10 National Lab. And then -- on the scaling. So that's 11 the review team. And we had quite a lot of input from 12 the -- our consultants and the national lab experts.
13 CO-CHAIR KIRCHNER: Let me interrupt you 14 just for a moment --
15 DR. LU: Sure.
16 CO-CHAIR KIRCHNER: -- and for the record 17 just say that Matt Sunseri, Chairman of the ACRS 18 Committee, has joined us. Okay.
19 DR. LU: Okay. All right.
20 CO-CHAIR KIRCHNER: Please proceed.
21 DR. LU: So let's take a quick look of the 22 design features and the scope. And actually NuScale 23 provided more detail. They include the DHRS, but as 24 part of a topical report since the LOCA methodology 25 did not take the credit of the DHRS, so I simply this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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159 1 ECCS system for the benefit of understanding what we 2 are dealing with. So this is very, very simple, the 3 simplest ECCS system I have ever seen from the PWR 4 design.
5 Three valves on top and then two valves --
6 the reactor recirculation valves at the bottom. And 7 you can see from the hydraulics perspective that's 8 pretty much three valves into the containment. And 9 the containment functions as part of ECCS. That's the 10 features that I think -- I think when NuScale was 11 presenting this part of the topical report, they 12 talked about this one.
13 And then one of the ACRS asked questions, 14 members asked questions. What's the -- really the big 15 change or difference? I think that's the major 16 difference. So our review actually also focus on the 17 major difference. Especially in the RELAP5 code has 18 been used so much extensively. And I think a couple 19 years ago we just gave the Subcommittee a presentation 20 on APR-1400, a large break LOCA. It's also RELAP5.
21 So let's turn to the next page. Okay. So 22 what's the scope? And the NuScale asked for the 23 approval for methodology to analyze NuScale LOCA 24 scenario. That's number one. That's primary goal.
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160 1 Revision 0 and then we added on one RAI which led to 2 the change of the appendix, addition of Appendix B to 3 provide a methodology to analyze the IORV. Okay?
4 And the methodology by itself and also the 5 testing of the -- the NIST testing. And that 6 supported the peak containment pressure analysis, so 7 the RELAP5 code plus the methodology and then the base 8 integral effect test supported the peak containment 9 pressure analysis, LOCA analysis methodology too and 10 the long-term cooling analysis. Although the approval 11 of this -- they ask us to approve is not the approval 12 for all those analysis and the methodology, but the 13 base. For example, the geometry, the base code 14 version are approved as part of this topical report.
15 That was the application scope.
16 Okay. Next slide. So with such a large 17 chunk of information coming in, so we decided to take 18 the approach to expedite the review, but at the same 19 time keep the sufficient adapt (phonetic) as much as 20 we can within the scope and the time and the resources 21 we have.
22 Okay. The initial part is the pre-23 application three years even before they submitted the 24 topical report. And we were engaged and as a part of 25 pre-application and we -- several staff -- several of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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161 1 us visited on site and we checked their NIST test 2 facility. And after we took on this topical report we 3 decided to review and they wait -- launch an initial 4 site audit. And then we have almost 10 people from us 5 and then I think from NuScale side they had almost 20.
6 So we met one -- for one week, expedited the review.
7 And then we have -- after that we had conducted 8 continuous audits until even this month. We finally 9 have a closeout of the audit to finish this one and to 10 finish SER.
11 So because of extensive audit and we 12 reduced the number of RAIs. Actually we have 13.
13 That's historically low number. But each RAI has its 14 own questions, so total of 45 issues there, the RAI 15 questions there under 13 RAIs. And through the audit 16 we resolve 210 audit issues. That's a lot of effort 17 there to get the issues resolved.
18 CO-CHAIR MARCH-LEUBA: So are you saying 19 that you will use the number of RAIs by just packing 20 the issues into -- like more issues into a single RAI, 21 or do you actually resolve the issues in the audits 22 and you have to ask the --
23 DR. LU: Well, both actually. Both.
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162 1 you're going to launch, how much merit you have to 2 launch an RAI. And so we make sure that each RAI 3 really indeed has significant safety implications 4 there. And also a lot of issues we resolve as part of 5 210 audit issues. Some of them actually evolve to be 6 an RAI. So later -- so we issue -- we said, okay, we 7 have to document this one. So that goes into the RAI.
8 CO-CHAIR MARCH-LEUBA: Yes, my experience; 9 and let's just put it on the record, is that audits 10 are very, very valuable because you can ask an 11 informal questions- and get an informal answers that 12 explain to the reviewer- where they're going. And 13 maybe you have to follow up with an RAI to document 14 it.
15 DR. LU: Yes.
16 CO-CHAIR MARCH-LEUBA: But this is a 17 documentation RAI that is just paperwork.
18 DR. LU: That's right.
19 CO-CHAIR MARCH-LEUBA: So I really would 20 like to encourage the -- like the fact that the 21 210:245 ratio on the audit to RAI is very good.
22 DR. LU: Thank you.
23 CO-CHAIR MARCH-LEUBA: I mean we should 24 even try to get it better for the next reactor.
25 DR. LU: Yes.
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163 1 MEMBER BLEY: Those audit issues aren't 2 documented in the audit reports?
3 DR. LU: Yes, we have -- actually I think 4 there were issues the two audit reports on LOCA side.
5 MEMBER BLEY: But they are in the audit 6 reports?
7 CO-CHAIR MARCH-LEUBA: But my experience 8 is often RAI --
9 MEMBER BLEY: There's probably more detail 10 in the RAI?
11 CO-CHAIR MARCH-LEUBA: Yes, often an RAI 12 comes from the staff because the staff does not 13 understand the question.
14 MEMBER BLEY: Yes. No, I agree on that.
15 CO-CHAIR MARCH-LEUBA: They don't know 16 what they're asking. So when you do it informally you 17 get answers. They say, oh, yes, you're right. I 18 don't have to issue this one. Ninety percent of the 19 issues are like that.
20 DR. SCHULTZ: Just for the public record 21 could you give some dates associated with this? You 22 say it's been going on for a long time, but dates of 23 when -- you said pre-engagement was three years before 24 the submittal and then these audits took place. Can 25 you just throw in a few annual -- year dates so we can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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164 1 get an appreciation for the time scale?
2 DR. LU: Oh, okay. For the pre-3 application I remember the first visit we had was back 4 to 2014.
5 DR. SCHULTZ: Yes.
6 DR. LU: All right? The initial on-site 7 visit was three years ago.
8 DR. SCHULTZ: Three years ago?
9 DR. LU: Three years ago, yes.
10 DR. SCHULTZ: Okay.
11 DR. LU: And then we had two phases of 12 continuing audits until actually several weeks ago.
13 DR. SCHULTZ: All right.
14 DR. LU: So the final closeout. We had it 15 at the meeting on the RAI.
16 DR. SCHULTZ: Oh, the peak of activity --
17 DR. LU: The peak is actually the first 18 year. And then -- but I'm going to talk about what's 19 the major issues. And then so the major issues 20 actually caused the -- a lot of iteration in the 21 additional testing.
22 DR. SCHULTZ: I just wanted to get on the 23 record the general timeframe that demonstrates what 24 was done.
25 DR. LU: Yes. Thank you.
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165 1 DR. SCHULTZ: Thank you.
2 DR. LU: All right. So those are the 3 issues and the number of issues. There are a number 4 of RAIs. But to really focus the staff's review --
5 and we performed our own NRELAP5 calculation, 6 sensitivity calculating confirmatory analysis and 7 TRACE. Our staff from Office of Research and -- had 8 developed this TRACE and also checked their --
9 independently checked the separate effects test and 10 also integral effect test to make sure that TRACE can 11 also do the job, too.
12 CO-CHAIR KIRCHNER: Clarification: The 13 staff did sensitivity analyses with NRELAP5 --
14 DR. LU: Yes.
15 CO-CHAIR KIRCHNER: -- or with RELAP5?
16 DR. LU: NRELAP5.
17 CO-CHAIR KIRCHNER: NRELAP5?
18 DR. LU: So we obtained the -- we normally 19 -- as part of the code topical report review we 20 normally get -- obtain the source code and then debug 21 and make sure that it was okay. Then we -- this time 22 we got the executable and we also had access to the 23 source code from their site. So we were able to dive 24 into. But we did a lot of confirmatory analysis 25 sensitivity cases. Actually Carl is the -- our expert NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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166 1 on that one.
2 MR. THURSTON: Yes, we do have RELAP5 3 also.
4 DR. LU: We also have RELAP5-3D. We have 5 3D, right.
6 (Simultaneous speaking.)
7 MR. THURSTON: -- some cases with RELAP5.
8 DR. LU: Yes.
9 MR. THURSTON: That's right.
10 DR. SCHULTZ: Carl or Shanlai, with the 11 TRACE analysis did you also do sensitivity analyses?
12 DR. LU: Yes, we did.
13 DR. SCHULTZ: Matching up what NuScale had 14 done?
15 DR. LU: Right.
16 MR. THURSTON: Yes, that's what we --
17 (Simultaneous speaking.)
18 DR. LU: I think that will be the --
19 DR. SCHULTZ: Good. Go ahead.
20 MR. THURSTON: No, I'm done.
21 DR. LU: Oh, you're done? Okay.
22 So we did. And then we -- especially when 23 we performed the NRELAP5 sensitivity calculations --
24 and then we want to make sure it was okay. So we gave 25 our research staff and say -- actually Dr. Peter Lien.
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167 1 And he's going to present a lot of those comparisons 2 here including the sensitivity cases we ran initially 3 with RELAP5 and later with TRACE.
4 DR. SCHULTZ: That's good. Thank you.
5 DR. LU: Okay. All right. I'm going to 6 talk about the major issues we identified and 7 resolved. So at this point -- and then the -- for 8 LOCA Topical Report we resolved all the issues here 9 and then in later side within a certain review area.
10 Okay. Next slide. Okay. I'm going to go 11 through this very quickly because I think the 12 Committee has been very familiar with 10 CFR 50.46.
13 And the peak cladding temperature, that's 14 acceptance criteria: 2,220 degree Fahrenheit and 15 localized maximum fuel cladding oxidation, 17 percent.
16 And the whole core fuel cladding oxidation with 17 hydrogen generation, that's one percent.
18 And the green is actually for -- is 19 NuScale decided to use Appendix K. And Appendix K 20 does give a lot of -- give them a lot of leverage to 21 simplify the -- to identify the limiting case because 22 they simply have so much margin in terms of PCT.
23 Therefore, they can afford that one. That were --
24 that both simplify their calculation and also our 25 review.
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168 1 And however Appendix K, that was a 2 specific -- a lot of specific features. So through 3 the pre-application stage they ask for the exemption.
4 So not all of them were actually needed, because for 5 example, the post center of heat transfer because they 6 never uncover the core. And no flood/refill issue, 7 the pump model. There's no pump, so therefore it's --
8 becomes very simple. And then that part of process we 9 approve that part, approach right before the topical 10 report was submitted to us.
11 And so with that so much margin because 12 the core -- if you have a figure, they presented early 13 design -- you know, design certification presentation 14 and the core, they wind up at total volume of water 15 inside the RPV. It's -- this reactor is a hot -- has 16 a list of the ratio. That's means the -- much more 17 water than power. So with that margin they can afford 18 the different figure of merit, not like operating 19 fleet.
20 So maintain containment, integrate, 21 provide water back to the RPV. And so let me see.
22 Okay. Yes. So the issue here is because once you 23 have the break and have LOCA -- and the limiting 24 scenario here is the -- whether the containment can 25 maintain to have its own integrity to protect the --
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169 1 provide the cooling. So that's part of ECCS. So 2 that's the figure of merit of the containment 3 integrity.
4 And for IORV maintain the critical heat 5 flux ratio to prevent significant clad heat-up and the 6 collapsed water level above the core. This just very 7 similar to the paths of the -- like the BWR. There's 8 so much water they have. They can say, okay, as long 9 as I have the collapsed water level above the core, 10 it's fine. So that's the figure of merit, and we 11 agree with that. And that's the baseline level 12 review.
13 Okay. That's criteria.
14 CO-CHAIR KIRCHNER: Can I -- just though 15 going back to one point. Did you also look at --
16 you've got collapsed water above the top of the core.
17 Did you also look at whether they have saturation 18 conditions above the core?
19 DR. LU: Yes, actually for most of time 20 the -- it's saturated.
21 CO-CHAIR KIRCHNER: Yes. Yes.
22 DR. LU: Yes, it's saturated. There is a 23 bubble coming out. So the reason the -- actually if 24 you -- yes, that's a good point. Actually because of 25 it's saturated and then the bubble is coming out of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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170 1 the core -- and that's a two-phase level. It's way 2 above the collapsed water level.
3 CO-CHAIR KIRCHNER: Right.
4 DR. LU: So from that perspective using a 5 collapsible level, it's very conservative. And 6 because your two-phase level is way above, and then 7 it's saturated there anyway.
8 CO-CHAIR KIRCHNER: Yes.
9 DR. LU: Okay. Any other questions after 10 this one? No? All right.
11 Okay. For IORV, because they added at 12 Appendix B and then declared this IORV case as 13 anticipated, operational occurrence and AOO. So 14 therefore, it's the SAFDL, typical SAFDL criteria into 15 CHF. So they maintain the margin to critical heat 16 flux. And then as long it's greater than the CHFR 17 limit, then it should be fine. So that's not the 18 typical LOCA topical report we -- you're able to see, 19 but that's the new item they added over there. So 20 that's the figure of merit IORV.
21 Next slide.
22 CO-CHAIR MARCH-LEUBA: Go back. Can you 23 go back to the previous slide on -- and point the 24 second -- the number -- well, the second bullet. I 25 thought that will require SAFDLs to be applied to IORV NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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171 1 is because there was the possibility of return to 2 power afterwards.
3 DR. LU: Okay.
4 CO-CHAIR MARCH-LEUBA: Is it because they 5 chose to characterize it with a higher frequency or 6 the return to the power?
7 DR. LU: Okay. Let me address this 8 because that's a good question, and that's the reason 9 -- if you go back to the scope of this topical report, 10 this one only covers short-term blowdown to long-term 11 -- to the point that it have the return to -- you have 12 a sufficient return flow to maintain the core 13 coverage. So anything beyond that point -- our 14 approval, we are approving the way they are using 15 RELAP5 code, but the actual application of the 16 methodology to address return to power and then peak 17 containment pressure, it's not covered by this topical 18 report.
19 CO-CHAIR MARCH-LEUBA: Yes, but the figure 20 of merit, if you allow return to power at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> you 21 cannot damage the fuel from the first 10 minutes.
22 DR. LU: Right.
23 CO-CHAIR MARCH-LEUBA: So it changes the 24 figure of -- the acceptance criteria.
25 DR. LU: That's true. And then that's --
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172 1 I think that that's related to the long-term cooling 2 power. I think we are going to cover that. That's 3 for sure.
4 CO-CHAIR MARCH-LEUBA: But --
5 DR. LU: Yes, agree.
6 CO-CHAIR MARCH-LEUBA: So maybe it's both.
7 Is the fact that they chose to make an IORV a high-8 probability event or high-frequency event --
9 DR. LU: Yes.
10 CO-CHAIR MARCH-LEUBA: -- and also the 11 fact that we allow return to power if there is no 12 operator action forces you to have a much more strict 13 criteria for LOCA than you will have operating a 14 reactor?
15 DR. LU: That's right. If you get to that 16 scenario. Yes, agree. Yes, that's the case. Yes.
17 That's true.
18 CO-CHAIR MARCH-LEUBA: Yes.
19 DR. LU: Any other question?
20 (No response.)
21 DR. LU: All right. Next slide. Okay.
22 NuScale already presented it at the -- NuScale develop 23 actually a very good extensive PIRT because part of 24 the design in the past 15 years and then their initial 25 PIRT panel was formed back to 2008. And then they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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173 1 came up with 21 phenomena important to capture the 2 LOCA evaluation model. So it's not only -- it's -- it 3 does not cover any other event, but just focus on 4 LOCA. And the -- we are very glad that they decide to 5 select the RELAP-3D as the base code. And then from 6 that base code they develop their own NRELAP5 code to 7 model to capture this LOCA analysis.
8 And they develop quite a lot of integral 9 test specifically tailored to capture the features of 10 NuScale design. And the good news with using RELAP5-11 3D is because it has had already extensive separate 12 effects test, and plus like integral test before. So 13 that's the good news from their perspective. And from 14 PIRT perspective we can go to the next slide.
15 CO-CHAIR MARCH-LEUBA: Stay here on the 16 PIRT.
17 DR. LU: Sure.
18 CO-CHAIR MARCH-LEUBA: Back to my favorite 19 topic of boron redistribution and boron dilution, did 20 the PIRT panel identify boron redistribution as an 21 issue that's important?
22 DR. LU: No. Actually, I can answer that 23 one. As particular part we reviewed that part. The 24 focus of -- at that time back to 2008, I think even 25 initially when they -- I think Dr. Jose Reyes NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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174 1 developed this concept. Initially it was a class 2 project for his graduate students. They were thinking 3 about how to mitigate the LOCA itself. So the -- any 4 long-term cooling related to any boron dilution, it 5 was not part of the PIRT panel's original task, so 6 they did not really address that one. And --
7 CO-CHAIR MARCH-LEUBA: But if --
8 DR. LU: Yes.
9 CO-CHAIR MARCH-LEUBA: It is the PIRT 10 panel responsibility to look at the transient and see 11 if you have reach a safe stable condition. And if 12 you're in a condition in which you're diluting the 13 boron, you have not yet reach a safe stable condition 14 when they choose the stop. So just by arbitrarily 15 stopping it at one hour doesn't save them for -- they 16 should have thought of it.
17 DR. LU: Right.
18 CO-CHAIR MARCH-LEUBA: And they didn't.
19 DR. LU: I think if NuScale started again, 20 they probably can definitely ask their PIRT panel to 21 do that. And specifically for this topical report as 22 a historical -- as it is right now what we have is 23 leading up to that, unfortunately.
24 CO-CHAIR MARCH-LEUBA: And I have reviewed 25 the long-term cooling technical report which has a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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175 1 PIRT on it and does not have boron dilution on it 2 either. So --
3 MEMBER BLEY: Yes, I want to kind of 4 support what Jose is saying here. Even if you define 5 the conditions a PIRT panel is supposedly looking 6 at --
7 DR. LU: Right.
8 MEMBER BLEY: -- it's on them. It's their 9 responsibility to see if you've over-specified the 10 case and there are some important issues that go 11 beyond what they've been specifically asked to look 12 at. Otherwise, it just doesn't work.
13 CO-CHAIR MARCH-LEUBA: Yes, and going back 14 to Charlie's topic this morning, the PIRT also didn't 15 identify the internal missiles --
16 DR. LU: Oh, okay.
17 CO-CHAIR MARCH-LEUBA: -- as an initiating 18 event even though they -- through probabilistic or 19 frequency of occurrence they can be discarded, but the 20 PIRT should have identified that there is a special 21 condition of the NPM design in which -- whereas in a 22 BWR the containment is maybe 100 feet away and is five 23 feet of concrete and there are three walls and floors 24 in between the core and the containment.
25 DR. LU: Right.
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176 1 CO-CHAIR MARCH-LEUBA: Here the 2 containment is only a couple of feet away and it's a 3 straight path for the control rod and it's only yea 4 thick of steel.
5 So in my opinion the PIRT panel should 6 have looked at the whole problem of LOCA and 7 identified special characteristics of this particular 8 design that are not in the operating fleet. And I'm 9 putting this on the record just to make sure that we 10 don't make a precedent on this. For new reactors when 11 they submit it to the staff and they have a PIRT --
12 DR. LU: Right.
13 CO-CHAIR MARCH-LEUBA: -- they don't have 14 to see which parts of the operating reactor apply to 15 me. But you also have to look at why is my design 16 different than the operating reactor and what problems 17 can I have? So and that should be one of the lesson 18 learned for the future reviews that when you do your 19 PIRTs, you should be looking at it from that point of 20 view.
21 DR. LU: Yes. Okay. So I think I 22 understand where you guys come from. And actually 23 related to the control rod housing failure potential 24 on the design. So that's related to that part, right?
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177 1 internal missile is not an issue.
2 DR. LU: Right.
3 CO-CHAIR MARCH-LEUBA: But my gut feeling 4 is not an approved or a method for analysis, but my 5 gut feeling tells me first is very low-priority then 6 that anything will keep the containment. And if it 7 hits, it doesn't have time to accelerate.
8 DR. LU: Yes.
9 CO-CHAIR MARCH-LEUBA: So but it was not 10 analyzed.
11 DR. LU: Right.
12 CO-CHAIR MARCH-LEUBA: It was not even 13 identified.
14 DR. LU: Yes. Well, I think you're 15 correct that the PIRT does not identify this. So, 16 okay. Yes, I see the problem. So you're correct that 17 they did not and then staff did. And we actually 18 launched RAIs. So that's the reason. And we launched 19 RAIs about the specific control rod housing failure 20 because there when this particular rod ejection went 21 as part of it, we had a Regulatory Guide when this was 22 developed and we were trying to figure out what can be 23 wrong with this one if the PIRT panel did not really 24 capture this one but we captured that one.
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178 1 summary. We don't think it's a problem, the reason is 2 the following. Okay. We asked them that RAI and 3 their answer is, oh, they design everything according 4 to ASME code, so therefore it's not design-basis. And 5 we disagree. And, but is that the really problem and 6 they are not? I think Carl, our expert -- actually we 7 did a lot of analysis trying to address in case.
8 What's wrong with this wrong? It was not ignore. It 9 was not captured.
10 So assuming there is a rod ejection, there 11 is a follow-on the -- controlled housing failure, 12 which can be a low-frequency event, but with Davis-13 Besse; everybody heard about that, it's the -- because 14 of the corrosion. So it's just one cycle away from 15 rod ejection. So it's not something that much that 16 can be treated without any basis.
17 So we launch an RAI, and the RAI came back 18 then that we disagree, but we close that RAI because 19 we did our extensive -- our own analysis. Our own 20 analysis, number one, is the -- by itself because 21 controlled rod housing we have a control rod nozzle, 22 a control rod driver nozzle. And then that's the 23 particular welding point with the control rod driver 24 system. There is a welding point. That's the place 25 that is suspect -- that it's the place potentially NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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179 1 have -- may have the break or may have a failure.
2 If that pops up, the control rod will be 3 ejected like a missile, right?
4 So from a reactor system perspective we 5 did dive into that point, into that. And so the good 6 news here, the nozzle diameter is less than the RVV, 7 the throat area. So they have a bounding case from 8 thermal hydraulics perspective, bound to this LOCA 9 scenario.
