ML18178A118
| ML18178A118 | |
| Person / Time | |
|---|---|
| Issue date: | 06/07/2018 |
| From: | Zena Abdullahi Advisory Committee on Reactor Safeguards |
| To: | |
| Abdullahi Z | |
| References | |
| NRC-3762 | |
| Download: ML18178A118 (33) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Open Session Docket Number: (n/a)
Location: Rockville, Maryland Date: Thursday, June 7, 2018 Work Order No.: NRC-3762 Pages 1-33 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 654th MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 + + + + +
8 OPEN SESSION 9 + + + + +
10 THURSDAY 11 JUNE 7, 2018 12 + + + + +
13 ROCKVILLE, MARYLAND 14 + + + + +
15 The Advisory Committee met at the Nuclear 16 Regulatory