10 And then Carl did additional -- quite a 11 lot of sensitivity cases trying to see if we extended 12 the rod -- because if you have a rod ejection, you 13 don't really right away shut down, you have a power 14 spike. So we have a simulated power spike based on a 15 rod ejection calculation assuming there's no void 16 effect, which is very conservative. And then we add 17 that to the core. Do we have a problem with peak 18 containment pressure? We don't. So we close this RAI 19 because we agree with them that it's not -- should not 20 be a design-basis issue. It's because we don't 21 believe that the failure of that particular control 22 rod housing where you resulting in more damaging event 23 than the limiting case they already analyzed from a 24 reactor assessment perspective.
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180 1 ejection and the missile, in fact that's the position 2 taken by the staff from Chapter 3. So that's the part 3 -- probably if you guys have a problem, you may want 4 to talk to Chapter 3 guys. But unfortunately I agree.
5 CO-CHAIR MARCH-LEUBA: I hate when you 6 guys say that.
7 (Laughter.)
8 CO-CHAIR MARCH-LEUBA: Okay? Because when 9 you say it's not my problem, then I'm here, so who did 10 it? Okay? So try not to say that.
11 DR. LU: We will pass along that -- your 12 comments.
13 MEMBER RICCARDELLA: I went back after the 14 discussion this morning and looked at the Chapter 3, 15 both the DCA and the SER for Chapter 3, and the SER 16 for this particular 3.5.1.1, it goes into a 17 probabilistic argument. It says if the combined 18 probability of the nozzle rupture -- it's the 19 probability of hitting a safety-related structure and 20 it -- probability of damaging the safety-related 21 structure. So it's all less than 10 to the minus 7th 22 and it's considered acceptable.
23 And so I don't know if anybody's done 24 those risk -- probabilistic calculations, but that was 25 the basis. It's not the basis that, oh, it's part of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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181 1 a vessel, therefore it can't fail. So I'm okay with 2 that consideration.
3 MEMBER BROWN: Did it address the 4 containment or just the reactor pressure vessel?
5 MEMBER RICCARDELLA: Both. It's a 6 combination of three probabilities they say. If the 7 probability of the rupture is less than 10 to the 8 minus 7th, it's okay. You don't have to consider it.
9 If the rupture --
10 MEMBER BROWN: So it's not going to become 11 a missile?
12 MEMBER RICCARDELLA: Well, the probability 13 of -- yes.
14 MEMBER BLEY: It's less likely than that.
15 MEMBER RICCARDELLA: If it's less likely 16 than that. If it's -- but if the probability of that 17 plus the probability of it hitting the safety-related 18 structure --
19 MEMBER BLEY: The containment.
20 MEMBER RICCARDELLA: -- the combined 21 probability is less than 10 to the minus 7th, it's 22 okay.
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182 1 to the minus 7th. So I think P1 times P3 is probably 2 less than 10 to the minus 7th, but it's -- that was 3 the basis for the staff acceptance of that Chapter --
4 MEMBER DIMITRIJEVIC: So for all three 5 probabilities the same application, same limit 6 applies, 10 -- less than 10 to the minus 7th?
7 MEMBER RICCARDELLA: Ten to the minus 8 seventh, yes.
9 MEMBER DIMITRIJEVIC: Well, what --
10 MEMBER RICCARDELLA: And I guess that's 11 the same as what's currently applied to operating --
12 I mean, that's out of a SRP.
13 MEMBER DIMITRIJEVIC: Well, it's 14 currently, but would that obviously with, you know, a 15 large release frequency like 10 to the minus 11 it 16 doesn't make sense to apply this criteria.
17 MEMBER RICCARDELLA: Right. I think --
18 MEMBER DIMITRIJEVIC: Because the LOCA, 19 the charging LOCA, I check, currently, which results 20 in containment bypass dominates large release, but is 21 right currently one minus nine challenging frequency.
22 So it doesn't -- if this is a real event which can 23 cause the bypass of containment in small LOCA, I mean 24 it would be much -- 10 to minus 7 is a hundred times 25 higher than 10 to the minus 9.
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183 1 CO-CHAIR MARCH-LEUBA: So maybe we should 2 have that discussion with the Chapter 19 guys when 3 they're here next month.
4 MEMBER DIMITRIJEVIC: No, the Chapter 19 5 -- yes, you have a specific picture with Chapter 19 6 that's takes care, but everything you guys miss in 7 Chapter 15, and that's now it --
8 CO-CHAIR MARCH-LEUBA: No, Chapter 15 9 tells you if the control rod breaks and hits the 10 containment, you get a LOCA, but it bypasses 11 containment. And now you have to tell me what the 12 frequency of that is. And that comes from 19. Well, 13 what we're thinking is that would be bad. If you have 14 a rod ejection --
15 MEMBER DIMITRIJEVIC: Well, why does it go 16 back to this 10 to minus 7 criteria in Reg -- because 17 we discuss that unfortunately when we are discussing 18 this. We have a criteria in regulation which doesn't 19 apply for these low-frequency things. And this is the 20 big thing. Because we saw this in many cases. This 21 is not the first case we see this. You cannot apply 22 the same criteria when you have a completely different 23 risk measures. We saw the selection of the scenarios 24 for the source terms and we are seeing it again here, 25 that this is not applicable. So somebody should be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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184 1 taking care how does this apply. And I don't think 2 that should be NuScale.
3 MS. PATTON: I just need to correct 4 something that was said earlier. It's not only in 5 Chapter 3 where we say that this isn't considered.
6 It's not in the design-basis. It's also in 15.
7 So what the LOCA Team is saying is correct 8 in terms of we ran a bunch of sensitivities and all 9 that to look at it. That was when we were trying to 10 determine at that time whether we needed to consider 11 it within the design-basis.
12 However, that decision is documented in 13 1506 in the return to power section, one of the 14 references I gave you earlier, because it was also 15 being looked at as a potential return to power 16 scenario. So we did determined that that was not 17 within the design-basis. Although we do have a lot of 18 sensitivities, that's solely from a LOCA perspective.
19 It would have been all right, but it's not considered 20 even within 15. It's in the design-basis.
21 CO-CHAIR MARCH-LEUBA: When you're doing 22 1506 and doing that evaluation, how do you determine 23 what the frequency of the event is? So it's not a 24 design-basis event.
25 MS. PATTON: Right, but that's -- the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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185 1 basis for that, it was like a regulatory basis 2 basically that -- and one of them was the -- part of 3 it was based on the Chapter 3 evaluation.
4 CO-CHAIR MARCH-LEUBA: Yes.
5 MS. PATTON: Also looking back at the 6 GDCs, the historical implementation of those, right?
7 And the rod ejection is considered a short-term type 8 of event.
9 CO-CHAIR MARCH-LEUBA: But it is because 10 in operating reactors the containment is a mile away 11 and there are three floors of concrete in between it 12 and the containment and the containment is five feet 13 of concrete.
14 DR. CORRADINI: No, that's not true for 15 all containments. There's a lot of free-standing 16 steel shell containments that are out there and 17 operating. So that's not true.
18 MS. PATTON: We actually researched the 19 entire regulatory history of the whole thing and that 20 whole event with the attorneys and looked at this in 21 a good amount of detail and determined that it was not 22 going to fall within the design-basis within the 23 regulations.
24 CO-CHAIR MARCH-LEUBA: It would have been 25 so much more satisfying to have made a simple missile NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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186 1 calculation showing that it cannot break the 2 containment, which again -- I mean, there is clearly 3 -- the rod doesn't have time to accelerate. But, 4 okay.
5 DR. LU: Okay. So next slide.
6 All right. Now let me talk about the 7 significant issues we identified and we resolved.
8 Just give a heads up. Resolved.
9 But the first, No. 1, that when they 10 initially submitted this one they did not have a 11 flange. So they just welded the valves with the RPV.
12 So later they change it to the flange. So between the 13 valve and then RPV there is a flange. So possibly 14 there -- has that been considered as part of a PIRT, 15 considered potentially that either the nozzle welding 16 point or the flange may become a break? And I think 17 that part has been resolved. They declared this one 18 as exclusion zone in the RPV. The unit containment 19 they -- outside of RPV of course. And then so 20 therefore there would be no more break location 21 considered as a design-basis for this one. And 22 Chapter 3 reviewer agreed with their special treatment 23 to declare this the break exclusion zone.
24 And so that's the part resolved. We 25 identified this issue as part of a PIRT and then our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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187 1 expert from previous Areva retired, Bert Dunn. And 2 also from NuMark, they identified this issue very good 3 issue. And then so we resolved that this point and 4 the Chapter 3 reviewer accepted this claim that they 5 are excluded from any potential break assumption 6 location, break location assumption.
7 CO-CHAIR KIRCHNER: So, Shanlai, just to 8 summarize. So you're saying -- if I remember 9 correctly, this is a bolted flange for the RRV.
10 DR. LU: It's a bolted flange. Yes, 11 that's right.
12 CO-CHAIR KIRCHNER: So you're saying you 13 can't have a break at the flange essentially?
14 DR. LU: NuScale says. And then the 15 Chapter 3 reviewer decided to approve that at this 16 point.
17 CO-CHAIR KIRCHNER: Okay. So therefore 18 that limits --
19 (Simultaneous speaking.)
20 MEMBER RICCARDELLA: -- you decided review 21 this --
22 DR. LU: To approve that claim, to have 23 the exclusion from --
24 CO-CHAIR KIRCHNER: But you cannot --
25 DR. LU: Yes.
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188 1 CO-CHAIR KIRCHNER: Okay. So you can't 2 have the valve body break loose, essentially is what 3 you're saying.
4 DR. LU: Right.
5 CO-CHAIR KIRCHNER: So therefore the RRV 6 is equivalent LOCA. I know the applicant doesn't want 7 to call it a LOCA, but it is a LOCA. If the RRV 8 opens, then you're just limiting to the actual valve 9 diameter, right?
10 DR. LU: Right.
11 CO-CHAIR KIRCHNER: Okay.
12 DR. LU: And another reason is I think we 13 are -- from our reactor system perspective we are 14 supporting that; I personally also support that point, 15 is because we did -- actually our staff handed that to 16 Carl and then the also Office of Research did 17 additional analysis. Just use the maximum possible 18 location. Let's assume it's flange also fail.
19 CO-CHAIR KIRCHNER: Yes.
20 DR. LU: What's going to happen? And it 21 doesn't really hurt. It does not really hurt the peak 22 containment pressure nor the clad 4 level above the 23 core, so figure of merit is covered. So therefore, 24 with additional measure NuScale decided to maintain 25 the flange and check that one during each cycle. And NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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189 1 I think with it or without it it's still okay because 2 by design they have sufficient water there. And they 3 also -- the nozzle size not that large, and so then 4 that will not challenge the peak containment pressure.
5 So I -- from reactor system perspective we supported 6 that decision. So that issue was resolved from that 7 perspective.
8 CO-CHAIR KIRCHNER: Did you actually do 9 TRACE or other analysis for --
10 DR. LU: Yes, we did.
11 CO-CHAIR KIRCHNER: So you did it with a 12 nozzle probably with a different --
13 DR. LU: Different size.
14 CO-CHAIR KIRCHNER: -- either Moody or 15 whatever factors you used?
16 DR. LU: Right.
17 CO-CHAIR KIRCHNER: One for the actual 18 valve and one for a nozzle?
19 DR. LU: Right.
20 CO-CHAIR KIRCHNER: Okay.
21 DR. LU: That's right. That's right. So 22 the nozzle size -- of course the throat size is much 23 smaller than a nozzle size. The nozzle size -- when 24 we enlarge it to the nozzle size, then it's not that 25 bad. And actually it's a -- there is an increase NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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190 1 there of the peak containment pressure, and there's no 2 really impact for the -- on the clad 4 level. So 3 therefore, we don't feel that it's not -- it's not as 4 much of challenge there for us to prove that.
5 CO-CHAIR MARCH-LEUBA: Yes, and that 6 conclusion is very intuitive because the peak pressure 7 happens with the steam valves when the RVV is open, 8 not when the RRV is open.
9 CO-CHAIR KIRCHNER: No, no. No, no.
10 PARTICIPANT: Peak pressure is --
11 CO-CHAIR KIRCHNER: No, actually it's the 12 opposite.
13 (Simultaneous speaking.)
14 PARTICIPANT: -- is the RRV.
15 CO-CHAIR KIRCHNER: Yes.
16 CO-CHAIR MARCH-LEUBA: I don't --
17 CO-CHAIR KIRCHNER: It's the limiting 18 case.
19 CO-CHAIR MARCH-LEUBA: No --
20 CO-CHAIR KIRCHNER: That's why they're 21 addressing it.
22 CO-CHAIR MARCH-LEUBA: No, the --
23 CO-CHAIR KIRCHNER: This is a liquid 24 discharge, two-phase. This is the worst break for 25 them, of the different valve system.
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191 1 (Simultaneous speaking.)
2 PARTICIPANT: Worse than a steam --
3 CO-CHAIR KIRCHNER: A burst in the steam 4 break on the top.
5 CO-CHAIR MARCH-LEUBA: No, it is an RRV 6 break that dumps liquid into containment. And when 7 you hit the ECCS saturation level, then it opens the 8 RVVs and that causes a big break.
9 CO-CHAIR KIRCHNER: No, when your primary 10 system blows down from full pressure and temperature, 11 this is the limiting case. Trust me. Not a steam 12 line break.
13 DR. LU: Yes, I think we are going to --
14 (Simultaneous speaking.)
15 DR. LU: Back to March, the staff is going 16 to talk about that one, so that was the details. But 17 the bottom line, I think this particular issue is 18 resolved. We identified. We launched the RAI. And 19 then very few RAIs. We -- this one of them. And then 20 the next one is also a sensitive one. And as the 21 Chairman mentioned about this one, the NuScale did not 22 want to declare as a LOCA, but it's a lower point of 23 the RPV can discharge coolant to the containment.
24 CO-CHAIR KIRCHNER: Yes.
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192 1 and think our consultant raise the question: How come 2 the lowest possible, largest potential discharging 3 point? We don't do this to say the breakpoint. Why 4 there was no test was not done? And is the PIRT 5 really covering this one?
6 So that took us a while and -- almost more 7 than year to talk to the NuScale. And then they 8 finally responded to this and with additional test.
9 And they did the RRV test and then discharge this one.
10 And the good news here turn out to be, yes, so 11 although it's a different number, came up with 12 different number, different peak containment pressure 13 number, different clad 4 level, but it's still within 14 the acceptable limit. And therefore there is no 15 challenge to the existing claim. And then -- or from 16 our perspective, from staff's perspective there's no 17 integrity issue, nor the core cover issue.
18 So therefore, this particular issue was 19 resolved. But that's part of a PIRT review 20 identifying this issue. How come never thought about 21 this one? The lowest level. There was no test.
22 CO-CHAIR KIRCHNER: Well, just one 23 member's opinion, to avoid using the word LOCA, 24 because their design is such that they don't have the 25 extended piping system of the commercial fleet --
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193 1 DR. LU: That's right.
2 CO-CHAIR KIRCHNER: -- so now we are 3 calling this -- and here I think in one of our write-4 ups that you will receive on the chapters there's 5 inconsistency because they've called it an AOO. By 6 definition an AOO is a frequent event, something that 7 happens one or more times in the lifetime of a plant.
8 And that's in Chapter 6 of your SER.
9 DR. LU: Right.
10 CO-CHAIR KIRCHNER: So there's some 11 inconsistent terminology that -- between Chapter 1.6 12 and 15.
13 DR. LU: I understand your comment.
14 Actually I personally understand it, but I think the 15 bottom line from a staff perspective if they are 16 applying AOO criteria of CHF or SAFDL and then they 17 definitely can meet 50.46.
18 CO-CHAIR KIRCHNER: Yes.
19 DR. LU: So what's the problem? There is 20 no problem there, so we let it go. That's how we 21 decide, made the decision. Say, okay, this issue is 22 resolved. With additional testing saying that there's 23 no additional new phenomena or change the PIRT. So 24 therefore, this particular issue was closed. But I 25 agree with you that's the -- there are a lot of places NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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194 1 we are having -- we have to learn how to jump out of 2 the historic thinking of what we have been doing with 3 operating fleet and then to accommodate new design 4 features. I think that's a part. There are some of 5 the regulatory inconsistency or consistency we have to 6 develop to address design features.
7 CO-CHAIR KIRCHNER: So one question that 8 occurs to me is once you admit to the RRV being an 9 AOO, is the other equipment in containment protected 10 against the blast that's going to come out of this 11 valve?
12 Where I'm going with this is that, for the 13 conventional valves up on the top of the reactor 14 vessel, there are usually shrouds or other things to 15 prevent the impingement zone affecting cables.
16 DR. LU: Right. Right.
17 CO-CHAIR KIRCHNER: We've got the lines 18 that come in to actuate the ECCS valves and other 19 things inside containment.
20 DR. LU: Right.
21 CO-CHAIR KIRCHNER: If this RRV lets go, 22 is there the possibility that that two-phase mixture 23 coming out could impact other equipment inside 24 containment?
25 DR. LU: That's a very good question.
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195 1 MR. THURSTON: Yes.
2 DR. LU: Actually I think we are going --
3 MR. THURSTON: So the valve design for the 4 RRV is a downward projection.
5 DR. LU: Yes.
6 MR. THURSTON: So the nozzle is projected 7 in the downward inside the containment. And in the 8 RRVs the valves at the top, they have some type of 9 device on the outlet of the RVVs.
10 DR. LU: Yes.
11 MR. THURSTON: So they have accounted for 12 those phenomena.
13 DR. LU: Yes, and that --
14 DR. CORRADINI: But I think what Walt's 15 question is even though it's pointed downward within 16 that annulus, is there --
17 CO-CHAIR KIRCHNER: Well, there's the --
18 DR. CORRADINI: -- equipment or cabling --
19 CO-CHAIR KIRCHNER: The level detection 20 sensors is the first things that occurs to me. The 21 second thing is all the lines that go to the ECCS 22 valves, et cetera. So there's --
23 DR. LU: Yes.
24 CO-CHAIR KIRCHNER: It's not as clean a 25 containment as we're lead to believe. There is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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196 1 equipment inside that annulus that could be impacted.
2 The first one that I would worry about would be the 3 level sensor.
4 DR. LU: Right. Actually I think the --
5 as Carl mentioned, is designed to be -- have a 6 different orientation of discharge. There was -- also 7 device was designed to protect the equipment against 8 that impingement. And that device review was done as 9 part of Chapter 3. We only heard of the -- but they 10 -- I think the final disposition of that one was done 11 by Chapter 3 guys. My understanding is that issue was 12 raised and it was also resolved.
13 CO-CHAIR KIRCHNER: I know it was for 14 the --
15 DR. LU: Right.
16 CO-CHAIR KIRCHNER: -- because you have to 17 worry about it for the reactor safety valve.
18 DR. LU: Right.
19 CO-CHAIR KIRCHNER: And also the reactor 20 vent valves. And I think there were shrouds or other 21 measures taken to reduce impingement on equipment 22 that's inside containment.
23 DR. LU: That's right.
24 CO-CHAIR KIRCHNER: But the RRV, the 25 potential of that opening, yes, it's oriented down NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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197 1 because its real function was to bring water into the 2 reactor vessel.
3 DR. LU: Right, right, right.
4 CO-CHAIR KIRCHNER: Hence, it's oriented 5 down, a little bit different than you would expect.
6 DR. LU: Right.
7 CO-CHAIR KIRCHNER: But to my knowledge 8 from reviewing Chapter 6, I didn't see any shrouds or 9 anything to prevent a two-phase jet impingement on 10 anything down in the containment.
11 DR. LU: Yes, that was --
12 CO-CHAIR KIRCHNER: So my concern would be 13 the level detector.
14 DR. LU: Okay. Yes.
15 MEMBER BLEY: Well, the --
16 DR. LU: Go ahead.
17 MEMBER BLEY: Remember the level detector 18 is expected to be a radar detector --
19 DR. LU: yes.
20 MEMBER BLEY: -- from the top down. Now 21 I don't know -- it wasn't specified --
22 DR. LU: Right.
23 MEMBER BLEY: -- at the time they did that 24 analysis, near as I heard.
25 DR. LU: That's right. That's right.
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198 1 MEMBER BLEY: So it probably didn't get 2 looked at at that time.
3 CO-CHAIR KIRCHNER: Yes. So I just raised 4 that as an observation.
5 DR. LU: Yes.
6 CO-CHAIR KIRCHNER: That's fine.
7 MS. PATTON: So that's done in Chapter 3 8 as part of the hydrodynamic loads evaluation. So I'm 9 trying to -- it's in their SER. It should be for 10 Chapter 3 where it talks about jet impingement and the 11 loadings and the review for that. But that's where 12 you would find it is in that chapter.
13 CO-CHAIR KIRCHNER: I know it's there for 14 the valves that are on the top of the reactor vessel, 15 but I don't know that it was considered for the RRV.
16 See where I'm going with this?
17 MS. PATTON: Yes, I mean, I'm trying --
18 CO-CHAIR KIRCHNER: I just make it as a 19 note.
20 MS. PATTON: It should be in that section.
21 DR. LU: Okay. We'll take back your 22 comment.
23 CO-CHAIR KIRCHNER: Yes, the upstairs 24 valves, they're shrouded.
25 DR. LU: Okay.
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199 1 CO-CHAIR KIRCHNER: I already looked at 2 that.
3 MS. PATTON: Right, but either rate that's 4 part of the DCA review.
5 DR. LU: Right. Yes, it's in our topical 6 report. Yes, indeed.
7 Next slide. That's a PIRT. Sorry. And 8 the models, the -- we are very glad that NuScale 9 selected RELAP5-3D computer come from INL, the base 10 code. And they developed the specific component to 11 model the NuScale reactor. And the one along the top 12 is a helical coil steam generator. So they have a 13 specific NRELAP5 component. They have pipe component, 14 they have a branch component, a whole bunch of 15 different specific hydraulic components, but they have 16 the specific steam generator component to model the 17 heat transfer and the pressure drops.
18 And then we have been talk about what the 19 -- our findings are and then also how -- what's after 20 it in detail as part of a non-LOCA topical report 21 presentation.
22 And they also develop their own and added, 23 selected the right -- the CHF correlation, and 24 specifically for the fuel selected there. And 25 obviously since containment functions as one part of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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200 1 ECCS system and provided the cooling, so therefore 2 steam condensation is a unique feature of this device 3 from a code and modeling perspective. So therefore, 4 they spent a lot of effort to identify and to make 5 sure that the correct correlation is used and also 6 benchmarked against their test facility. So that's 7 the part. And we can talk about details in -- during 8 the proprietary session.
9 In the interfacial draft for large 10 diameter pipes, because of the riser above the core, 11 basically it's a large -- huge, large-diameter pipe, 12 excuse me. So therefore, the interfacial draft 13 becomes important there. So that's part of the model.
14 They had a unique adjustment there.
15 Okay. And we also -- staff also review 16 their NPM model, NuScale power module, the action 17 model from the actual access model to the point to 18 change it to the nodal RELAP5 model. So every step --
19 and then we audit it and we did the calculation. We 20 identified issues and we resolved. So what we are 21 proving here is they really -- the final -- the NPM 22 model based on our review and the resolution of the 23 comments we had.
24 CO-CHAIR KIRCHNER: Shanlai, I'd just like 25 to make an observation.
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201 1 DR. LU: Yes.
2 CO-CHAIR KIRCHNER: You have brackets 3 after stem condensation and critical flow models.
4 DR. LU: Right.
5 CO-CHAIR KIRCHNER: We'll get to it in 6 closed session, but --
7 DR. LU: Right.
8 CO-CHAIR KIRCHNER: -- most of these 9 models are in the public domain.
10 DR. LU: That's right. The --
11 (Simultaneous speaking.)
12 CO-CHAIR KIRCHNER: -- the public domain.
13 DR. LU: Yes, you're right. And that --
14 CO-CHAIR KIRCHNER: So I just personally 15 object to the idea that that somehow becomes 16 proprietary, which is different than CHF where they 17 actually developed the CHF correlations.
18 DR. LU: Right.
19 CO-CHAIR KIRCHNER: But in general, like 20 the break discharge models, those are right out of the 21 public domain.
22 DR. LU: That's right.
23 CO-CHAIR KIRCHNER: That's an observation.
24 DR. LU: In Appendix K you see they are --
25 (Simultaneous speaking.)
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202 1 CO-CHAIR KIRCHNER: Or Appendix A. Yes.
2 DR. LU: There is a specific critical flow 3 break discharge. I think that's just a proprietary 4 check we would like to have, so I can -- I respect 5 that decision. So we will talk about in details --
6 CO-CHAIR MARCH-LEUBA: In defense of the 7 applicant, which I don't do often, they spent a lot of 8 money verifying that those open domain correlations 9 work for their system.
10 DR. LU: That's right.
11 CO-CHAIR MARCH-LEUBA: And the fact that 12 this particular correlation work for my system, I 13 spend a lot of money doing it. And they have the 14 right to keep it.
15 DR. LU: That's right.
16 CO-CHAIR KIRCHNER: It's an expensive 17 choice.
18 CO-CHAIR MARCH-LEUBA: Yes.
19 DR. LU: The choice is they decided to use 20 public domain. It's still --
21 PARTICIPANT: That one I'm confident is 22 all public domain, but --
23 DR. LU: That's right.
24 (Off-microphone comments.)
25 DR. LU: Then you can see that point, yes.
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203 1 Indeed.
2 Okay. Next slide. So we have done 3 extensive testing into -- with a scaled -- I think a 4 partial height, not full height, in one-third height 5 test facility. And then this one is -- as NuScale 6 already presented at their NIST facility, covers 7 reactor pressure vessel, steam generator and DHRS, 8 containment and the containment pool. And so we have 9 all the valves and vessels, whatever the NPM module 10 has. So what is the -- really the extensive review by 11 Dr. Peter Lien and also Upendra from Brookhaven 12 National Lab, who was on the scanning analysis.
13 So this one-third height and devices was 14 -- can that be scaled back to NPM and do we have any 15 distortions or do they have identified sufficient 16 detail to justify the distortions that are acceptable?
17 Those are the issues we dived into.
18 And because we dived into this one and as 19 part of scaling and the distortion analysis review, 20 and I think the team identified one issue, which was 21 also -- at that time it was a significant issue we had 22 to report to the office director. And by the time we 23 hit this issue we found, oops, all the previous LOCA 24 test, scaled LOCA test was bound with only half of the 25 decay heat power. It was supposed to do with a scaled NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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204 1 to this much of power. All the previous test was done 2 half the power. And then that was a shock. And then 3 so we issued RAI and then we pushed the -- it took a 4 long time and then also to push that for another test.
5 Most limiting LOCA test with full decay heat power, 6 scale decay heat power.
7 So staff -- I think our team, we did our 8 job and then we identify this issue. And then the 9 good news is the conceptual design by itself has so 10 much water. You add 100 percent additional power 11 level into the core. And then for this NIST facility, 12 they say okay. Now all other designs, all other PWR 13 designs you have this kind of condition. You raise 14 the power, decay heat power by a factor of two.
15 Pretty much none of the power can now make it, but 16 they made it. So that's good to see that. And that's 17 issue resolved. And then resulting additional tests.
18 That's the reason why it's three years.
19 And I think early the staff -- you know, ACRS was 20 asking why took that long? Because there are 21 additional tests being -- that were done.
22 DR. CORRADINI: Where is that documented 23 in the SER? I can't find that.
24 DR. LU: I think that we --
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205 1 find in the SER.
2 DR. LU: Okay. The SER -- okay. The --
3 what --
4 DR. CORRADINI: Because the scaling 5 portion of the SER is -- the key scaling parameter is 6 on keeping time in variant at the various scales.
7 DR. LU: Right.
8 DR. CORRADINI: But this is a power 9 mistake. I don't remember reading that.
10 DR. LU: Yes. Well, this -- the staff was 11 working on that. The final SER was based on all the 12 revised version of the topical report, Revision 1.
13 And there's a final version of the NRELAP5. So we are 14 approving although the goal is still the same as the 15 Rev. 0. But by the time we reach Rev. 1, those are 16 the -- so when we wrote our SER --
17 DR. CORRADINI: There's no historical 18 reference to this observation.
19 DR. LU: Yes. Yes. Yes.
20 DR. CORRADINI: Okay.
21 DR. LU: So we're --
22 DR. SCHULTZ: Isn't there --
23 DR. LU: Yes, go ahead.
24 DR. SCHULTZ: Isn't there an RA response 25 which documents the results of the test as compared to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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206 1 what had been done previously?
2 DR. LU: Yes, we actually have a --
3 specifically it's the middle of the -- I think that 4 was part of HP-06b about this one. So this was --
5 that's additional test.
6 DR. SCHULTZ: Can you get the RAI response 7 for us?
8 DR. LU: We'll get back to you. And then 9 either RAI responses or audit report, because we had 10 that -- this issue.
11 DR. SCHULTZ: Well, not just identifying 12 the issue, but the results of the test --
13 DR. LU: Okay. Results of test, yes.
14 DR. SCHULTZ: -- in comparison to the 15 original half-power test.
16 DR. LU: Yes, actually the -- yes, we'll 17 get back to you about the ML number or whatever 18 information. And actually part of --
19 MS. PATTON: Just to explain a little bit 20 more how we're doing it. So the office undertook and 21 effort to change a little bit how we're writing the 22 safety evaluations. So now it's -- right now is like 23 the latest revision. Like you said, when you're 24 Revision 1, you write the SE against that revision.
25 So issues that were brought up and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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207 1 resolved are documented through audit reports, through 2 the RAI process and the issuance of those. But it's 3 whatever that final rev -- and so if that final rev 4 resolves the issue, that's what you're writing the SE 5 against. But we're presenting some of the issues that 6 have been encountered during the review so that ACRS 7 could see some of those major issues, because we 8 understood that that was part of what you wanted to 9 see.
10 MEMBER BLEY: I admit to being a little 11 confused because other -- and maybe it's different 12 section by section and the way you treat the SERs. I 13 think for the chapters -- the NuScale design cert, 14 we've been told the SERs include the revision they're 15 written on plus all of the resolved RAIs up to that 16 point, which it really looks like the next rev of the 17 design cert.
18 MS. PATTON: So the way that we write SEs 19 evolves throughout the review. And so I can't speak 20 to some of the other chapters as to how much they went 21 towards the office director's goal of writing SEs this 22 way. Some of those chapters I think they may have 23 been further along in the writing at the time that 24 decision was made and it may have been written more 25 the old way. But the new way that the office is going NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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208 1 is you write it against whatever the current rev is.
2 So in other words, if you had written RAIs 3 that said you guys need to specify X, Y and Z in this 4 document. And then later they end up specifying X, Y 5 and Z. When you write the SER, you're writing it 6 against the revision that does have that 7 specification. So you don't go through all the 8 history of we had a problem with this. We made them 9 fix it. They fixed it. Here's what they did.
10 MEMBER BLEY: Okay.
11 MS. PATTON: That's historical. But we do 12 have enough for the legal basis of why everything is 13 okay as it's currently written.
14 DR. CORRADINI: At least in this one I 15 guess this is a big enough difference. At least I'd 16 like to know what the RAI number is so we can --
17 MS. PATTON: Yes, we can get you that.
18 DR. SCHULTZ: No, and the reason is that 19 there was a spectrum of tests that were performed with 20 the half power.
21 DR. LU: Right.
22 DR. SCHULTZ: And a limiting case 23 performed with the full power.
24 DR. LU: Right.
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209 1 to know. You said it was good to go, but the 2 technical --
3 DR. LU: Yes. Want to know the technical 4 basis.
5 DR. SCHULTZ: -- basis behind that would 6 be good to have.
7 DR. LU: Yes.
8 DR. SCHULTZ: Thank you.
9 DR. LU: That's fair. And actually I 10 think the slides we are going to present you today 11 probably has already had some of those in comparison 12 there. So that can may be --
13 (Simultaneous speaking.)
14 DR. SCHULTZ: That will be helpful also.
15 DR. LU: Right. Yes.
16 DR. SCHULTZ: Thank you.
17 DR. LU: Okay. All right. Let's get to 18 the confirmatory analysis. As usual that for a new 19 design, and especially for light water reactor design, 20 we perform extensive confirmatory analysis to just to 21 make sure everything is okay. And unlike AP1000 and 22 these PWR, which we had develop our own test facility, 23 this one we did not. So we put our effort on the 24 confirmatory analysis with the great support from 25 Office of Research.
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210 1 And so we went through their separate 2 effect tests. We set it to the NuScale power module, 3 KAIST model relate to DHRS tube condensation. It's 4 almost -- I think I recall it's full scale of the 5 tubes there. And the SIET test model. And then they 6 had their -- over the test facility Italy. And then 7 we run our code and also develop our TRACE model. And 8 we identify the issues in them. Currently is resolved 9 at this point.
10 And also the NIST-1 model. By itself the 11 NIST facility, we model that one, so that we can -- be 12 make sure that how much uncertainty they are claiming 13 is really agreeable with our own independent analysis, 14 which is a part of our confirmatory analysis effort.
15 The integral effects test, and then as I 16 mentioned, they did a LOCA/non-LOCA test. And so 17 therefore we used a TRACE code and NRELAP5 code to 18 understand their integral effect test and then make 19 sure that our code also has the capability to have a 20 same accuracy, or similar accuracy, or even better to 21 assess their NRELAP5 model. And I think very good 22 agreement were obtained at the end, so we were happy 23 with that one. So that's how we end up saying, okay, 24 this model are good.
25 And of course we used both TRACE and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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211 1 NRELAP5 code. And we had -- Carl ran many, many 2 cases, 55 sets of calculations with sensitivity cases 3 of different separated factors or integral factors and 4 NPM models. There are a lot there.
5 And then because of that effort we 6 identify a lot of audit issues and those are result by 7 resolving those all the issues. Plus we had a QA 8 audit. And then they decided to upgrade the code 9 version from 1.3 to 1.4. And then good agreements 10 were obtained between their code prediction against 11 the test results. So therefore, we feel comfortable 12 to prove this topical.
13 All right. Keep going.
14 DR. SCHULTZ: Shanlai, for the open 15 session, can you categorize those items that you've 16 identified in terms of what were the important items 17 that you identified that --
18 DR. LU: Those are -- yes, we --
19 (Simultaneous speaking.)
20 DR. SCHULTZ: -- code modifications and so 21 forth?
22 DR. LU: Yes, for example, we already 23 mention about the RRV location. The RRV, the nozzle 24 size versus the throat size. We also are trying to 25 understand the impact on the peak containment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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212 1 pressure. There are a lot of stuff can impact the 2 peak containment pressure. And then if we -- I also 3 mention about that we were diving into whether that 4 rod ejection went with -- assuming it's a rod in 5 controlled housing failure --
6 DR. SCHULTZ: Yes.
7 DR. LU: -- what would be the impact on 8 peak power? Those are part of 54. We have our 9 excellent expert on NRELAP5. He ran all those cases.
10 And then, so we also have the Office of Research 11 running the same things as we are, too, and with 12 different subset, with a focus on -- we normally ran 13 NRELAP5 with the initial runs and then we stay focused 14 with TRACE focusing on specifically do we have a 15 figure of merit and -- of the issue? And then Dr.
16 Weidong Wang and Dr. Peter Lien, those are the key 17 contributors of that.
18 DR. SCHULTZ: And the good agreements were 19 obtained on Version 1.4?
20 DR. LU: Yes, 1.4 --
21 (Simultaneous speaking.)
22 DR. LU: Yes, the 1.4. 1.4 and then some 23 of the 1.3 version for specific tests. Yes, 24 normally --
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213 1 difference. There are minor changes and corrections 2 that they made to Version 1.4. So it's -- I would 3 characterize it --
4 DR. SCHULTZ: There were some changes 5 there that you --
6 (Simultaneous speaking.)
7 MR. THURSTON: Yes, and we will do those 8 in the closed session.
9 DR. SCHULTZ: Okay. safety 10 DR. LU: Yes.
11 DR. SCHULTZ: The Version 3 results 12 cover --
13 MR. THURSTON: Are very similar.
14 DR. SCHULTZ: -- most of the issue, most 15 of the good comparisons, but then some modifications 16 were done that --
17 MR. THURSTON: That's correct.
18 DR. SCHULTZ: -- helped out others?
19 MR. THURSTON: That's correct.
20 DR. SCHULTZ: Okay. Thank you.
21 DR. LU: We actually had -- we are going 22 to have the slides comparison.
23 DR. SCHULTZ: No, that's very helpful to 24 know that this wasn't a major modification that 25 changed everything to --
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214 1 DR. LU: No, the -- yes, we have --
2 DR. SCHULTZ: Some things needed to be 3 tweaked?
4 MR. THURSTON: Correct.
5 DR. LU: Right. Okay, let's go back to, 6 so, CHF model, after they introduce the IORV, of 7 course the SAFDL and also CHF becomes key. But even 8 before that, because they also claim that the collapse 9 level is so high and they never reach CHF, even doing 10 LOCA.
11 So, that's also, you know, we dived into 12 their CHF correlation, make sure they are really okay.
13 And the good news here is, most of the limiting time, 14 it's only a couple of seconds.
15 Right after that one, if both LOCA and 16 IORV goes down, you have a scram, the power goes down 17 and now CHF is not limiting. But because of that 18 fissures of only a couple of seconds of CHF is 19 limiting, if you just think about it for IORV or LOCA 20 case, your initial couple of seconds of 21 destructuration.
22 If you have a sudden from single phase to 23 suddenly you have flashing of liquid in the core.
24 Actually, that takes a lot of energy away from the 25 cladding.
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215 1 So therefore you, anybody apply the steady 2 state, no matter whether public domain or proprietary 3 correlation, which are thoroughly conservative. So 4 therefore we don't, we agree with them about the 5 design with such a passive feature.
6 And also, so much margin. And the CHF is 7 not an issue for them, for IORV and LOCA.
8 All right, let's keep going. All right, 9 since this can be a public domain, so I'm going to 10 talk about Item 5. And so that I don't have to repeat 11 this pro-limitation as a condition is part of the 12 proprietary session presentation. So you have any 13 questions on that one, let me know, okay?
14 So first, of course, they ask for the 15 removal, or reduce of requirement, from Appendix K.
16 Even before they submitted a topical report.
17 And so anybody, any applicant or licensee 18 decided to view the plan with this one, they need to 19 ask for the exemption from Appendix K requirement.
20 Which is, can be done fairly easily.
21 So, since the figure of merit of this, the 22 LOCA EM model is limited to the evaluation of LOCA and 23 of CHF is not exceeded, it has to be, any revision of 24 this in this application of this topical report 25 limited to the point that the containment temperature NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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216 1 and pressure remain below the delay limit. And the 2 core collapsible level is above the core.
3 If it goes into core uncovery, this 4 approval can no longer be valid.
5 And the, no credit for DHR heat removal 6 exits right now. Which is conservative I think to 7 their credit. And I think if they do have, DHRS doing 8 LOCA, it definitely will remove a lot of heat. There 9 is no doubt about it.
10 And therefore, but they decided not to 11 take the credit from that, which is very conservative.
12 But if they decided to take on that one, they have to 13 accredit that one, DHR's heat removal at certain 14 point. And they need to submit a revision of these 15 topical report with further review and approval.
16 CO-CHAIR MARCH-LEUBA: So once ECCS 17 actuates and you reach agreement, this simulator is 18 completely uncovered, right?
19 DR. LU: Yes.
20 CO-CHAIR MARCH-LEUBA: This DHRS doesn't 21 really work.
22 DR. LU: That's right.
23 CO-CHAIR MARCH-LEUBA: That's the only 24 reason --
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217 1 does, you still have a cooling there. If you have 2 cooling there --
3 CO-CHAIR MARCH-LEUBA: Cooling steam 4 condensing but --
5 (Simultaneous speaking.)
6 DR. LU: But, yes, you're right that the, 7 actually, because the steam bays above the wall level 8 in the collapsible level in the downcomer are very 9 stagnant.
10 But if you have condensation going on, so 11 you can, still can calculate steam coming down to the, 12 through the steam generator and the condensate 13 ejection water towards the downcomer collapsible 14 level. That's can happen to it.
15 CO-CHAIR MARCH-LEUBA: If DHRS is 16 available for LOCA, at least long-term, it's a very 17 small addition to the heat removal.
18 DR. LU: Yes. So --
19 CO-CHAIR MARCH-LEUBA: It's not worth for 20 them if they can live without to have to answer all 21 your 45 RAIs on the subject.
22 DR. LU: Yes, I got it.
23 CO-CHAIR MARCH-LEUBA: So, everything is 24 good.
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218 1 of LOCA EM is limited to evaluations of analysis of 2 the downcomer described in the TR, which is the 3 condensation pressure CHF and the collapsible level 4 for LOCA and IORV event.
5 And the approval is limited on version 1.4 6 and NPM Model Revision 2. And if they decide to make 7 a change they need to submit a change process for our 8 approval for any other revision change of the dyn 9 (phonetic) changes there.
10 Other part of the features of this design, 11 single failure criteria, electrical power assumption 12 and the need for operator action are not approved 13 within this methodology. This just methodology 14 approved for LOCA and IORV analysis.
15 And we did have specific CHF limitation 16 conditions there. And we are going to talk about that 17 one in detail during the proprietary session.
18 So those are the approval limitations and 19 conditions. Yes?
20 CO-CHAIR MARCH-LEUBA: So, I'd like to 21 bring up the same topic I brought up this morning with 22 Applicant.
23 DR. LU: Okay.
24 CO-CHAIR MARCH-LEUBA: After the ECCS 25 valves were tested and the actuation pressure values NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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219 1 were identified, which I call set points, and I want 2 to call it set points, they had to rerun all of their 3 analysis because they have calculated with the wrong 4 set point.
5 And one of the things they did was stagger 6 the opening of the valves. You have to open the 7 second IORV valve first and then you open the RPV.
8 DR. LU: Right.
9 CO-CHAIR MARCH-LEUBA: Where is that 10 specified that that's part of the methodology?
11 It has to be either on this topical report 12 or it has to be on Chapter 15 specifying what we mean 13 by the IORV. And I don't --
14 DR. LU: Okay.
15 CO-CHAIR MARCH-LEUBA: -- see it in either 16 of the two.
17 And I think the proper place to have it is 18 a condition here that say, for future applications of 19 this methodology to IORV, though shall use the worst 20 staggering of valves.
21 DR. LU: Right.
22 CO-CHAIR MARCH-LEUBA: What do you think?
23 I mean, that's what I would do. I would add a 24 condition here.
25 DR. LU: Yes.
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220 1 MR. THURSTON: So, the LOCA TR does 2 acknowledge that there's a potential for staggered 3 release of the ECCS valves. But the actual evaluation 4 of those set points are targeted, are appointed to 5 Chapter 15 analysis.
6 CO-CHAIR MARCH-LEUBA: Yes.
7 MR. THURSTON: Yes.
8 CO-CHAIR MARCH-LEUBA: Well, it is 500, 9 700 or 1,200 psi, that's a design issue. But the fact 10 that this methodology, though shall use the worst 11 opening time possible within your tolerances. It's 12 part of the methodology.
13 MR. THURSTON: It should be. It should be 14 yes. So that I agree with. Yes.
15 DR. LU: Okay.
16 CO-CHAIR MARCH-LEUBA: So, I mean, I would 17 like to add the recommendation to overview. With 18 somebody at overview.
19 MR. THURSTON: Well --
20 MS. PATTON: So I think, and Carl, maybe 21 you can explain this, but I think you looked at some 22 of those sensitivities when we audited, right? So how 23 big was the difference?
24 MR. THURSTON: Yes. So the differences 25 between the staggered and changing, so initially the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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221 1 release set point was at 1,000. And through the 2 testing, the verification testing that they did, they 3 found that it could be as low as 900.
4 CO-CHAIR MARCH-LEUBA: That might be 5 proprietary.
6 MR. THURSTON: Oh, that might be 7 proprietary.
8 DR. LU: Yes.
9 MS. PATTON: I guess what I meant is, how 10 much of a difference did it make, Carl?
11 MR. THURSTON: Thanks.
12 CO-CHAIR MARCH-LEUBA: The difference was 13 a few psi --
14 DR. LU: Yes.
15 CO-CHAIR MARCH-LEUBA: -- but it was half 16 of the margin.
17 MR. THURSTON: So, for the LOCA cases that 18 we considered --
19 CO-CHAIR MARCH-LEUBA: I don't know, maybe 20 we can postpone this to the closed session.
21 DR. LU: It's better to talk about it 22 later on.
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222 1 is small. For the limiting case of collapsed liquid 2 level. Which is the limited case for LOCA.
3 CO-CHAIR MARCH-LEUBA: I see it refers to 4 peak containment pressure. It was a small change, but 5 that small change equals half the margin.
6 MR. THURSTON: No, that's not the case.
7 CO-CHAIR MARCH-LEUBA: No?
8 MR. THURSTON: Yes, that's not the case.
9 CO-CHAIR MARCH-LEUBA: That's what I 10 remember.
11 MR. THURSTON: It's a small change.
12 CO-CHAIR MARCH-LEUBA: They have to 13 increase the margin, the operating pressure of the 14 containment just in case.
15 MR. THURSTON: Well --
16 CO-CHAIR KIRCHNER: No, that's not 17 correct.
18 MR. THURSTON: -- that was NuScale's 19 prerogative.
20 CO-CHAIR KIRCHNER: They, NuScale changed 21 it once they admitted to the fact that you could have 22 inadvertent operation, opening of the RRV. That's the 23 dominant scenario for peak containment pressure.
24 MR. THURSTON: Yes.
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223 1 the containment pressure from 1,000 psi --
2 CO-CHAIR MARCH-LEUBA: To 1,050.
3 CO-CHAIR KIRCHNER: -- in Rev 2, to 1,050 4 in Rev 3. I think that's a matter of public record.
5 The actual change in the sequences, 6 sequencing the valves, the various combinations was 7 less than ten psi difference.
8 CO-CHAIR MARCH-LEUBA: Which was half of 9 the margin.
10 MR. THURSTON: No.
11 CO-CHAIR MARCH-LEUBA: To the existing --
12 CO-CHAIR KIRCHNER: No.
13 MR. THURSTON: No, it's not half of the 14 margin.
15 CO-CHAIR KIRCHNER: Not half of the 16 margin.
17 DR. LU: Maybe that's the part I think 18 that more detail --
19 CO-CHAIR MARCH-LEUBA: Maybe we can --
20 CO-CHAIR KIRCHNER: Yes, we can do that --
21 MR. THURSTON: This margin is 22 significantly larger.
23 DR. LU: Yes. Actually, I think Carl is 24 right. And because of the margin of so much margin we 25 have in terms, from the PCT perspective, peak NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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224 1 containment pressure --
2 CO-CHAIR MARCH-LEUBA: PCT, this --
3 DR. LU: -- in the peak pressure, even for 4 the peak containment pressure side, we did a lot of 5 sensitivity cases.
6 When we had IAV issued, they, you know, 7 when the opening and the timing and the closed timing.
8 So we had those calculations there. And --
9 CO-CHAIR MARCH-LEUBA: Let's talk in the 10 closed session --
11 DR. LU: Yes. Yes.
12 CO-CHAIR MARCH-LEUBA: -- so we can talk.
13 DR. LU: We can talk about it, yes.
14 CO-CHAIR MARCH-LEUBA: And I have --
15 DR. LU: We can go to the closed session 16 and see whether we can address your comments, yes.
17 CO-CHAIR MARCH-LEUBA: In my opinion, if 18 the staggering of the valves makes a difference, does 19 it make a difference? Does it make a significant 20 difference?
21 DR. LU: Yes. If it does make 22 significant, I think I would agree with you. At this 23 point we have not seen that.
24 MR. THURSTON: That's right.
25 DR. LU: Because that's the point --
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225 1 MR. THURSTON: Because think about it, 2 when one of the valves open --
3 CO-CHAIR MARCH-LEUBA: That's not what --
4 MR. THURSTON: Well, it's not going to be 5 a large difference.
6 (Off-microphone comments.)
7 CO-CHAIR MARCH-LEUBA: Yes, let's postpone 8 it to closed.
9 DR. LU: Okay. I exceeded my time by a 10 factor of three, sorry.
11 CO-CHAIR KIRCHNER: We'll take it out of 12 your next colleague.
13 DR. LU: Okay. All right. Conclusion 14 slide. And then we are proving that the version 15 NRELAP5.14 and we agree that it's in compliance with 16 10 CFR 50.46, Appendix K.
17 And the 1.4 is associated with the NPM 18 model is approved. And then we also proved that CHF 19 modeling is up to limitation and condition for both 20 LOCA and IORV. Those are the conclusions that we 21 have.
22 So, pretty much what they asked, and then 23 we approved.
24 CO-CHAIR KIRCHNER: Shanlai, could you 25 just go back to, you don't have to show it, it was on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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226 1 Slide 11 where you mentioned the new limiting LOCA 2 test was conducted.
3 For the record, how late in the game, so 4 to speak, was that discovered?
5 DR. LU: It's hard --
6 CO-CHAIR KIRCHNER: That they are running 7 at half power.
8 DR. LU: Half power it was, when we 9 launched the first initial face-to-face onsite audit, 10 and then we got a lot of information, so we came back.
11 We were digesting a lot of stuff. The team had to 12 work, and then somehow we asked them, what is 13 distortion reported, at that time the distortion 14 report was not completed yet.
15 So we want to know, where is this, what 16 exactly there. So our scaling team, and then also the 17 staff, we dive into their, later they submit a 18 technical report regarding this scaling and distortion 19 analysis. The update scaling and the distortion 20 analysis.
21 So we checked that one and said how come, 22 a very specific statement saying, the power, decay 23 heat power previously done before was half of the 24 scale of the power.
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227 1 asked what's going on there. It's confirmed that 2 actually the previously LOCA tests were done at half 3 of the decay power.
4 So we said, sorry --
5 CO-CHAIR KIRCHNER: But that was early in 6 the process.
7 DR. LU: Early in the process. That's 8 very early phase of, within half, half year. Nine 9 months.
10 CO-CHAIR KIRCHNER: Okay.
11 DR. LU: Yes.
12 CO-CHAIR KIRCHNER: Because I'm reflecting 13 on Jose's statement earlier this morning where had you 14 discovered that very late in the process of approving 15 we would perhaps, maybe this, they have enough margin 16 it didn't have an impact, but that would be something 17 of concern.
18 DR. LU: Yes. Sure.
19 CO-CHAIR KIRCHNER: And throw a question 20 about the validity of RELAP validation for the Chapter 21 15 analysis.
22 DR. LU: Yes. I think that's exactly the 23 reason I think the management asked the team to engage 24 with NuScale as early as possible. And in support, 25 asked for additional onsite audit and then face-to-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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228 1 face meeting. Which enabled us to identify the, on 2 the proper issues there.
3 CO-CHAIR KIRCHNER: So, at some point we 4 will do a lessons learned, from our perspective, of 5 our review, of this application. And we're taking 6 notes and looking forward to future reviews of advance 7 concepts.
8 And so there is something learned there 9 about doing a good thorough, early engagement and 10 audit with the Applicant.
11 DR. LU: Yes.
12 CO-CHAIR KIRCHNER: You wouldn't want this 13 to be uncovered in December of last year --
14 DR. LU: At our 11th hour.
15 CO-CHAIR KIRCHNER: -- finishing your SER.
16 DR. LU: Thank you.
17 CO-CHAIR KIRCHNER: Okay.
18 DR. LU: And actually -- that's actually 19 all our review and our branch review. We worked even 20 late in March in order to do that.
21 CO-CHAIR KIRCHNER: Okay.
22 DR. LU: As early as possible.
23 CO-CHAIR MARCH-LEUBA: I'm getting a 24 little tired of being corrected, so I do on the 25 record. I'm looking at the slide that we were NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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229 1 presented earlier this month and it says 2 nonproprietary everywhere, so I can talk about these 3 numbers.
4 When the containment pressure was 1,000 5 psi the original analysis of peak pressure come out 6 with 986 psi.
7 DR. LU: Right.
8 CO-CHAIR MARCH-LEUBA: When you stagger, 9 this here. When you stager the opening of the valves 10 it moved to 994. Very small change.
11 But this is more than half of the margin 12 to 1,000. And after this test, the pressure was 13 increased to 1,050.
14 MR. THURSTON: The margin for peak 15 containment pressure was initially 1,000, you're 16 right. And the first analysis that provided us with 17 peak containment pressure I think was 951.
18 CO-CHAIR MARCH-LEUBA: 986.
19 DR. LU: 951.
20 MR. THURSTON: It was 951.
21 CO-CHAIR MARCH-LEUBA: The Revision 1, 22 something.
23 DR. LU: Right.
24 MR. THURSTON: So the numbers that you are 25 referring to are later iterations of the peak NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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230 1 containment calculation. At that time, the limiting 2 design pressure had been increased to 1,050.
3 CO-CHAIR MARCH-LEUBA: Right. But the 4 staggering of the IAV increased by eight psi, which is 5 not much.
6 MR. THURSTON: Under ten, that's right.
7 CO-CHAIR MARCH-LEUBA: But it got to 994.
8 MR. THURSTON: That is correct. So that 9 is the new --
10 CO-CHAIR MARCH-LEUBA: So it's not 11 insignificant.
12 MR. THURSTON: Right. So, in comparison 13 to the original design pressure, I see your point.
14 But, all right.
15 CO-CHAIR MARCH-LEUBA: Yes.
16 MR. THURSTON: Okay.
17 DR. LU: With the 1,050 there might be 18 enough to accommodate --
19 MR. THURSTON: Yes. With 1,050.
20 CO-CHAIR MARCH-LEUBA: You have 50 psi 21 margin, which you didn't have it.
22 MR. THURSTON: That's right.
23 DR. LU: Right. Okay, that's --
24 CO-CHAIR KIRCHNER: Okay, Members, further 25 questions of Shanlai before he escapes?
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231 1 All right, so let's move on then to the 2 non-LOCA, no, is this right?
3 CO-CHAIR MARCH-LEUBA: Non-LOCA. Yes.
4 Non-LOCA open.
5 CO-CHAIR KIRCHNER: Yes, the non-LOCA open 6 session. And their staff. Alex, do we turn to you?
7 MS. SIWY: Yes.
8 CO-CHAIR KIRCHNER: Please, proceed.
9 MS. SIWY: If everybody is all set. My 10 name is Alex Siwy and I am the technical reviewer in 11 NRR. And I will present to you the staff's review of 12 NuScale's non-LOCA methodology topical report.
13 As you can see from this slide, we had a 14 lot of contributors to the review. This included 15 staff from NRR, the Office of Research, as well as 16 consultants from Energy Research Incorporated.
17 And at the table with me here today are 18 Dr. Mohsen Khativ-Rahbar of ERI, and Walter Tauche 19 supporting ERI.
20 So, to provide a pretty broad overview of 21 the staff's review process, the staff safety 22 evaluation is based on Revision 2 of the topical 23 report.
24 The staff conducted two audits in four 25 phases as part of its review. And we examined about NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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232 1 140 different audit issues as part of the audits.
2 Ultimately the audits really helped to 3 confirm the staff's understanding of docketed 4 information, as well as to inform RAIs.
5 And that's also evidence in the number of 6 RAIs that were ultimately issued. We issued 33 7 different RAI questions in five RAI letters.
8 And as of now, all of these issues are 9 resolved. And the responses have been incorporated 10 into Revision 2 of the topical report as appropriate.
11 The staff conducted its review in 12 accordance with NRC requirements and guidance.
13 Several general design criteria are applicable to non-14 LOCA transient analysis.
15 While the staff didn't make direct 16 findings with regard to the GDC as part of the non-17 LOCA review, the staff did consider how the non-LOCA 18 methodology would be applied in order to support those 19 findings.
20 The staff also utilized several guidance 21 documents as part of its review. These include SRP 22 Section 1502, review of transient and accident 23 analysis methods, Regulatory Guide 1.203, transient 24 and accident analysis methods, DSRS Section 15.0, 25 introduction, transient and accident analyses.
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233 1 As well as the event specific SRP in DSRS 2 sections that are applicable to the various non-LOCA 3 events.
4 So NuScale already discussed some of the 5 Chapter 15 acceptance criteria, but this slide just 6 provides an overview of the acceptance criteria 7 outlined in DSRS Section 15.0. Which is applicable to 8 NuScale design basis events.
9 And the key takeaways here are that there 10 are acceptance criteria for primary and secondary 11 pressures, fuel failures and event escalation.
12 NuScale also covered the scope of the non-13 method, non-LOCA methodology in detail, but just to go 14 through it quickly again.
15 The non-LOCA methodology is, it provides 16 a conservative methodology for performing system 17 transient analysis of the specific non-LOCA design 18 basis events for the NPM.
19 The non-LOCA methodology directly 20 evaluates the primary and secondary pressure figures 21 and merit. And it's also used to evaluate whether a 22 staple condition has been reached at the end of the 23 transient.
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234 1 acceptable fuel design limits and radiological 2 consequences. So the non-LOCA methodology itself 3 doesn't evaluate against those acceptance criteria.
4 The non-LOCA EM covers the timeframe 5 during which mixture level is maintained above the top 6 of the riser and natural circulation is maintained.
7 There are separate methodologies that cover the longer 8 term periods after a non-LOCA.
9 And finally, the scope of the non-LOCA EM 10 includes certain event specific assumptions, as well 11 as conservative bias directions for initial 12 conditions.
13 So this slide --
14 CO-CHAIR MARCH-LEUBA: We have seen this 15 slide before.
16 MS. SIWY: Yes, we can move on. But the 17 key here is that the methodology covers several 18 different events from all event categories. With the 19 exception of the decreased in RCS flow events because 20 they're not applicable to the natural circulation 21 design.
22 I just wanted to highlight some of the 23 items that the staff is not reviewing in detail or 24 that the staff is not making a finding on as part of 25 the non-LOCA topical report review because these are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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235 1 more appropriate for a design specific application of 2 the methodology.
3 These include the representative 4 calculation results in Section 8 of the topical 5 report. Example, values of parameters or parameter 6 biases that are discussed in Section 7 and 8 of the 7 topical report.
8 The limiting loss of power and single 9 failure scenarios. Any need for operator actions. As 10 well as credit for secondary main steam isolation 11 valves that are not safety related.
12 So the staff reviewed the entire non-LOCA 13 transient analysis process to make sure that it would 14 result in an acceptable calculation. The non-LOCA EM 15 takes input from upstream methodologies, including 16 nuclear and fuel design.
17 And the steps within the non-LOCA EM 18 include adapting NRELAP5 base model for event specific 19 analyses. This includes assumptions related to the 20 initiating event, conservative biases and things of 21 that nature.
22 Then steady state initialization 23 calculations are performed. This includes running no 24 transients. Or basically running a steady state 25 calculation in transient mode to ensure that the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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236 1 steady state convergence criteria have been met.
2 The next step is performing the transient 3 calculations, which is pretty self-explanatory. Then 4 the results of the transient calculations will be 5 evaluated against the primary and secondary pressure 6 acceptance criteria.
7 And finally, the cases for downstream 8 analyses are identified. These include the subchannel 9 analyses and radiological analyses.
10 CO-CHAIR MARCH-LEUBA: And I've been 11 complaining about this step the whole morning. And 12 part of the afternoon now.
13 Are those acceptance criteria, and 14 criteria, to define downstream, need for downstream 15 analysis more than identified in the methodology, I 16 don't think so.
17 MS. SIWY: So, I would say that there are 18 no numerical types of criteria. But the methodology 19 does specify that the pre-screening process can be 20 used as an aid. And if necessary, a spectrum of cases 21 will be sent for downstream subchannel analysis.
22 CO-CHAIR MARCH-LEUBA: Yes. My problem 23 is, when the CHF correlation that is used is so way 24 off reality, how, unless you use a very conservative 25 screening criteria, how can you rely on it?
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237 1 MS. SIWY: So, we have pretty good 2 confidence that the trend remains valid. I mean, as 3 NuScale said, it's not going to correctly assess the 4 margin but it does, from what we've reviewed, provide 5 a reliable indicator as to which cases would be 6 limiting.
7 And for the challenging events, there will 8 be more cases that are subjected to the downstream 9 subchannel analysis.
10 CO-CHAIR MARCH-LEUBA: What really saves 11 us is that none of these transients' challenge limits 12 anyway. We don't even have to run to know that 13 everything is going to be okay.
14 But the, as long as you say, will run them 15 all and pick the five worst and then analyze those 16 instead of leaving them to the next, you know, they 17 still going to happen to work out there to do this 18 issue. I mean, there should not be judgement.
19 You should not have to rely on the good 20 judgement, on the analyst, 20 years from now, to do 21 the right thing. I mean, you should just say, pick 22 the worst five in these categories, or whatever, and 23 then those need to be run with VIPRE. You know what 24 I mean?
25 I mean, we're leaving it hanging. We're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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238 1 relying on the good judgment of, grab the student that 2 just has been hired to run these calculations. To do 3 the right thing.
4 DR. KHATIB-RAHBAR: I think it is 5 desirable to have a perspective value.
6 CO-CHAIR MARCH-LEUBA: Pull the 7 microphone.
8 DR. KHATIB-RAHBAR: It is desirable to 9 have a --
10 CO-CHAIR KIRCHNER: Down below.
11 CO-CHAIR MARCH-LEUBA: There.
12 DR. KHATIB-RAHBAR: -- descriptive number.
13 But what they've done I think there is sufficient 14 margin here that a qualitative way of doing it is 15 probably deceptive.
16 CO-CHAIR MARCH-LEUBA: What I'm saying is 17 that we know what the result is.
18 DR. KHATIB-RAHBAR: Yes.
19 CO-CHAIR MARCH-LEUBA: We're just 20 demonstrating it. But I would have liked to see 21 something more prescriptive.
22 MR. TAUCHE: You know you have margin, but 23 you have to have something for your licensing basis.
24 So I need to determine which are the limiting cases 25 that I want to show as my licensing basis.
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239 1 And since you have potentially a large 2 number of different cases for the various events, the 3 question becomes then, how do you categorically decide 4 I need to run five cases out of these ten or two cases 5 out of these 20.
6 And I think what they have done is a 7 reasonable approach to say that these cases have been 8 clearly identified as based upon the dummy hot rod as 9 those cases that are going to be the closest to the 10 limit and therefore those are the ones that need to be 11 subjected to the VIPRE analysis.
12 CO-CHAIR MARCH-LEUBA: What I mean is 13 that, out of the cases that we run VIPRE already 14 identify?
15 MR. TAUCHE: They're not already 16 identified but -- and they are subject to 17 interpretation, but what they have shown so far is 18 that their ability to identify the cases that are 19 going to be most limiting, can come from that 20 selection from the hot rod criteria.
21 So the question is, do I run three or four 22 cases out of 20, do I run five or ten cases out of 20 23 or do I need to only run those that I know based upon 24 the hot rod dummy analysis are the ones that are going 25 to give me the worst situation.
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240 1 CO-CHAIR MARCH-LEUBA: And my point is 2 that you are the experts on it, on the applicant, 3 you've done it now. Make the tough decisions now 4 instead of postponing to allow us to do it 20 years 5 from now. That's what I would have done if I was 6 applying this year.
7 DR. SCHULTZ: Yes. With potentially a 8 higher power core.
9 CO-CHAIR MARCH-LEUBA: Oh yes. Yes. And 10 we'll go into closed session and you can see how bad 11 the CHF correlation is. I mean, it's embarrassing.
12 DR. CORRADINI: I'd be very careful on 13 continuing saying bad. Different, maybe, bad, I would 14 disagree.
15 CO-CHAIR MARCH-LEUBA: It's not 16 conservative.
17 DR. CORRADINI: I just think that's not, 18 I mean personally, I would call it different.
19 CO-CHAIR MARCH-LEUBA: What do you mean 20 different?
21 DR. CORRADINI: Different doesn't mean bad 22 or good, it means different.
23 MS. SIWY: Right. So I think if a limit 24 were to be established on that correlation that would 25 also look different from the limit on the NSP4 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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241 1 correlation.
2 CO-CHAIR MARCH-LEUBA: Okay, we'll wait 3 for the closed session.
4 MS. SIWY: Okay. So anyway, to conclude 5 this slide, the staff reviewed the entire analysis 6 process, including the interfaces with the other 7 methodologies and finds this general process to be 8 acceptable.
9 The staff focused its review on several 10 key design features of the NuScale design that 11 include, but aren't limited to, the natural 12 circulation design. Which introduces different 13 transient behavior as well as different initialization 14 characteristics relative to a traditional large light 15 water reactor.
16 In addition, the staff really focused on 17 the helical coil steam generators which are used to 18 transfer heat from the RCS to the feedwater. As well 19 as the passive DHRS condensers. Which transfer the 20 decay heat to the reactor pool via the steam 21 generators.
22 And the staff also considered the 23 evacuated containment vessel, which, as NuScale 24 pointed out, presents the possibility for a new type 25 of overcooling event.
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242 1 The staff reviewed the applicability of 2 NRELAP5 to non-LOCA analysis. The staff reviewed how 3 NuScale develops the non-LOCA EM, starting from the 4 LOCA EM, using the graded approach in Regulatory Guide 5 1.203. And agrees that the regulatory guidance has 6 been followed.
7 In addition, the staff audited the entire 8 non-LOCA PIRT. And also reviewed the highly ranked 9 phenomena that are described in the topical report.
10 And the staff really wanted to ensure that the 11 important phenomena were adequately addressed as part 12 of the evaluation model.
13 So, as the Applicant indicated earlier, 14 they identified the deltas and highly ranked phenomena 15 relative to the LOCA EM and addressed those deltas 16 using a variety of methods. Including the separate 17 effects tests at NIST as well as the KAIST and SIET 18 assessments, which the staff really evaluated as part 19 of the non-LOCA review since those assessments are 20 more applicable to non-LOCA transients.
21 In addition, there were three integral 22 effects tests included as part of the non-LOCA topical 23 report. A code-to-code benchmark against RETRAN-3D.
24 As well as use of bounding input values and other 25 analysis methodologies. In particular, a lot of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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243 1 highly arranged phenomena were addressed using the 2 subchannel analysis.
3 And this brings up one of the first 4 significant issues that the staff encountered during 5 its review. And it's very high level for this slide.
6 But I wanted to mention it for the benefit of the 7 public.
8 The staff requested additional 9 justification for how certain multidimensional flow 10 effects in the RCS, as well as thermal stratification 11 in the reactor pool are addressed as part of the non-12 LOCA EM.
13 The Applicant provided an RAI response, 14 which the staff was able to confirm via audits and 15 some audit discussions with the Applicant.
16 DR. CORRADINI: Which one was that?
17 MS. SIWY: This is RAI 9351, Question 18 15.0.2-31.
19 DR. CORRADINI: Right. Okay.
20 (Laughter.)
21 DR. CORRADINI: You were ready for this, 22 I --
23 MS. SIWY: Some of this is just ingrained 24 into my memory by now.
25 (Laughter.)
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244 1 MS. SIWY: Anyway. The RAI response 2 ultimately confirmed, through the staff's audit, 3 resolved the issue. And I'll provide more details on 4 this in the closed session.
5 The staff reviewed --
6 DR. CORRADINI: If I might back you up.
7 MS. SIWY: Sure.
8 DR. CORRADINI: So we are going to hear 9 about this in the closed session?
10 MS. SIWY: Yes.
11 DR. CORRADINI: Okay.
12 MS. SIWY: There will be more details 13 then.
14 DR. CORRADINI: Okay, fine. Thank you.
15 MS. SIWY: Yes. Okay, so the staff 16 reviewed NRELAP5 assessments against the KAIST. And 17 this one, HP-03 and HP-04 tasks, which predominately 18 deal with DHRS phenomena. Particularly, the 19 condensation inside and the heat transfer across the 20 DHRS tubes. This cooling pool vessel fluid.
21 The staff confirmed that the ranges of the 22 KAIST and missed one validations adequately cover the 23 expected ranges during DHRS operation in the NPM. In 24 addition, the staff finds that the facility 25 nodalization is acceptable relative to the NPM NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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245 1 nodalization.
2 DR. CORRADINI: So, can I ask about this?
3 MS. SIWY: Sure.
4 DR. CORRADINI: Did the staff do an audit 5 calculation?
6 MS. SIWY: By audit calculations do you 7 mean --
8 DR. CORRADINI: Something else with 9 somebody else?
10 MS. SIWY: So did we perform confirmatory 11 calculations?
12 DR. CORRADINI: Yes. Yes.
13 MS. SIWY: The staff performed 14 confirmatory calculations against KAIST in order to 15 validate its TRACE models. So that was one thing that 16 was performed.
17 DR. CORRADINI: Use TRACE --
18 MS. SIWY: Yes.
19 DR. CORRADINI: Okay. Because, I don't 20 know how much I should say now or not. If you're 21 going to come back to this in closed session I'll just 22 wait.
23 MS. SIWY: We will touch a little bit more 24 on the NIST tests in the closed session, but I wasn't 25 planning to cover KAIST.
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246 1 DR. CORRADINI: Well, there was a 2 discussion about force convection versus natural, 3 force circulation versus natural circulation.
4 MS. SIWY: That will come up in the closed 5 session.
6 DR. CORRADINI: Okay, fine.
7 MS. SIWY: That relates to the NIST tests.
8 DR. CORRADINI: Okay, thank you.
9 MS. SIWY: Sure. And through the staff's 10 audit of underlying calculations, as well as review of 11 the topical report and RAI responses on these topics, 12 the staff finds that the NRELAP5 prediction of the 13 KAIST and HP-03 data is reasonable to excellent. And 14 the agreement, the code-to-data agreement for HP-04 is 15 reasonable.
16 DR. CORRADINI: What do those adjectives 17 means? Reasonable versus excellent.
18 So, is there a number that goes with 19 reasonable and a number that goes with excellent or --
20 MS. SIWY: They are --
21 DR. CORRADINI: I'm trying to understand 22 because I'm reading it here and --
23 MS. SIWY: Right. So this terminology is 24 taken out of Regulatory Guide 1.203.
25 DR. CORRADINI: Okay.
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247 1 MS. SIWY: So, excellent agreement could 2 be characterized as the trends follow the data in 3 there within, well within the uncertainty bands.
4 DR. CORRADINI: Okay, fine.
5 MS. SIWY: And reasonable is they still 6 follow the trends reasonably well but they might 7 disagree in magnitude a little bit more.
8 DR. CORRADINI: All right, thank you very 9 much.
10 MS. SIWY: Sure. So, this is a good segue 11 into one of the significant issues that we encountered 12 during our review. And just to preface this, there 13 will be more detail during the closed session.
14 But one of the issues was removal of the 15 DHRS heat transfer bias from the methodology that was 16 described in Revision 0 and 1 of the topical report.
17 And the Applicants decided to remove this 18 bias because they performed extensive sensitivity 19 studies that showed that there was an insignificant 20 effect of DHRS performance and nodalization on non-21 LOCA figures of merit.
22 And there was some key underlying reasons 23 for this. The first is that the minimum margin to 24 acceptance criteria typically occurs before the DHRS 25 is fully active.
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248 1 In addition, peak RCS pressure is 2 typically limited by the lifting of the RSB that's 3 credited in NRELAP5. And peak secondary pressure is 4 really a function of the secondary inventory and 5 primary site conditions at the time of DHRS actuation.
6 So it tends to be insignificantly affected 7 by DHRS heat transfer variations. Especially given 8 the margin in secondary pressure.
9 So the staff, in conclusion, finds that 10 the sensitivity studies that the Applicant performed, 11 as well as the physical dynamics, support removal of 12 the DHRS heat transfer bias for NPM Model Revision 2.
13 However, there is an associated limitation and 14 condition that I'll discuss at the end of the 15 presentation.
16 The staff also reviewed the assessment of 17 NRELAP5 against SEIT TF-1 and TF-2 test data. The TF-18 1 test assessed flow within the steam generator tubes, 19 and TF-2 assessed the primary to secondary heat 20 transfer.
21 DR. CORRADINI: Can I just, I'm sorry that 22 I'm slow to catch up to you, but can I make sure I 23 understand.
24 MS. SIWY: Okay.
25 DR. CORRADINI: When you say bias, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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249 1 mean it's a timing bias?
2 I turn it on or off and it's essentially 3 inapplicable based on SRV operation and timing of the 4 event? That's how I read your three sub-bullets 5 there.
6 MS. SIWY: Okay. I'm sorry that wasn't 7 clear for you. I'll back up.
8 DR. CORRADINI: It's okay. A lot of 9 things are unclear for me.
10 (Laughter.)
11 MS. SIWY: I take for granite that I've 12 been working on this for quite some time.
13 DR. CORRADINI: No, no, no, I was just 14 trying to understand. Okay, go ahead.
15 MS. SIWY: Okay. So the heat transfer 16 bias is basically applied as a multiplier on the heat 17 transfer through the DHRS tubes. So it doesn't have 18 anything to do with timing.
19 What I was trying to get at with those 20 three bullets is the underlying reasons why that bias 21 doesn't have a significant effect for non-LOCA events.
22 DR. CORRADINI: But I read the three 23 dashes to mean, the first one occurs because, the MCFR 24 occurs before it kicks in, it's fully active.
25 MS. SIWY: Yes.
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250 1 DR. CORRADINI: The second one is because 2 the valve opens before it fully gets to high pressure.
3 So, in some sense it's a matter of timing and 4 operation of a component that limits you. So the bias 5 is --
6 MS. SIWY: Okay. Yes.
7 DR. CORRADINI: That's how I read what you 8 were saying.
9 MS. SIWY: Yes.
10 DR. CORRADINI: Am I close?
11 MS. SIWY: Yes.
12 DR. CORRADINI: Okay.
13 MS. SIWY: I would agree with that.
14 DR. CORRADINI: Okay, got it. Thank you.
15 MS. SIWY: Yes.
16 DR. CORRADINI: Got it.
17 MS. SIWY: Okay. So picking up where we 18 left off here.
19 So, for the SIET RF-1 and TF-2 20 assessments, the staff audited the underlying 21 calculations and also issued RAIs. Based on this 22 review, the staff finds that there is reasonable to 23 excellent agreement between the NRELAP5 predictions 24 and the SIET test data.
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251 1 concern about the range of conditions validated by the 2 TF-2 test compared to expected ranges in the NPM and 3 what impacts there might be on uncertainties and steam 4 generator heat transfer as a result.
5 This leads into another significant issue 6 that the staff looked at in its review. Which is the 7 removal of the steam generator heat transfer bias.
8 DR. CORRADINI: Was the bias up or down?
9 MS. SIWY: It depended. But it could be 10 plus or minus 30 percent. Depending on --
11 DR. CORRADINI: At a small effect?
12 MS. SIWY: Yes. That's ultimately what 13 we're getting to. The Applicant performed a lot of 14 sensitivity studies for various events and came to the 15 conclusion that there really wasn't a significant 16 impact on non-LOCA figures and merit.
17 The staff audited the sensitivity 18 calculations and agrees with those conclusions. The 19 staff also considered the impacts of steam generator 20 heat transfer on both post-trip conditions and 21 ultimately concludes that you're really limited by the 22 DHRS heat transfer at that point because it's so much 23 less than the heat transfer you can get through the 24 steam generators.
25 In addition --
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252 1 CO-CHAIR MARCH-LEUBA: Because at this 2 point you have very low heat transfer from the inside 3 of the DHRS tubes. I mean, I realize that there is a 4 limit on the condensation rate on the outside, in the 5 pool simulator, but on the inside of the vessel, when 6 you are in DHRS there is no flow.
7 You're just basically a stagnant core with 8 water with boiling, pool boiling in there. That's not 9 very high heat transfer.
10 MS. SIWY: Yes, I think what we're trying 11 to say is that the DHRS is designed to remove decay 12 heat while when the steam generators are removing heat 13 during normal operation. That's a lot --
14 CO-CHAIR MARCH-LEUBA: When you have flow 15 through there.
16 MS. SIWY: Yes. That's fair.
17 CO-CHAIR MARCH-LEUBA: Now, you remove the 18 flow, then you're not removing as much heat as you 19 used to. Anyway, keep going.
20 (Off-microphone comments.)
21 CO-CHAIR KIRCHNER: You better turn on 22 your green light.
23 MR. TAUCHE: Well, it builds up over time 24 through the DHRS once you open the valves. So of 25 course you have low flow at the very beginning, then NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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253 1 the steam condensation occurs, you build the flow 2 through the system.
3 MS. SIWY: Right. Yes. Then it moves 4 more constant.
5 CO-CHAIR MARCH-LEUBA: The moment you 6 isolate the secondary, the steam line and you open 7 DHRS, no more amounts of circulation. You only have 8 whatever steam comes out.
9 MR. TAUCHE: You'll have a decay in the 10 natural circulation while the DHRS builds up.
11 CO-CHAIR MARCH-LEUBA: Oh. No, the system 12 won't by having a water level and having steam coming 13 out and condensing and dropping down.
14 MR. TAUCHE: Yes.
15 CO-CHAIR MARCH-LEUBA: So your flow is 16 sealed. Whatever steam comes up from the top is that 17 water that goes in from the bottom.
18 MR. TAUCHE: Right. When you open the 19 valve initially there is zero flow. From the DHRS, 20 through the DHRS. So it relies upon the condensation.
21 But that effect is, builds up rather rapidly.
22 CO-CHAIR MARCH-LEUBA: Right. Some people 23 can call it natural circulation. It's not natural 24 circulation, it's boil off condensation.
25 MR. TAUCHE: Yes.
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254 1 MS. SIWY: Boiling condensing, yes.
2 CO-CHAIR MARCH-LEUBA: The boil off, the 3 velocity of the liquid in the inside is essentially 4 zero?
5 MR. TAUCHE: At times zero, that's 6 correct.
7 CO-CHAIR MARCH-LEUBA: On a timing -- that 8 too.
9 DR. CORRADINI: It's small.
10 CO-CHAIR KIRCHNER: It's very small.
11 DR. CORRADINI: I think he was saying 12 small, you're saying super small.
13 CO-CHAIR MARCH-LEUBA: It's equal --
14 DR. CORRADINI: But I think you guys are 15 in violent agreement.
16 CO-CHAIR MARCH-LEUBA: It's equal to 17 whatever --
18 MR. TAUCHE: Condensate rate basically is 19 what returns to the vessel. That's what it is.
20 CO-CHAIR MARCH-LEUBA: Which isn't 21 reachable.
22 MR. TAUCHE: Yes. It's just three parts 23 condensation.
24 CO-CHAIR MARCH-LEUBA: As far as this 25 board is concerned it is zero.
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255 1 (Laughter.)
2 CO-CHAIR MARCH-LEUBA: Anyway, keep going.
3 MS. SIWY: Okay. The staff also 4 considered the effects of steam generator heat 5 transfer on normal operation and found that technical 6 specifications would account for the effect of 7 variations in steam generator heat transfer.
8 So these considerations, in addition to 9 the sensitivity studies that the Applicant provided, 10 support removal of the steam generator heat transfer 11 bias for NPM model Revision 2. Again, there will be 12 more details on this during the closed session.
13 The staff also reviewed NRELAP5 14 assessments against the NIST-1 integral effects test.
15 This included auditing the underlying calculation 16 notes.
17 The NLT-02a test was a loss of feedwater 18 transient to the point of a reactor trip. And 19 ultimately the staff concluded that there was 20 reasonable to excellent code-to-date agreement.
21 NLT-02b test was DHRS driven cooling 22 following initial DHRS actuation. And the staff finds 23 that there is reasonable or a better agreement for 24 this test.
25 Due to staff questions on the version of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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256 1 the assessment summarized in Revision 0 and 1 of the 2 non-LOCA topical report, the Applicants also added the 3 NLT-15P2 test, which is a loss of feedwater transient 4 with subsequent DHRS cooling. And the staff finds 5 that there was reasonable code-to-date agreement for 6 this test.
7 So, looking at all three of these tests, 8 the staff finds that they demonstrate the 9 applicability of NRELAP5 to evaluate non-LOCA 10 transients.
11 In addition, the staff evaluated the 12 NRELAP5 assessment using the benchmark against RETRAN-13 3D. RETRAN-3D is a general purpose thermal hydraulics 14 code that the NRC staff has approved for generic use 15 as well as plant-specific applications.
16 So the staff found that the comparisons of 17 NRELAP5 to RETRAN-3D are useful as part of the NRELAP5 18 assessment.
19 The staff reviewed some of the modeling 20 assumptions that the Applicant implemented in the 21 RETRAN-3D model. Which included a simplification 22 since RETRAN-3D does not account for helical coil 23 steam generators.
24 The staff reviewed, do you have a 25 question? No, okay.
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257 1 CO-CHAIR MARCH-LEUBA: I was waiting for 2 you to finish.
3 MS. SIWY: Okay.
4 CO-CHAIR MARCH-LEUBA: But I do. I do 5 have a question.
6 MS. SIWY: Okay. The staff reviewed the 7 Applicant's comparison of four different reactivity 8 transients. This includes both a slow and a fast, 9 uncontrolled control rod withdrawal from full power, 10 a power reduction from 100 to 50 percent in a dropped 11 control rod assembly.
12 Based on these comparisons, the staff 13 finds that there is reasonable or better agreement for 14 each of these transients, which provides confidence in 15 the ability of the NRELAP5 connects model.
16 CO-CHAIR MARCH-LEUBA: Now, I want to mess 17 with your memory of this event. On this code-to-code 18 comparison, did the Applicant calculate the reactivity 19 worth of the control rod using CASMO/SIMULATE or did 20 they use a RETRAN-3D edit of the reactivity?
21 Because if they used a written 3D edit of 22 the reactivity, written 3D output, the number that you 23 need to put into the point kinetics to get down sort 24 of the return to RELAP. And I suspect that's what 25 they did.
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258 1 MS. SIWY: Yes, I don't recall.
2 DR. KHATIB-RAHBAR: Yes. The reactivity 3 worth of control rods were input to the RELAP model.
4 So RELAP is not used to calculate --
5 CO-CHAIR MARCH-LEUBA: That is correct.
6 DR. KHATIB-RAHBAR: -- is retrying this 7 case.
8 CO-CHAIR MARCH-LEUBA: RETRAN-3D was also 9 point kinetics?
10 DR. KHATIB-RAHBAR: My understanding was, 11 RETRAN was used with the same control rod worth as it 12 was used in NRELAP.
13 CO-CHAIR MARCH-LEUBA: Why do you even 14 bother to do the benchmarking. Number one, it's 15 worthless, number two, its guaranteed to succeed. So 16 --
17 (Simultaneously speaking.)
18 DR. KHATIB-RAHBAR: -- getting from one 19 point versus the other --
20 CO-CHAIR MARCH-LEUBA: If both of the 21 models are point kinetics and both of the models have 22 the same reactivity input, both of the models get the 23 same output. I mean, duh. Please, let me repeat it, 24 duh.
25 MS. SIWY: Okay.
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259 1 CO-CHAIR KIRCHNER: Remember, this is just 2 one member's opinion.
3 (Laughter.)
4 MS. SIWY: Point taken.
5 CO-CHAIR MARCH-LEUBA: One member in fact 6 understands what you're doing.
7 MS. SIWY: Okay. Now that we've covered 8 the assessments of NRELAP --
9 DR. KHATIB-RAHBAR: I guess all I was 10 trying to show, if we use the same feedback effects, 11 the same for hydraulic conditions, the same point 12 kinetics approximation, you get the same answer.
13 That's all it does.
14 CO-CHAIR MARCH-LEUBA: Was there any 15 doubt?
16 DR. KHATIB-RAHBAR: Well, there may have 17 been, yes.
18 MR. TAUCHE: But if RETRAN 3, if RETRAN is 19 acceptable as an NRC method to calculate reactivity 20 effects, then the use of NRELAP5 should be acceptable 21 by the same means for future reactivity calculations.
22 So, sure, it's an extrapolated step. A 23 equals B, therefore B equals C for future use.
24 CO-CHAIR MARCH-LEUBA: And it explains 25 your margin for all these terms anyway. So, it's not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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260 1 an issue but --
2 MS. SIWY: All right, moving on to some 3 important general aspects of the non-LOCA methodology 4 that the staff reviewed. The staff reviewed the 5 Applicants use of suitable conservative input rather 6 than performing an uncertainty evaluation.
7 This includes the use of bounding 8 reactivity parameters, conservatively biased initial 9 conditions, as well as conservative valve 10 characteristics.
11 The methodology also specifies that plant 12 control systems, operation, single failure and loss of 13 power assumptions --
14 MEMBER BROWN: Can I calibrate myself the 15 first bullet?
16 MS. SIWY: Yes.
17 MEMBER BROWN: The way that was read off 18 to me, that sounds like they used a deterministic 19 approach vice a uncertainty, uncertainty type 20 approach. They made judgements as to what was 21 conservative and then went with that as part of their, 22 that's the way I read the fancy language.
23 I just wanted to make sure I can related 24 it to my past experience. Thank you.
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261 1 anyway --
2 CO-CHAIR KIRCHNER: Also, Charlie, then 3 it's, so they're not using best estimate and 4 uncertainty. I mean, that's really what she's saying.
5 MEMBER BROWN: That's what I meant.
6 MS. SIWY: Right.
7 CO-CHAIR KIRCHNER: Yes.
8 MEMBER BROWN: Yes, they kind of, so I'm 9 going to weight it here, I'm going to stop it at that 10 point. That's the worst thing I can think of.
11 Roughly. And therefore it's okay. I'm very familiar 12 with that technic.
13 MS. SIWY: Okay. So the methodology 14 specifies that plant control systems operation, single 15 failure and loss of power assumptions should maximize 16 consequences with regard to acceptance criteria of 17 interests, which is conservative and therefore 18 acceptable.
19 In addition, the methodology specifies 20 that if acceptance criteria are not sensitive to 21 certain inputs or assumptions, nominal values may be 22 used. Which is fine because there is really not a 23 clear conservative bias direction for those.
24 In addition, if an acceptance criterion is 25 not challenged by a particular event, it will not be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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262 1 subjected to in-depth sensitivity studies. And the 2 staff agrees that performing representative 3 calculations for those non-challenged figures of merit 4 is sufficient to determine that they are not 5 challenged.
6 So, overall the staff finds the general 7 approach for the non-LOCA methodology to be acceptable 8 subject to the applicable limitations and conditions, 9 which we'll get to towards the end of the 10 presentation.
11 The staff also reviewed the event specific 12 non-LOCA methodology. And these included methodology 13 aspects specific to each initiating event.
14 The general assessment of which acceptance 15 criteria are challenged by each events. As well as 16 the initial conditions, biases and conservatism.
17 The staff always finds it acceptable for 18 sensitivity studies to be performed on various 19 parameters as part of a plant specific calculation.
20 Because it's going to ensure that the most limiting 21 cases are captured.
22 In addition, some conservative bias 23 directions, or assumptions, are specified as part of 24 the non-LOCA methodology. And for those items the 25 staff confirmed that they are indeed appropriate NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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263 1 conservative with respect to their effect on the 2 acceptance criteria.
3 Overall the staff finds that the event 4 specific methodologies will ensure that conservative 5 results are achieved when the methodology is 6 implemented.
7 Finally, the staff looked at the 8 representative calculations in Section 8 of the 9 topical report. As far as they demonstrate the 10 application of the non-LOCA methodology. Because the 11 staff is not approving those representative 12 calculations as part of the topical report review.
13 The staff concludes that the calculations 14 provided show how the non-LOCA methodology can be used 15 for conservative transient analysis simulations.
16 And the final calculations for the NuScale 17 design are reviewed in the safety evaluation for DCA 18 Chapter 15.
19 So this slide summaries the limitations 20 and conditions in the staff SER. The first is that 21 future changes to the LOCA topical report must be 22 assessed for impacts to the non-LOCA evaluation model.
23 And that's because the non-LOCA model was developed 24 from the LOCA model.
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264 1 with the scope of the non-LOCA evaluation model.
2 Specifically that it's limited to the events described 3 in the topical report prior to the time of riser 4 uncovery for evaluation of primary and secondary 5 pressures and the potential for loss of system 6 functionality.
7 And the actual limitation and condition 8 specifies some things that it is not approved for.
9 The third limitation and condition states 10 that additional justification must be provided for the 11 elimination of steam generator in DHRS heat transfer 12 biases if the methodology is to be applied to a design 13 other than NPM model Revision 2 or a model update that 14 is made pursuant to a change process, specifically 15 approved by the NRC for changes to the NPM model.
16 DR. CORRADINI: So, translate that.
17 MS. SIWY: Basically, a lot of the 18 justification that the Applicant provided for 19 eliminating those biases was based on sensitivity 20 studies that we don't know how the results might 21 change if there are evolutions of the design.
22 DR. CORRADINI: Okay. All right, so as 23 designed and as analyzed it's fine, but if they start 24 changing the operating space they may have to go back, 25 someone may have to go back and do additional NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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265 1 sensitivities?
2 MS. SIWY: Yes, that's correct.
3 DR. CORRADINI: Thank you.
4 CO-CHAIR MARCH-LEUBA: Would that include 5 a power operate?
6 MS. SIWY: That's one potential scenario.
7 CO-CHAIR MARCH-LEUBA: The other scenario.
8 MS. SIWY: Yes.
9 CO-CHAIR KIRCHNER: Now, when you say 10 model, you're referring to actually the nodalization 11 that's used in the code. So if they were to change 12 the approach that was used in that model, how they 13 nodalize the plant, there are a few instances that 14 come to mind where they use pipes for certain 15 components and so on.
16 Any change to that RELAP model, NRELAP5 17 model, would then have to be approved.
18 MS. SIWY: So, we plan, or NuScale plans 19 to submit a change process because we recognize that 20 there are certain changes that might be made that are 21 not important and would not necessitate additional 22 justification.
23 So we're not saying that any little change 24 to the model would require this. We're still working 25 on this. But we do recognize the burden that could be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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266 1 placed if we required justification for any little 2 modeling change.
3 CO-CHAIR KIRCHNER: Okay. See, where I 4 was going with this is, even though my esteemed 5 colleague impugned our ability to understand what's 6 going on here, I do have some background in this area.
7 And I do know that, actually, the 8 nodalization that you would use with RELAP to try and 9 simulate the actual plant versus the integral tests, 10 is going to be different.
11 And I would be interested in the closed 12 session to hear from the staff on their own 13 confirmatory calculations. Because it seems to me 14 that without those confirmatory calculations you 15 wouldn't have a lot of confidence going from the 16 integral effects test that NuScale conducted to the 17 actual NPM. Because of scaling and other issues.
18 And quirkiness, if I might say, of 19 applying, in how the user applies the models and 20 nodalizes the plant going from the integral effects 21 test to the actual plant.
22 So this is where I see great value in the 23 staff doing confirmatory analysis. Because in many 24 cases, that's a big leap in modeling an integral 25 facility and then modeling the actual prototypical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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267 1 plant.
2 And so, in the closed session I'll be 3 interested to hear from the staff, both in LOCA and 4 non-LOCA, how their confirmatory analyses compared 5 with those of the Applicant.
6 MS. SIWY: Okay. And I will add that one 7 of the areas we reviewed was the, how comparable the 8 nodalization of the test facility was to the actual 9 NPM.
10 CO-CHAIR KIRCHNER: Yes, precisely. Thank 11 you.
12 MS. SIWY: Yes.
13 MEMBER BROWN: Can I ask another? I want 14 to understand the changes.
15 MS. SIWY: Okay.
16 MEMBER BROWN: I mean, they've based all 17 this on a specific revision. And obviously, from my 18 past experience in the Naval nuclear program, our labs 19 always were trying to improve the models that for any 20 specific design reactor plant and its application, we 21 had specific versions that were used.
22 And they were not allowed to alter or 23 change anything for application, for operations of the 24 plants already in place, without our specific 25 agreement to the changes. The headquarters agreement.
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268 1 And it sounds like you're looking at a way 2 to allow them to make changes without headquarters 3 agreement. Even though you, how you define 4 insignificant versus significant changes is kind of 5 difficult.
6 Particularly when the licensee thinks he, 7 I'm not impugning anybody's integrity when I say that 8 thinks that these are insignificant whereas your 9 conclusion may be different. It just seems to be a 10 reach to me.
11 I'm just making a question out of that, 12 it's not a statement. Again, one member's opinion of 13 practice, it's kind of hard to imagine that you would 14 allow the analysis basis that we've used for licensing 15 to be modified, in any way, without you all knowing 16 what the changes were first.
17 MS. SIWY: Right.
18 MEMBER BROWN: Now, that's just a comment.
19 MS. SIWY: Okay.
20 MEMBER BROWN: That's all it is.
21 MS. SIWY: That's kind of the idea behind 22 the change process because we recognize that there 23 will be minor error corrections that --
24 MEMBER BROWN: I hate to say this but we, 25 error corrections still had to go through us, they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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269 1 were not done without headquarters approval.
2 MS. SIWY: Okay.
3 MEMBER BROWN: Before we used them to draw 4 conclusions to acceptability on operations. Maybe we 5 were picky, but that's just the way it is.
6 DR. CORRADINI: Yes.
7 MEMBER BROWN: We were really picky. Just 8 as I never allowed changes to my reactor plant 9 protection safeguard systems or instrumentation 10 systems, no changes were made unless they went through 11 me. I was the director of the division, so they all 12 had to go through me one way or the other. And my 13 staff.
14 So, I'm really particularly picky when it 15 comes to models and --
16 DR. CORRADINI: We wouldn't have noticed 17 that.
18 MEMBER BROWN: Pardon?
19 (Laughter.)
20 MEMBER BROWN: If you hadn't figured that 21 out yet.
22 DR. CORRADINI: That escaped my --
23 MEMBER BROWN: You what?
24 DR. CORRADINI: That escaped me.
25 MEMBER BROWN: It did?
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270 1 DR. CORRADINI: Yes.
2 MEMBER BROWN: Okay. I'm just, that's a 3 real caution. I think you're walking a very tenuous 4 path when you start delegating change process to what 5 you'd approve for your licensing basis. It's similar 6 to the changes made to the plant issue.
7 MS. PATTON: So, just to clarify. You 8 know, there are other vendors that have approved 9 changed processes as well. So this is to basically 10 say, if they came in with a change process, and if we 11 approved that change process, then that could be used.
12 That hasn't been done yet.
13 And obviously any changes to the plant, to 14 the DCA, still have to go through the change processes 15 that are specified in the regulations. So this one is 16 specific to the methodology and this one aspect. And 17 it allows that out, if we approve it.
18 Right. But it's not inconsistent with 19 what's been done for other vendors.
20 CO-CHAIR KIRCHNER: I think, Charlie, that 21 if you look at the last bullet, if they actually 22 wanted to change the code and/or the model of the 23 plant they're using, that limit and condition would 24 essentially force them to come back.
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271 1 have to come back and I would expect the staff would 2 make them run all the separate effects and integral 3 tests over again to --
4 MEMBER BROWN: That's not what she said.
5 CO-CHAIR KIRCHNER: But isn't that the 6 process for change?
7 You are limiting it to this version of 8 RELAP, Version 1.4.
9 MEMBER BROWN: You can always go to 1.5, 10 if you get their approval subsequently. But they're 11 saying they're going to allow them to make, a process 12 may be put in place that allows them to make changes 13 under the process, if they fit some category where NRC 14 does not have to review them.
15 CO-CHAIR MARCH-LEUBA: 1.4 refers to a 16 major release. In-between 1.4 and 1.5 there is 17 probably a hundred changes.
18 CO-CHAIR KIRCHNER: No, I understand that.
19 Yes.
20 CO-CHAIR MARCH-LEUBA: Okay. And each of 21 those changes are tested and only major releases get 22 the whole suit of, over education, in this review.
23 And, Charlie, when you were asking for 24 that review in your office, it probably was taking two 25 days. Almost a week if you're on vacation.
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272 1 MEMBER BROWN: I hate to say it never just 2 the two days. It was like sucking blood out of rocks.
3 Okay.
4 CO-CHAIR MARCH-LEUBA: Hold on.
5 MEMBER BROWN: You had to go through 6 reactor engineering, the safeguards people, the 7 materials people, the physics director, me, who did 8 the plant analysis. Plus the fluid system guys to 9 make sure we didn't muck with their analysis.
10 CO-CHAIR MARCH-LEUBA: When you send the 11 topical report for Alex to review it takes 18 months.
12 MEMBER BROWN: Yes. Some of our changes 13 took a year, 18 months. Some of them. They were not 14 done casually, that's all I'm saying.
15 CO-CHAIR MARCH-LEUBA: Yes.
16 MEMBER BROWN: I'm just, just an 17 observation.
18 CO-CHAIR MARCH-LEUBA: All the cause that 19 the industry used are modified, I wouldn't say daily 20 --
21 MEMBER BROWN: I'm not arguing about that.
22 It's using, when you started using them to do your, 23 evaluate your plant when you've used an earlier 24 version for setting everything in the basis of your 25 operations, your AOO responses, et cetera, et cetera, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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273 1 et cetera.
2 CO-CHAIR MARCH-LEUBA: Okay.
3 MEMBER BROWN: I just think that's a 4 potential problem.
5 DR. CORRADINI: I think all we're hearing 6 is that there's going to have to be a change process 7 and that that change process has to be approved.
8 That's what I sense is over there in that backroom.
9 MEMBER BROWN: Their words were used, 10 changes was out, NRC having to see them in terms of --
11 DR. CORRADINI: Okay. That isn't what I 12 heard by --
13 MEMBER BROWN: That's what --
14 DR. CORRADINI: Don't blame her, she's up 15 front. I think that's what Becky said to us.
16 MEMBER BROWN: She's doing the talking so 17 I presume she talks and then she is amplifying.
18 (Laughter.)
19 MS. PATTON: I'm just louder.
20 MEMBER BROWN: She's putting the brakes on 21 she is providing the insight that we needed.
22 MS. PATTON: So we would have to approve 23 the change process.
24 MEMBER BROWN: I understand that.
25 MS. PATTON: Right.
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274 1 MEMBER BROWN: Details matter.
2 (Laughter.)
3 CO-CHAIR KIRCHNER: Keep going, Alex.
4 MEMBER BROWN: Excuse me.
5 MS. SIWY: All right.
6 MEMBER BROWN: Go ahead. Thank you.
7 MS. SIWY: So, condition in Limitation 4 8 is that any credit for the secondary MSIVs, which are 9 not safety related, must be approved through the 10 specific design review.
11 Limitation Condition 5 is that event-12 specific electrical power assumptions, single failures 13 and operator actions must be approved through the 14 specific design review.
15 And finally, Number 6 looks similar to one 16 from the LOCA topical report that the non-LOCA 17 evaluation model use is limited to NRELAP5 Version 1.4 18 and Version 2 of the NPM model, unless changes are 19 made pursuant to the previously described change 20 process.
21 So, in conclusion, all technical issues 22 from the course of the non-LOCA review have been 23 resolved. And the staff finds that the use of NRELAP5 24 with the non-LOCA methodology described in the TR is 25 acceptable for non-LOCA safety analyses of the NuScale NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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275 1 NPM design, subject to the limitations and conditions 2 as summarized on the last slide.
3 CO-CHAIR KIRCHNER: Thank you, Alex.
4 Members, any further questions of Alex?
5 PARTICIPANT: That is all.
6 CO-CHAIR KIRCHNER: Okay. So at this 7 point I would like to use this opportunity for public 8 comment. This finishes are, essentially, open 9 session.
10 So it's time to turn to anyone in the 11 audience. Member of the public in the audience or 12 others.
13 If you would wish to make a comment, 14 please come forward to the microphone. State your 15 name and your comment. And then we will open up our 16 bridge line next.
17 CO-CHAIR MARCH-LEUBA: It's open.
18 CO-CHAIR KIRCHNER: Seeing that no one has 19 stepped forward, is there anyone out there in the 20 public on the bridge line who wishes to make a 21 comment? If so, state your name and please make your 22 comment.
23 MS. FIELDS: This is Sarah Fields. The 24 only comment I have is on the NRC slides. They didn't 25 have a list of acronyms at the end. Having that list NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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276 1 at the end to refer to is helpful.
2 I have a master list that I got from the 3 staff, I don't know, a year or so ago and some of the 4 -- a lot of the acronyms being used were not on that 5 list. So, it's for future slides really helps. Thank 6 you.
7 CO-CHAIR KIRCHNER: Thank you for your 8 comment. Anyone else wish to make a comment?
9 Using the five second rule, not hearing 10 further, okay, we can close the bridge line. And at 11 this point I would like to recess for a break. And 12 then we will re-engage at 3:30 on this clock here and 13 go into closed session. Thank you.
14 (Whereupon, the above-entitled matter went 15 off the record at 3:15 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-0220-68846 February 17, 2020 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 Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on reactor Safeguards (ACRS) NuScale Subcommittee Meeting on February 19, 2020. The materials support NuScales presentation of the Rod Ejection Accident Methodology topical report.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-8H12 Michael Snodderly, NRC, OWFN-8H12 Christopher Brown, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12
Enclosure:
ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0 NuScale Pow er, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0220-68846
Enclosure:
ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0 NuScale Pow er, 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 Topical Report Rod Ejection Accident Methodology February 19, 2020 1
PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presenters Kenny Anderson Nuclear Fuels Analyst Matthew Presson Licensing Project Manager 2
PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Opening Remarks - NuScale T/H Methods NRELAP5 System T/H Analysis Basis code TR-0516-49422-P
- Modified to address NuScale- Containment specific phenomena/systems response analysis Control rod TR-0516-49416-P
Non-LOCA EM
- LOCA EM extended to derive TR-0716-50350-P EMs for other events as shown in this figure. FSAR Ch 5, RAI 9508
- Nuclear Analysis Codes -
TR-0716-50350-P-A Overcooling return to power
- Critical Heat Flux -
TR-0116-21012-P-A TR-0916-51299-P
- Subchannel Analysis - Long term TR-0915-17564-P-A cooling with ECCS 3
PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Agenda
- Event Overview
- Acceptance Criteria
- PCMI Criteria - DG-1327
- Method Flowchart
- Steady State Initialization
- Event Evaluations
- Summary 4
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Overview
- NuScale seeks approval of methodology for modeling rod ejection accident (REA) events
- Bounding reactivity initiated accident (RIA) from General Design Criteria (GDC) 28
- REA is unique in comparison to other Ch. 15 events Description Rod Ejection Other Events Dominant Physics Nuclear Thermal-Hydraulics Timing milli-sec sec to hr Spatially Local Global Peak power ~5x Full Power ~1.2x Full Power Integrated Energy Low Low to High Failure of ASME Class 1 Postulated Cause Single Equipment Failure Pressure Boundary Acceptance Criteria Specialized Generic 5
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Unique Event Acceptance Criteria Topical Criteria Description Unique?
Section Maximum reactor coolant system pressure 5.3 No Hot zero power (HZP) fuel cladding failure 5.5.2 Yes FGR effect on cladding differential pressure N/A Yes Critical heat flux (CHF) fuel cladding failure 5.4.1 No Cladding oxidation-based PCMI failure 5.5.3 Yes Cladding excess hydrogen-based PCMI failure N/A Yes Incipient fuel melting cladding failure 5.5.1 No Peak radial average fuel enthalpy for core cooling 5.5.2 Yes Fuel melting for core cooling 5.5.1 No Fission product inventory (failed fuel census) 5.6 Yes
- Submitted NuScale design and method inherently precludes fuel failure, thus no accident radiological consequences are evaluated.
- PCMI: Pellet-Clad Mechanical Interaction 6
PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Revised PCMI Criteria
- In general, the NuScale REA methodology has adopted the limiting criteria of the Clifford Letter (ML14188C423), now included in draft guide DG-1327 (ML16124A200). In spirit, NuScale is prepared for this regulatory change:
- Closed session presents example results, showing large margins for enthalpy rise
- A technical formality inhibits complete adoption at this time. NuScale does not currently have a validated cladding H2 model to convert local exposure to excess cladding hydrogen
- Oxidation criteria from NUREG-0800 Section 4.2, Appendix B (ML07074000) is used
- To simplify method, no exposure is credited (Limit: 75 cal/gm)
- NuScale M5 cladding less susceptible than other zirc alloy-type clad used in the industry NUREG-800, Sec. 4.2, Fig. B-1 DG-1327, Fig. 2 NuScale criteria NuScale max ratio per corrosion limit.
7 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Unique Event Method (Flowchart)
NRELAP5 Dynamic Peak RCS Pressure System Response Below 120% Limit SIMULATE-3K VIPRE-01 MCHFR Above SIMULATE5 Steady Dynamic Core Subchannel CHF Correlation Design State Initialization Response Evaluation Limit Fuel Temperature Adiabatic Heatup and Enthalpy Below Fuel Response Regulatory Criteria 8
PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Steady-State Initialization
- SIMULATE5: Setup the core response analysis
- Code shown to be appropriate in TR-0616-48793-A (Nuclear Analysis Codes and Methods Qualification)
- Determination of the worst rod stuck out (WRSO)
- Assumption bounds potential for ejected assembly to damage adjacent control rod assembly
- Due to rapid nature of the event, location does not significantly affect the results in NuScale application 9
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Dynamic Core Response
- SIMULATE-3K: Model transient core response
- Benchmarked to SPERT-III experiment and NEACRP computational benchmark
- Benchmarks demonstrate the combined transient neutronic, thermal-hydraulic, and fuel pin modeling capabilities
- SIMULATE-3K results generally in excellent agreement with the results from the two benchmark problems
- Uncertainties applied for each simulation:
- Delayed Neutron Fraction
- Ejected Rod Worth
- Doppler Temperature Coefficient
- Moderator Temperature Coefficient 10 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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CHF Evaluation
- VIPRE-01: Model detailed thermal-hydraulics
- Evaluate critical heat flux (CHF) acceptance criteria
- Code shown to be appropriate in TR-0915-17564-A (Subchannel Analysis Methodology)
- Unique event differences in method:
- Smaller axial nodalization (smaller time steps)
- Radial power distribution (case-specific)
- Axial power distribution (peak assembly)
- Convergence parameters
- Additional parametric sensitivity cases performed with each application to holistically justify differences 11 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Adiabatic Fuel Heatup
- Hand-Calculation: Model fuel response
- Total energy (from SIMULATE-3K) during the transient is integrated
- Conservative as no energy is allowed to leave the fuel rod
- Energy is then converted into either a temperature or enthalpy increase
- Fuel rod geometry, heat capacity, and power peaking factors taken into account
- Calculated values compared to NRC developed acceptance criteria
- Example values provided in closed session 12 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Dynamic System Response I
- NRELAP5: Evaluate system response for input to CHF Evaluation
- Code shown to be appropriate in TR-0516-49416 (Non-LOCA Methodologies)
- Transient power from SIMULATE-3K utilized as input
- No reactivity calculation performed in NRELAP5
- Provides system thermal-hydraulic conditions to subchannel (CHF) evaluation
- System flow, pressure, and inlet temperature
- Screens cases for potential to be limiting
- Family of limiting cases evaluated with VIPRE-01 13 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Dynamic System Response II
- NRELAP5: Evaluate system response for pressurization
- Limiting scenario: Low ejected worth that raises the power quickly to just below both the high power and high power rate trip setpoints
- Point-kinetics model used based on bounding static worth
- Peak system pressure calculated compared to acceptance criteria
- Example results to be presented in closed session 14 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Summary
- A conservative analysis method for the unique rod ejection accident
- Topical Report provides details and justification for:
- Software tools and acceptance criteria used
- Applicability of the method and tools
- Appropriate treatment of uncertainties
- Results from application of the method provide input to FSAR Chapter 15 15 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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- CHF - Critical Heat Flux
- GDC - General Design Criteria
- HZP - Hot Aero Power
- MCHFR - Minimum Critical Heat Flux Ratio
- NEACRP - Nuclear Energy Agency Committee on Reactor Physics
- PCMI - Pellet Clad Mechanical Interaction
- REA - Rod Ejection Accident
- RIA - Reactivity Initiated Accident
- WRSO - Worst Rod Stuck Out 16 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
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 Charlotte Office Corvallis Office 2815 Coliseum Centre Drive, 1100 NE Circle Blvd., Suite 200 Suite 230 Corvallis, OR 97330 Charlotte, NC 28217 541.360.0500 980.349.4804 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301.770.0472 http://www.nuscalepower.com Twitter: @NuScale_Power 17 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
LO-0220-68918 February 17, 2020 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 Topical Report - Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on February 19, 2020. The materials support NuScales presentation of the Loss-of-Coolant Accident Evaluation Model topical report.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Presentation: NuScale Topical Report - Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-8H12 Michael Snodderly, NRC, OWFN-8H12 Christopher Brown, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12
Enclosure:
ACRS Subcommittee Presentation: NuScale Topical Report - Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0 NuScale Pow er, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0220-68918
Enclosure:
ACRS Subcommittee Presentation: NuScale Topical Report - Loss-of-Coolant Accident Evaluation Model, PM-0220-68917, Revision 0 NuScale Pow er, 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 Topical Report Loss-of-Coolant Accident Evaluation Model February 19, 2020 1
PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
Revision: 0 Template #: 0000-21727-F01 R5
Presenters Matthew Presson Licensing Project Manager Dr. Pravin Sawant Supervisor Code Validation and Methods Dr. Selim Kuran Thermal Hydraulic Analyst 2
PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
Revision: 0 Template #: 0000-21727-F01 R5
Agenda
- Methodology Overview
- Background
- Regulatory Requirements
- Methodology Roadmap
- NPM Safety Systems Overview
- Element 1: PIRT
- Element 2: Assessment Base
- Element 3: NRELAP5 Evaluation Model
- Element 4: Applicability Evaluation
- Conclusions 3
PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Background
- Unique NPM Design Features
- Integrated design eliminates piping and limits potential breaks
- Coolant captured completely in containment, cooled and returned to RPV using a large pool as ultimate heat sink
- Simple LOCA Progression with Well-Known Phenomena
- Choked/un-choked flow through break and ECCS valves
- Core decay heat and RCS stored energy release
- CNV heat transfer to pool (condensation, conduction, convection)
- EM Development Approach
- Follows Regulatory Guide 1.203 EMDAP (Table 2-1)
- Compliance with 10 CFR 50.46 and Appendix K requirements (Table 2-2) 4 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Regulatory Requirements
- 10 CFR 50.46 Acceptance Criteria
- Max. clad temperature < 2200 ºF
- Cladding oxidation > 0.17 times thickness
- Hydrogen generation < 0.01 times total hydrogen from oxidation of all cladding
- Core remains amenable to cooling
- Long-term cooling maintained
- Maximum PCT at steady state, no clad heat up
- Core remains covered: collapsed level > TAF
- MCHFR > CHFR Limit (1.29)
- Containment pressure and temperature below design limit 5
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Methodology Roadmap
- 10 CFR 50.46 Appendix K Compliance (Section 2.2.3 of LTR)
Element 1 Establish Requirements for EM Capabilities LOCA PIRT (LTR Ch. 4)
Element 3 Element 2 Develop Evaluation Model Develop Assessment Base NRELAP5 Code (LTR Ch. 6)
SET and IET Assessment (LTR Ch. 7) Evaluation model (LTR Ch. 5)
Break spectrum and sensitivity calcs. (LTR Ch. 9)
Element 4 Assess EM Adequacy Top-down and Bottom-up evaluation (LTR Ch. 8) 6 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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NPM Safety Systems
- Opens a boiling/condensing circulation flow path to transfer decay and residual heat to reactor pool
- Reactor Recirculation Valves (RRV): 2 valves
- Reactor Vent Valves (RVV): 3 valves
- Actuation Signals: High CNV level, 24-hour loss of AC power
- Fail safe: ECCS trip valves open on loss of DC power
- Inadvertent Actuation Block (IAB)
- Prevents inadvertent opening of ECCS valves at high RCS pressure
- Actuation based on differential pressure between RPV and CNV
- Module Protection System (MPS)
- Reactor scram
- Steam Generator (SG) and Containment (CNV) Isolation
- Passive safety system activation (ECCS and DHRS)
- Decay Heat Removal System (DHRS)
- Passive, boiling-condensation system
- Removes heat from RCS through SG via two trains
- Each trains capable of removing 100% decay heat
PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Element 1 PIRT 8
PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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PIRT Process
- Assessment of relative importance of phenomena
- Unique phases
- Key components
- PIRT panel included recognized experts and NuScale subject matter experts
- State-of-knowledge, design description, LOCA description, NRELAP5 calculations
- Figures-of-Merit
- CHF, Collapsed level above top of the active fuel, CNV P & T
- Rankings
- Importance: High, Low, Medium, Inactive
- Knowledge: Well known (small uncertainty), Known (moderate uncertainty, partially known (large uncertainty), very limited 9
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Spatial and Temporal Decomposition
- Phenomena identified for Systems, Structure, Components (SSCs) and LOCA phases o Phase 1a: Blowdown o Phase 1b: ECCS activation (opening)
System/Subsystem/Module decomposition Distinct phases of a typical NPM LOCA 10 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Element 2 Assessment Base 11 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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NRELAP5 Code
- RELAP5-3D© v4.1.3 used as a baseline code o Two-fluid model (thermal and mechanical non-equilibrium) for hydrodynamics with
- Non-condensable gases with gas phase
- Semi-implicit scheme for time integration o Heat conduction across 1D geometries (slab, cylinder, sphere) o Neutron Kinetics with thermal hydraulic feedback o Special Process Models o Comprehensive control/trip system modeling
- Code configuration control and development consistent with NuScales NQA-1 2008 / 2009a QA program
- Modifications for NRELAP5:
- NuScale specific components (e.g., helical coil SG)
- Regulatory requirements (i.e., Appendix K)
- Error correction 12 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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IET and SET Data
- Extensive database with adequate coverage of all high-ranked phenomena
- Integral effects tests (IET)
- Six (6) NIST-1 tests
- Separate effects tests (SET)
- Two (2) NIST-1 SETs
- Four (4) other NuScale SETs
- Nine (9) Legacy SETs 13 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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NIST-1 Facility
- Primary source of NuScale-Specific IET and SET data
- Design Features
- Integral Reactor Vessel with electrically heated rod bundle core, helical coil steam generator, and pressurizer
- Containment with HTP and Cooling Pool
- DHRS, ECCS, CVCS lines represented
- ~700 instruments
- Scaling Basis
- Power/Volume Scaling
- Reduced height and reduced volume scale
- Full Pressure and Temperature
- Same Time Scale (isochronicity) 14 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Element 3 NRELAP5 NPM LOCA 15 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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NPM LOCA Model Overview
- Analysis and Justifications
- NRELAP5 model nodalization and input options
- Time-step control
- Initial and boundary condition biases
- Treatment of setpoints and trips
- LOCA break spectrum
- Break location and sizes
- Single failures
- Power availability
- Methodology sensitivity calculations
- Required by Appendix K
- Phenomena-specific
- To establish conservative biases 16 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Element 4 Applicability Evaluation 17 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Applicability Evaluation
- Evaluated models and correlations (bottom-up)
- Identified dominant models/correlations for H phenomena (Table 8-1 of LTR)
- Identified key model/correlation parameters and phenomenological domain where models/correlations are used (Tables 8-2 and 8-4)
- Reviewed models/correlations (Table 8-18 of LTR)
- Pedigree, Applicability range, Fidelity to SET data, Scalability
- Evaluated integral performance of EM (top-down)
- Reviewed code governing equations and numerics
- Evaluated integral performance of code using IET data (Table 8-19 of LTR)
- Evaluated IET data applicability and NRELAP5 scalability
- Scaling and distortion analysis
- Differences and distortions between NPM and NIST can be accounted using NRELAP5 18 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Conclusions
- Other methodology conservatisms
- Cycle independent bounding LOCA analysis
- Supported by extensive experiment database, well qualified code, and several sensitivity calculations
- Applicability evaluation consistent with RG 1.203
- CHF not challenged
- No clad or fuel heat-up
- CNV P&T remain below design limits 19 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Appendix B to LOCA LTR Extension to IORV Event 20 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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IORV Background
- Liquid space (RRV) and steam space (RVV, RSV) discharge
- Similar transient phenomena and progression
- EM Development Approach
- Compliance with DSRS for NuScale SMR Design 15.6.6
- Follows RG 1.203 EMDAP
- Element 1 (PIRT), Element 2 (Assessment), and Element 4 (Applicability) remains same as LOCA EM
- Initial LOCA PIRT addressed IORV
- Element 3 (NRELAP5 Model) unique due to event classification 21 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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- Minor methodology differences given AOO classification
- Key Acceptance Criteria
- MCHFR Limit ( 1.13 high flow range, 1.37 low flow range)
- Conservatisms same as LOCA with exceptions:
- Fuel properties still biased to maximize stored energy, but additional 15% bias removed
- Limiting axial power shapes and radial peaking based on subchannel analysis
- Moody choked flow model for 2-phase flow choking applied to initiating valve
- Initial conditions biased to minimize MCHFR 22 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Conclusions
- PIRT, Assessment, and Applicability same as LOCA
- Minor methodology differences for AOO classification
- Focused on conservative CHFR evaluation
- Collapsed level in RPV remains above TAF 23 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Acronyms 1-D one-dimensional HP high pressure 3D three-dimensional HS heat sink AC alternating current HTP heat transfer plate ANS American Nuclear Society H2TS hierarchical two-tiered scaling CCFL counter current flow limitation IAB inadvertent actuation block CHF critical heat flux IET integrated effects test CNV containment vessel INL Idaho National Laboratory CVCS chemical and volume control system KATHY Karlstein thermal-hydraulic test facility DC direct current kW kilowatt DCA Design Certification Application LOCA loss-of-coolant accident DHRS decay heat removal system LTR Licensing Topical Report ECCS emergency core cooling system Max maximum EM evaluation model MCHFR minimum critical heat flux ratio EMDAP evaluation model development and Min minimum assessment process Mlb/ft2.hr pounds mass per square foot per hour FW feedwater MPS module protection system FSAR Final Safety Analysis Report MSIV main steam isolation valve FOM figure of merit NIST-1 NuScale Integral System Test Facility HL hot leg NPM NuScale Power Module 24 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Acronyms P&T pressure and temperature PCT peak cladding temperature PIRT phenomena identification and ranking table psi pounds per square inch psia pounds per square inch absolute PZR pressurizer QA Quality Assurance RCS reactor coolant system RG Regulatory Guide RRV reactor recirculation valve RPV reactor pressure vessel RVV reactor vent valve SG steam generator SET separate effects test SIET Societ Informazioni Esperienze Termoidrauliche StDev standard deviation TAF top of active fuel 25 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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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 Charlotte Office Corvallis Office 2815 Coliseum Centre Drive, 1100 NE Circle Blvd., Suite 200 Suite 230 Corvallis, OR 97330 Charlotte, NC 28217 541.360.0500 980.349.4804 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301.770.0472 http://www.nuscalepower.com Twitter: @NuScale_Power 26 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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Backup IORV Slides 27 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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RRV Opening Results Sequence of Events Time (s)
RRV opens 0 Minimum CHFR occurs 0.5 Containment pressure reaches analytical limit 0.7 Control rods begin to fall 2.7 Peak steam generator pressure is reached 64 Remaining ECCS valves open 161 Minimum collapsed liquid level above the core 170 Peak containment pressure is reached 171 Natural circulation from containment to RPV is 197 established 28 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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RVV Opening Results Sequence of Events Time (s)
RVV opens 0 Containment pressure reaches analytical limit 0.3 Minimum CHFR occurs 0.3 Control rods begin to fall 2.3 Peak steam generator pressure is reached 25 Peak containment pressure is reached 52 Remaining ECCS valves open 3925 Natural circulation from containment to RPV is 4072 established Minimum collapsed liquid level above the core 4192 29 PM-0220-68917 Copyright 2020 by NuScale Power, LLC.
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LO-0220-68854 February 17, 2020 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 Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on February 19, 2020. The materials support NuScales presentation of the Non-Loss-of-Coolant Accident Analysis Methodology topical report.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-8H12 Michael Snodderly, NRC, OWFN-8H12 Christopher Brown, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12
Enclosure:
ACRS Subcommittee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0 NuScale Pow er, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0220-68854
Enclosure:
ACRS Subcommittee Presentation: NuScaleTopical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0 NuScale Pow er, 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 Topical Report Non-Loss-of-Coolant Accident February 19, 2020 1
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Presenters Ben Bristol Supervisor, System Thermal Hydraulics Meghan McCloskey Thermal Hydraulics Analyst Matthew Presson Licensing Project Manager Paul Infanger Licensing Specialist 2
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Outline
- Scope of non-LOCA LTR
- Non-LOCA events
- Events and acceptance criteria
- Interface to other methodologies
- Factors controlling margin to acceptance criteria
- Development of non-LOCA EM
- PIRT and gap analysis
- Focus of NRELAP5 validation for non-LOCA
- General event analysis methodology
- Specific event analysis 3
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Scope of Non-LOCA Topical Report In Scope Out of Scope
- NRELAP5 system
- Accident radiological dose
- Interface to subchannel analysis and accident radiological
- Control rod ejection analysis
- LOCA and valve opening
- Short-term transient events progression with DHRS
- Peak containment cooling pressure/temperature analysis
- Long term transient progression with DHRS Riser uncovery Return to power 4
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Non-LOCA EM EM applicable to NuScale Power Module plant design Applicable initiating events:
- Cooldown events
- Reactivity events
- Decrease in FW temperature - Uncontrolled bank withdrawal from subcritical
- Increase in FW flow
- Uncontrolled bank withdrawal at power
- Increase in steam flow Inadvertent opening of SG relief or safety valve - Control rod misoperation
- Steam piping failures (postulated accident)
- Single rod withdrawal
- Loss of containment vacuum Containment flooding
- Control rod drop
- Inadvertent decrease in RCS boron concentration
- Heatup events
- Loss of external load
- Inventory increase event Turbine trip
- Loss of condenser vacuum - CVCS malfunction
- Closure of MSIV
- Loss of non-emergency AC power
- Inventory decrease events
- Loss of normal FW flow
- Feedwater system pipe breaks (postulated accident) - Small line break outside containment (infrequent event)
- Inadvertent operation of DHRS
- Steam generator tube failure (postulated accident)
NuScale unique event 5
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Non-LOCA Event Acceptance Criteria AOO Infrequent Event Accident Description Analysis Acceptance Criteria Acceptance Criteria Acceptance Criteria Reactor Coolant Non-LOCA System Pressure 110% of Design 120% of Design 120% of Design NRELAP5 (Pdesign = 2100 psia)
Steam Generator Non-LOCA Pressure 110% of Design 120% of Design 120% of Design NRELAP5 (Pdesign = 2100 psia)
Minimum If limit exceed, If limit exceed,
> Limit Subchannel Critical Heat Flux Ratio fuel assumed failed (1) fuel assumed failed (1)
Maximum Fuel If limit exceed, If limit exceed,
< Limit Subchannel Centerline Temperature fuel assumed failed (1) fuel assumed failed (1)
< Limits < Limits < Limits Containment Containment Integrity (pressure, temperature) (pressure, temperature) (pressure, temperature) P/T analysis Escalation of an AOO to an accident (AOO)
If other or No No No acceptance Consequential loss of criteria are met system functionality (IE or accident)?
Normal or Normal Radiological Dose < Limit < Limit Accident Operations radiological (1) NuScale safety analysis methodologies developed to demonstrate fuel cladding integrity maintained.
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Evaluation Models - General Non-LOCA Approach Plant design, M&E releases Core design, NRELAP5 VIPRE-01 from T/H Fuel rod design, system T/H subchannel response, other Plant initial conditions, response analysis input SSC performance RCS pressure, Accident secondary Fuel cladding radiological pressure, integrity analysis Safe stabilized condition Radiological Non-LOCA topical report Subchannel topical report TR-0516-49416-P TR-0915-17564-P-A dose acceptance criteria Accident source term topical report TR-0915-17565-P 7
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Non-LOCA Events -
Margin to Acceptance Criteria Design characteristics governing non-LOCA event transient response and margin to acceptance criteria
- MCHFR: Limited by combination of high power, high pressure, high temperature conditions occurring around time of reactor trip, for reactivity insertion events
- Primary pressure: Protected by RSV lift
- Secondary side pressure: Limited by primary side temperature conditions
- Radiological release: MPS designed to rapidly detect and isolate based on measured conditions
- Establishing a safe, stable condition: MPS designed to trip, actuate DHRS to protect adequate inventory in at least 1 steam generator 8
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Non-LOCA EM Development
- Non-LOCA evaluation model developed to perform conservative analyses, following intent of the RG 1.203 EMDAP and applying a graded approach
- Element 1 - Establish applicable transients and acceptance criteria, develop non-LOCA PIRT
- Element 2, 3, 4
- Leverage NRELAP5 development, NRELAP5 assessments performed during LOCA evaluation model development.
- Gap analysis performed to evaluate how high ranked phenomena are addressed
- Focused on differences in high ranked PIRT phenomena between LOCA and non-LOCA
- Additional NRELAP5 code validation performed focused on DHRS and integral non-LOCA response
- Suitably conservative initial and boundary conditions applied for non-LOCA analyses
- Sensitivity calculations used to demonstrate factors controlling margin to acceptance criteria 9
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Non-LOCA PIRT Development Event Types SSCs Considered in PIRT Increased heat removal Reactor coolant system Main feedwater system Decreased heat removal Containment vessel Main steam system Reactivity anomaly Decay heat removal system Chemical volume control system Increase in RCS inventory Reactor pool Containment evacuation Steam generator tube failure system Phase Identification RCS Response DHRS Operation
- PIRT Figures of merit 1 pre-trip transient higher flow levels at full inactive CHFR power levels RCS pressure 2 post-trip transitional flow levels at startup CHFR transition transitioned power levels RCS, secondary, containment pressures 3 stable natural lower flow levels at decay fully effective CHFR circulation power levels RCS mixture level Subcriticality
- If DHRS actuated by protection system
- Different non-LOCA events involve different plant systems and responses
- PIRT developed considering all non-LOCA event types and important SSCs
- Short-term response divided into 3 generic phases with associated FoM 10 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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NRELAP5 Applicability for Non-LOCA After non-LOCA PIRT developed, Key areas identified from gap analysis for short-term non-LOCA analysis:
gap analysis performed to
- DHRS modeling and heat transfer determine how to address high-NRELAP5 validation against KAIST tests; ranked phenomena: NIST-1 SETs HP-03, HP-04 NPM sensitivity calculations
- Validation performed as part of NRELAP5 assessment for LOCA
- Steam generator modeling and heat transfer evaluation model NRELAP5 validation against
- Additional validation or benchmark SIET-TF1, SIET-TF2 tests for non-LOCA NPM sensitivity calculations
- Reactivity event response
- Conservative input NRELAP5 benchmark against RETRAN-3D
- Subchannel analysis
- NPM non-LOCA integral response NRELAP5 validation against NIST-1 IETs NLT-2a, NLT-2b, NLT-15p2 Overall conclusion: NRELAP5 code, with NPM system model, is applicable for calculation of the NPM non-LOCA system response 11 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Non-LOCA Analysis Process Topical report Section 4
- 1. Develop plant base model 5. Identify cases for subchannel NRELAP5 input (geometry, control analysis and extract boundary and protection systems, etc) conditions (if applicable)
- 2. Adapt NRELAP5 base model as Conservative bias directions:
necessary for specific event
- Maximum reactor power analysis and desired initial
- Maximum core exit pressure conditions
- Maximum core inlet temperature
- 3. Perform steady state and transient
- Minimum RCS flow rate analysis calculations with NRELAP5 CHF calculations for NRELAP5 dummy hot rod may be used as a screening tool to assist analysts in
- 4. Evaluate results of transient determining limiting cases to be analysis calculations: evaluated in downstream subchannel Confirm margin to maximum RCS analysis pressure acceptance criterion
- 6. Identify cases for radiological Confirm margin to maximum SG pressure analysis (if applicable) acceptance criterion Maximum mass release case Confirm appropriate transient run time execution to demonstrate safe, stabilized Maixmum iodine spiking case condition achieved 12 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Non-LOCA Methodology General Methodology Event-specific Methodology (Section 7.1): (Section 7.2)
- Steady-state conditions
- Description of event initiation
- Treatment of plant controls and progression
- Acceptance criteria of interest
- Loss of power
- Single failure
- Limiting single failure, loss of power scenarios, or need for
- Bounding reactivity sensitivity calculations parameter input
- Initial condition biases and
- Biasing of other parameters: conservatisms, or need for initial conditions, valve sensitivity calculations characteristics, analytical
- Tabulated representative results limits and response times of sensitivity calculations
- Operator action Example analysis results provided in Section 8 13 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Conclusions
- Non-LOCA system transient evaluation model developed following a graded approach in accordance with guidance provided in RG 1.203
- Applies to NPM-type plant design natural circulation water reactor with helical coil SG and integral pressurizer
- NRELAP5 used to simulate the system thermal-hydraulic response
- Demonstrate primary and secondary pressure acceptance criteria are met
- Demonstrate safe, stabilized condition achieved
- System transient results provide boundary conditions to downstream subchannel and radiological analyses 14 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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- AOO - Anticipated Operational
- MPS - Module Protection System Occurrences
- CNV - Containment Vessel
- MSS - Main Steam System
- CVCS - Chemical and Volume Control System
- NIST NuScale Integral System Test-1
- NPM - NuScale Power Module
- DHRS - Decay Heat Removal System
- PIRT - Phenomena Identification and Ranking Table
- EM - Evaluation Model
- EMDAP - Evaluation Model Development and Assessment Process
- RSV - Reactor Safety Valve
- RVV - Reactor Vent Valve
- SET - Separate Effects Test
- IET - Integral Effects Test
- KAIST - Korea Advanced Institute of
- SG - Steam Generator Science and Technology
- SSC - Structures, Systems, and
- LOCA - Loss of Coolant Accident Components
- MCHFR - Minimum Critical Heat Flux Ratio 15 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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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 Charlotte Office Corvallis Office 2815 Coliseum Centre Drive, 1100 NE Circle Blvd., Suite 200 Suite 230 Corvallis, OR 97330 Charlotte, NC 28217 541.360.0500 980.349.4804 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301.770.0472 http://www.nuscalepower.com Twitter: @NuScale_Power 16 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Additional Material for Public Presentation Previously presented background material 17 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Power Module Overview Integral Pressurized Water Reactor steam line
- Core, steam generator and pressurizer in one vessel feedwater line
- Integrated reactor design, no large-break containment vessel loss-of-coolant accidents
- Reactor coolant system operated in single reactor vessel phase (liquid) density driven flow pressurizer
- Safety decay heat removal systems are passive and fail safe steam generator
- Module protection system designed to automate event mitigation actuations (no operator actions) feedwater header reactor core module support skirt 18 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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ECCS Emergency Core Cooling System
- ECCS valves open to a boiling/condensing circulation flow path to transfer decay and residual heat to reactor pool
- Liquid from containment vessel enters RCS through reactor recirculation reactor vent valve reactor vent valves valves
- Vapor vented from RCS to containment vessel through reactor vent valves
- Steam condenses on inside surface of containment vessel
- Heat transfer through vessel walls to the reactor pool
- Actuation Signals: High CNV reactor recirculation valve reactor recirculation valve level, 24hr loss of AC power
- Fail safe: ECCS valves open on loss of DC power 19 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Decay Heat Removal System (DHRS)
- Boiling/condensing loop DHR actuation valves
- Two redundant trains reactor pool
- Redundant actuation and isolation valves for each train DHR passive condenser
- Initiates on:
Loss of power Loss of cooling indication (ESFAS Signal) 20 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Deterministic Event Mitigation Module Protection Functions Event Mitigation Reactivity Control Increase in heat removal transients
- Reactor trip Secondary Isolation
- CVCS/Demineralized Water Decrease in heat removal transients Isolation
- Reactor trip DHRS Actuation RCS and Secondary Inventory Reactivity and power transients Control
- Reactor trip Demineralized Water Isolation
- Containment Isolation Increase in RCS inventory transients
- Secondary Isolation
- Reactor trip CVCS Isolation Heat Removal Decrease in RCS inventory transients
- DHRS Actuation
- Reactor trip CNV Isolation
- Reactor trip 21 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Pressure vs. Temperature Operation Map High Temperature Analytical Limit (610°F) 2100 High Pressure Analytical Limit (2000 psia) 2000 High Pressure Operating Limit (1920 psia) 1900 Minimum Critical Temperature (420°F)
Low Initial TAvg (535F) High Initial TAvg (555F)
Pressurizer Pressure (psia)
Normal Operating Pressure (1850 psia)
TCold TAvg THot 1800 Low Normal Pressure (1780 psia)
Low Pressure Analytical 1700 Limit (1720 psia)
Saturation Curve 1600 Low Low Pressure Analytical Limit (1600 psia)
Subcooled Margin 5°F 1500 400 450 500 550 600 650 Reactor Coolant System Temperature (°F)
Module protection system (Ch. 7, red) Technical specification LCOs (Ch. 16, blue) 22 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Loss of Power - Non-LOCA Event
- Availability of AC, DC power affects whether ECCS valves actuate, and what time they open 23 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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5.1 - Summary Description 620 RCS temperature and flow data 580 Temperature (°F) 540 500 460 T_Hot T_Cold T_Ave T_Cold for Maximum Flow T_Cold for Minimum Flow 420 0 20 40 60 80 100 Reactor Power (%)
700 600 500 Primary Flow (kg/s) 400 300 200 100 Primary Flow Minimum Flow Maximum Flow 0
0 20 40 60 80 100 Reactor Power (%)
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Presentation to the ACRS Subcommittee Staff Review of NuScale Topical Report TR-0716-50350, REVISION 1 Rod Ejection Accident Methodology Presenters:
Chris Van Wert - Senior Reactor Systems Engineer, Office of New Reactors Jeff Schmidt- Senior Reactor Systems Engineer, Office of New Reactors February 19, 2020 (Open Session)
Non-Proprietary 1
NRC Technical Review Areas/Contributors NUCLEAR METHODS, SYSTEMS & NEW REACTORS BRANCH / NRR:
Rebecca Patton (BC)
ADVANCED REACTOR TECHNICAL BRANCH / NRR:
Jeff Schmidt Chris Van Wert 2 Non-Proprietary
Staff Review Timeline TR-0716-50350, ROD EJECTION ACCIDENT METHODOLOGY NuScale submitted Topical Report (TR)-0716-50350, Rod Ejection Accident Methodology, Revision 1, on November 15, 2019, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19319C684).
Staff plans to brief advisory committee on reactor safeguards (ACRS) full committee on March 5, 2020.
Staff plans to issue its final SER in March 2020.
Staff plans to publish the -A (approved) version of the TR prior to finishing Phase 6 of the NuScale DCA.
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Staff Review
- The staffs review included:
- Evaluation of the analysis criteria
- Evaluation of the code suite used within the analysis methodology
- Evaluation of the plant and cycle assumptions used in the analysis methodology
- Evaluation of the rod ejection accident analysis methodology
- The staffs review does not include the licensing basis RIA analysis for the NuScale DCA.
- Contained in Section 15.4.8 of the SER for the NuScale Design Certification
- During its review, staff audited calculations and other supporting information 4 Non-Proprietary
Analysis Criteria
- Reactor Coolant System Pressure
- Maximum RCS pressure 120% of the design pressure
- Fuel Cladding Failure
- High temperature at zero-power conditions:
- Set to 100 cal/g for all cladding differential pressures
- For intermediate and full-power conditions:
- Failure assumed if local heat flux exceeds CHF limits
- PCMI failure threshold follows Figure B-1 of SRP Section 4.2 Appendix B
- Failure is assumed if fuel temperature anywhere in the pellet exceeds incipient fuel melting conditions 5 Non-Proprietary
Analysis Criteria (cont)
- Core Coolability
- Peak radial average fuel enthalpy will remain below 230 cal/g
- Fuel melt is not allowed
- Fission Product
- NuScale does not rely on SRP guidance regarding fission product inventory. Instead, the methodology precludes this need by:
- Requiring that no failure occurs due to fuel melt, enthalpy increase, or exceeding MCHFR limits
- Incorporating the most limiting criteria which SRP 4.2 Appendix A associated with the largest cladding differential pressure, therefore any increase in pressure due to FGR will not change allowed peak radial average fuel enthalpy
- Staff reviewed criteria and found them acceptable relative to the guidance in SRP Section 4.2, Appendix B.
- Staff also noted that DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents is currently being developed.
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Comparison of Current versus Draft Regulatory Guidance 7 Non-Proprietary
Evaluation of Code Suite
- The NuScale REA analysis is based on the following codes and packages:
- CASMO5/SIMULATE5: provides reactor core physics parameters
- SIMULATE-3K: 3-dimensional nodal reactor kinetics code which supplies power input to downstream analyses
- NRELAP5: transient system response
- VIPRE-01: subchannel analysis
- Applicability of CASMO5, SIMULATE5, NRELAP5, and VIPRE-01 has been reviewed and approved for NuScale in TR-0616-48793-P-A, Revision 1, Nuclear Analysis Codes and Methods Qualification.
- The validation of SIMULATE-3K is included as part of TR-0716-50350 and is therefore included in the staffs review.
- Staff concluded that NuScale successfully validated S3K against experimental data and the NEACRP control rod ejection problem computational benchmark 8 Non-Proprietary
Plant and Cycle Assumptions
- The staff reviewed the plant and cycle assumptions used in the NuScale rod ejection analysis methodology
- The staff determined that the methodology included ranges in power, time in cycle, and core power that covered a wide range of operating conditions and would capture the most limiting condition
- The staff agreed that the assumptions associated with the automatic system response of non-safety systems were conservative
- The staff determined that the methodology regarding the timing of loss of AC power conservatively biases the RCS pressure evaluation 9 Non-Proprietary
Rod Ejection Accident Analysis Methodology
- The staff reviewed the analysis methodology including steady-state initialization, dynamic core response, dynamic system response, subchannel critical heat flux evaluation, and the adiabatic heatup fuel response
- The staffs review included the methodology by which information is passed between codes, application of uncertainties, modelling assumptions used for inputs, and handling of reactor trips.
- The staff concluded that the methodology for calculating the system response, subchannel, and fuel response analyses was conservative and acceptable for demonstrating compliance with the acceptance criteria 10 Non-Proprietary
Staff SER Conclusions
- The staff concludes that the NuScale criteria used for evaluating REA either follows or is more conservative than staff guidance
- The staff concludes that the methodology accounts for the various potential operating conditions and time in life, and conservatively addresses uncertainties and plant conditions
- The staff finds the use of TR-0716-50350-P acceptable for evaluating reactivity initiated accidents for the NuScale plant design.
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Questions?
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Review of TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology NuScale Design Certification Application ACRS Subcommittee Meeting February 19, 2020 Non-Proprietary
NRC Staff Review Team
- NRC Technical Reviewers:
- Jeff Schmidt, NRR
- Alex Siwy, NRR
- Ray Skarda, RES
- Peter Lien, RES
- Ron Harrington, RES
- Consultants (Energy Research, Inc.):
- Mohsen Khatib-Rahbar
- Walter Tauche (subcontractor)
- Morgan Libby
- Michael Zavisca February 19, 2020 Non-Proprietary 2
Review Process Overview
- SER is based on TR-0516-49416, Revision 2
- Two audits conducted in four phases
- About 140 audit issues
- Helped to confirm staffs understanding and inform RAIs
- 33 RAI questions issued
- All resolved and responses incorporated into TR-0516-49416, Revision 2, as appropriate February 19, 2020 Non-Proprietary 3
Regulatory Requirements and Guidance
- Regulatory requirements include several GDC (5, 10, 13, 15, 17, 20, 25, 26, 27, 28, 31, 34)
- Guidance Documents
- SRP Section 15.0.2, Review of Transient and Accident Analysis Methods
- RG 1.203, Transient and Accident Analysis Methods
- DSRS Section 15.0,Introduction - Transient and Accident Analyses
- Event-specific SRP/DSRS sections February 19, 2020 Non-Proprietary 4
DSRS Section 15.0 Acceptance Criteria
- Primary and secondary pressures < 110% of design values
- Minimum DNBR > 95/95 limit
- Should not generate a postulated accident without other faults occurring independently or result in a consequential loss of function of the RCS
- Postulated Accidents/Infrequent Events
- Primary and secondary pressures < acceptable design limits, considering potential brittle and ductile failures
- Fuel assumed to be failed if minimum DNBR 95/95 limit
- Offsite doses guidelines of 10 CFR Parts 52.47(a)(2)(iv) and 100 (for infrequent events, 10 percent of 10 CFR 52.47(a) and Part 100)
- Shall not, by itself, cause a consequential loss of required functions of systems needed to cope with the fault, including RCS and containment February 19, 2020 Non-Proprietary 5
Non-LOCA Methodology Scope
- Provides a methodology for performing system transient analysis of specified non-LOCA design-basis events for the NPM
- Directly evaluates primary and secondary pressure figures of merit
- Provides input to downstream analyses that evaluate margin to SAFDLs and radiological consequences
- Covers time frame during which mixture level is above top of riser and natural circulation is maintained
- Includes certain event-specific assumptions and conservative bias directions for initial conditions February 19, 2020 Non-Proprietary 6
Non-LOCA Methodology Analyzed Events Increase in heat removal by the secondary system Reactivity and power distribution anomalies
- Decrease in feedwater temperature
- Uncontrolled control rod assembly bank withdrawal
- Increase in feedwater flow from a subcritical or low power startup condition
- Increase in steam flow
- Uncontrolled control rod assembly bank withdrawal at
- Inadvertent opening of SG relief or safety valve power
- Steam system piping failure inside or outside
- Control rod misoperation (single rod withdrawal, single containment rod or bank drop)
- Containment flooding/loss of containment vacuum
- Inadvertent decrease in boron concentration in the reactor coolant
- Inadvertent DHRS actuation Increase in reactor coolant inventory Decrease in heat removal by the secondary system
- Chemical and volume control system malfunction that
- Loss of external load increases reactor coolant system inventory
- Loss of condenser vacuum Decrease in reactor coolant inventory
- Main steam isolation valve closure
- Failure of small lines carrying primary coolant outside
- Loss of nonemergency AC power to station containment auxiliaries
- Steam generator tube failure
- Loss of normal feedwater flow
- Inadvertent DHRS actuation
- Feedwater system pipe break inside or outside of containment February 19, 2020 Non-Proprietary 7
Non-Review Scope Items Several items described in the Non-LOCA TR are outside the scope of the staff TR review and within the scope of a design-specific application of the methodology:
- Representative calculation results
- Example values of parameters or parameter biases
- Limiting loss of power and single failure scenarios
- Need for operator actions
- Credit for secondary MSIVs that are not safety related February 19, 2020 Non-Proprietary 8
Non-LOCA Transient Analysis Process Upstream Methodologies
- Adapt NRELAP5 Basemodel
- Nuclear Design
- Fuel Design for Specific Event Analysis
- Perform NRELAP5 Steady-State Initialization Calculations
- Evaluate Results of Transient Calculations against Acceptance Criteria Downstream Methodologies
- Identify Cases for Downstream
- Subchannel (VIPRE-01) Analyses
- Accident Source Term February 19, 2020 Non-Proprietary 9
Key Design Features for Non-LOCA
- Staff focused its review on several key features of the NuScale Design:
- Natural circulation design
- Helical coil SGs
- Passive DHRS condensers
- Transfer decay heat to reactor pool using the SGs
- Evacuated containment vessel February 19, 2020 Non-Proprietary 10
Applicability of NRELAP5 to Non-LOCA Analysis
- Applicants non-LOCA PIRT identified highly ranked phenomena
- Applicant identified deltas in highly ranked phenomena relative to LOCA and addressed them using:
- Separate effects tests: NIST HP-03, HP-04
- Staff also evaluated KAIST and SIET assessments
- Integral effects tests: NIST NLT-02a, NLT-02b, NLT-15p2
- Code-to-code benchmark against RETRAN-3D for reactivity-initiated events
- Use of bounding input values
- Other analysis methodologies (e.g., subchannel)
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Significant Issue - Multi-Dimensional Flow Effects
- Staff requested additional justification for how multi-dimensional flow effects in the RCS and thermal stratification in the reactor pool are addressed
- RAI response, as confirmed by audit and multiple audit discussions, resolved the issue
- More details in closed session February 19, 2020 Non-Proprietary 12
NRELAP5 Assessments:
KAIST, NIST-1 HP-03, and HP-04
- Phenomena addressed: condensation inside and heat transfer across DHRS tubes to CPV fluid
- Staff confirmed that the ranges of the KAIST and NIST-1 validations adequately cover expected ranges during DHRS operation in the NPM
- Staff finds the facility nodalization acceptable relative to NPM nodalization
- Staff finds that the NRELAP5 prediction of KAIST and HP-03 data is reasonable to excellent, and the prediction of HP-04 data is reasonable February 19, 2020 Non-Proprietary 13
Significant Issue - Removal of DHRS Heat Transfer Bias
- The applicant performed extensive sensitivity studies that showed an insignificant effect of DHRS performance and nodalization on non-LOCA figures of merit
- Underlying reasons:
- Minimum margin to acceptance criteria typically occurs before DHRS is fully active
- Peak RCS pressure limited by RSV
- Peak secondary pressure a function of secondary inventory and primary side conditions - insignificantly affected by DHRS heat transfer variations
- The staff finds that the sensitivity studies and physical dynamics support removal of a DHRS heat transfer bias for NPM model Revision 2 February 19, 2020 Non-Proprietary 14
NRELAP5 Assessment:
SIET TF-1 and TF-2
- TF-1 assessed flow inside SG tubes
- TF-2 assessed primary-to-secondary heat transfer
- Staff finds reasonable to excellent agreement between NRELAP5 predictions and SIET test data
- Staff raised concerns about the range of conditions validated by TF-2 compared to expected ranges in the NPM and associated uncertainties in SG heat transfer February 19, 2020 Non-Proprietary 15
Significant Issue - Removal of SG Heat Transfer Bias
- The applicant performed comprehensive sensitivity studies that showed an insignificant effect of SG heat transfer on non-LOCA figures of merit
- Staff audited the sensitivity calculations
- The staff finds that these sensitivity studies, in addition to post-trip and normal operation considerations, support removal of SG heat transfer bias for NPM model Revision 2
- More details in closed session February 19, 2020 Non-Proprietary 16
NRELAP5 Assessments:
NIST-1 Integral Effects Tests
- Staff audited detailed assessments
- NLT-02a, loss of feedwater transient to the point of a reactor trip
- Reasonable to excellent code-to-data agreement
- NLT-02b, DHRS-driven cooling following initial DHRS actuation
- Reasonable or better agreement
- Test assessed by NuScale to address staff questions
- Reasonable agreement
- The staff finds that these assessments together demonstrate applicability of NRELAP5 to evaluate non-LOCA transients February 19, 2020 Non-Proprietary 17
NRELAP5 Assessments:
Benchmark against RETRAN-3D
- Primary purpose: assess performance of NRELAP5 point kinetics model
- Applicant made appropriate modeling simplifications since RETRAN-3D does not account for helical coil SG
- Staff reviewed applicants comparison of four reactivity transients:
- Slow uncontrolled control rod withdrawal
- Fast uncontrolled control rod withdrawal
- Power reduction from 100 to 50 percent
- Dropped control rod assembly
- Staff finds that there is reasonable or better agreement for these transients, providing confidence in NRELAP5 point kinetics model February 19, 2020 Non-Proprietary 18
General Non-LOCA Methodology
- Use of suitably conservative input in lieu of uncertainty evaluation, e.g.:
- Bounding reactivity parameters
- Biased initial conditions
- Valve characteristics
- Plant control systems operation, single failure, and loss of power assumptions should maximize consequences with regard to acceptance criteria of interest
- If acceptance criteria are insensitive to certain inputs/assumptions, nominal values may be used
- If an acceptance criterion is not challenged by a particular event, it will not be subjected to in-depth sensitivity studies
- The staff finds the general approach acceptable subject to applicable Limitations and Conditions February 19, 2020 Non-Proprietary 19
Event-Specific Non-LOCA Methodology
- The staff evaluated each event-specific methodology, including:
- Methodology aspects specific to the initiating event
- General assessment of challenge to the acceptance criteria
- Initial conditions, biases, and conservatisms
- Always acceptable for a parameter to be varied as part of a plant-specific calculation
- Staff ensured that bias directions or conservatisms specified are appropriately conservative with respect to effect on acceptance criteria
- The staff finds that the event-specific methodologies will ensure conservative results when implemented February 19, 2020 Non-Proprietary 20
Representative Calculations
- Staff reviewed the representative calculations insofar as they demonstrate application of the non-LOCA methodology
- Staff concludes that the calculations show how the non-LOCA methodology can be used for conservative transient analysis simulations
- Final calculations for the NuScale design are reviewed in the SER for DCA Chapter 15 February 19, 2020 Non-Proprietary 21
Staff SER Limitation and Condition Summary
- 2. Non-LOCA EM scope limited to non-LOCA events defined in the TR prior to the time of riser uncovery for evaluation of primary and secondary pressures and potential for loss of system functionality
- 3. Additional justification must be provided for elimination of SG and DHRS heat transfer biases if applying methodology to a design other than NPM model Revision 2 or a model update made pursuant to a change process specifically approved by NRC for changes to the NPM model
- 4. Any credit for secondary MSIVs (not safety-related) must be approved through design review
- 5. Event-specific electrical power assumptions, single failures, and operator actions must be approved through design review
- 6. Non-LOCA EM use limited to NRELAP5 v1.4 and NPM model Revision 2, unless changes are made pursuant to a change process specifically approved by the NRC staff for changes to NRELAP5 and the NPM model February 19, 2020 Non-Proprietary 22
Conclusions
- All technical issues from the course of the review have been resolved
- Use of NRELAP5 with the non-LOCA methodology described in the TR is acceptable for the non-LOCA safety analyses of the NuScale NPM design subject to the specified limitations and conditions February 19, 2020 Non-Proprietary 23
Backup Slides February 19, 2020 Non-Proprietary 24
NuScale DCA GTS 3.4.1 February 19, 2020 Non-Proprietary 25
NuScale DCA GTS 3.4.1 (2)
February 19, 2020 Non-Proprietary 26
Loss-of-Coolant Accident (LOCA) Evaluation Model Topical Report Review (TR-0516-49422-P)
NuScale Design Certification Application ACRS Subcommittee Meeting February 19, 2020 February 19-20, 2020 Non-Proprietary 1
Review Team
- NRC Mr. Carl Thurston Dr. Shanlai Lu Dr. Peter Lien Mr. Antonio Barrett Dr. Weidong Wang Dr. Tim Drzewiecki Mr. Ron Harrington Dr. Syed Haider
- NuMark Associates Mr. Marvin Smith Dr. Donald Rowe Dr. Leonard Ward Mr. Bert Dunn
- Brook Haven National Lab Dr. Upendra Rohatgi February 19-20, 2020 Non-Proprietary 2
Design Features And Scope 3 Reactor Vent Valves 2 Reactor Recirculation Valves Each ECCS valve has its own IAB, trip valve and trip reset valve Containment functions as part of ECCS February 19-20, 2020 Non-Proprietary 3
Topical Report Application Scope
- Provide a methodology to analyze NuScale LOCA scenarios
- Provide a methodology to analyze the inadvertent opening of RPV Valves event
- Support other methodologies:
Peak containment pressure analysis Non-LOCA analysis Long-Term Cooling analysis February 19-20, 2020 Non-Proprietary 4
Review Approaches
- Early Engagement And Extensive Audits Through Electronic Reading Room Pre-application engagement Initial on-site visits and audit meetings Two phases of continuing audits throughout review period
- Issues Raised:
45 RAI Questions 210 Audit Issues
- Staff performed sensitivity analysis with NRELAP5 and confirmatory analysis with TRACE
- Major Issues Summarized on Later Slides February 19-20, 2020 Non-Proprietary 5
Regulatory Basis and Acceptance Criteria
- Regulatory Basis For LOCA Analysis: 10CFR50.46 Appendix K
- Peak Cladding Temperature < 2,200 ºF
- Local maximum fuel cladding oxidation < 17%
- Whole core fuel cladding oxidation < 1%
- Phenomenon Excluded:
- Post CHF heat transfer, Reflood & Refill, Pump Model
- With this NPM design, LOCA Peak Clad Temperature (PCT) is the fuel temperature during normal plant operation and PCT is not a FOM
- The FOMs for the NuScale NPM during a LOCA are:
- Maintaining containment integrity to provide water flow back to the RPV after ECCS valve operation and pressure equalization between the PRV and CNV
- Maintaining the fuel heat flux below the critical heat flux ratio (CHFR) to prevent significant clad heat-up
- Keeping the RPV collapsed liquid level (CLL) above the core is to demonstrate that there is an adequate supply of liquid water available to the core February 19-20, 2020 Non-Proprietary 6
Regulatory Basis and Acceptance Criteria
- NuScale categorized as Anticipated Operational Occurrence (AOO)
- GDC 10 requires protection of Specified Acceptable Fuel Design Limits (SAFDLs)
- The FOM for the NuScale NPM during an IORV event is:
- Maintain margin to critical heat flux: CHFR > CHFR-limit February 19-20, 2020 Non-Proprietary 7
Phenomena Identification and Ranking Table
- NuScale engaged expert panels to develop a Phenomena Identification and Ranking Table (PIRT) to identify important phenomena and processes in the NPM during a LOCA
- The PIRT panels identified twenty-one phenomena as important to capture in the NuScale LOCA Evaluation Model
- NuScale then performed NRELAP5 code and model validation to show that their NRELAP5 code and LOCA EM is able to evaluate these important phenomena and processes
- The NuScale validation suite included legacy separate effects tests (SETs) and integral effects tests (IETs), and SETs and IETs developed and run specifically for the NuScale NPM design.
February 19-20, 2020 Non-Proprietary 8
Phenomena Identification and Ranking Table
- Staff identified the following significant issues:
- RRV flange not considered a break location
- Not clear whether all the phenomena were identified for the lowest discharge point of the system (RVV or RRV opening).
- NuScale performed additional NIST tests with RRV opening
- Confirmed that the PIRT is appropriate for IORV discharge February 19-20, 2020 Non-Proprietary 9
NRELAP5 Code and Models
- Models and correlations used by NRELAP5 (V 4.1.3) were reviewed and modified as required for use within the NuScale LOCA EM Helical coil S. Generator heat transfer and pressure drop Core CHF, MCHFR limit, and run termination (mCHFR<1)
Steam condensation heat transfer { }
Critical flow break discharge { }
Interfacial drag for large diameter pipes
- The development of NPM model and nodalization were reviewed resulting in model update February 19-20, 2020 Non-Proprietary 10
NIST Tests And Scaling Analysis
- NIST-1 scaled facility test used in validation of NRELAP5 code
- Main components included RPV, helical coil SG system with DHRS, CNV, and CPV representing the reactor pool (separated vessels)
- NIST-1 ECCS connects the RPV to the CNV and consists of two RVVs and two RRVs, each on separate lines
- The scaling analysis was developed and reviewed.
- A RAI was issued to question the validity of all half power test results.
- A new limiting LOCA test was conducted - Full decay heat power test (HP-06B)
February 19-20, 2020 Non-Proprietary 11
NRC Sensitivity & Confirmatory Analyses
- Separate Effect Tests (SETs):
- KAIST model: DHRS tube condensation experiment, non-LOCA
- SIET model: helical coil steam generator tube/shell side heat transfer, non-LOCA
- NIST-1 model: high pressure condensation test (HP-02)
- Integral Effect Tests (IETs):
- NIST-1 models: loss of coolant accident (LOCA) and inadvertent emergency core cooling system (ECCS) operations NPM models: licensing calculation confirmation and sensitivity studies, LOCA, non-LOCA
- Both TRACE and NRELAP5 codes were used. More than fifty five sets of calculations were performed. RAIs were issued and NRELAP5 code was updated from V1.3 to V1.4. Good agreements were obtained with NuScale analysis results February 19-20, 2020 Non-Proprietary 12
CHF Correlations For LOCA and IORV CHF Models:
} and { } CHF correlations are used to calculate CHF.
- { } is used for high-flow conditions
- { } is used for low-flow conditions
- STERN and KATHY test facilities provide CHF databases which are used to assess the NRELAP5 models ability to predict CHF.
- Stern CHF database is used to determine CHFR limit for low-flow conditions for LOCA and LOCA-like events
- Preliminary Prototype Fuel
- KATHY CHF database is used to determine CHFR limit for high-flow conditions for LOCA and LOCA-like events
- Prototypical fuel February 19-20, 2020 Non-Proprietary 13
Approval Limitations & Conditions With all the issues resolved, the staff approved this topical report with the following limitations and conditions:
- Exemption required for an applicant or licensee to exclude required features of Appendix K
- NuScale EM is limited to the evaluation of LOCAs where CHF is not exceeded, the collapsed water level above the core and containment temperature and pressure remain below the design limits
- No credit for DHRS heat removal unless it is further reviewed and approved
- Use of the LOCA EM is limited to evaluations of the analyses for the FOMs described in the TR: the short term LOCA or an IORV event
- Limited to NRELAP5 v1.4 and NPM model Revision 2, unless changes are made pursuant to a change process specifically approved by the NRC staff for changes to NRELAP5 and the NPM model
- Single Failures, Electrical Power Assumptions (ac/dc) and Need for Operator Actions Not Approved Within this Methodology - (done in design review)
- Several CHF C/Ls (covered in closed session slides)
February 19-20, 2020 Non-Proprietary 14
Conclusions
- NuScale LOCA EM and NRELAP5 V1.4 are approved for determining critical heat flux and collapsed liquid level for NuScale reactor in compliance with 10CFR 50.46 Appendix K requirements
- NRELAP5 computer code V1.4 is also determined applicable to predict containment pressure and temperature subject to specific modeling requirements
- The CHF modeling is approved subject to limitations and conditions February 19-20, 2020 Non-Proprietary 15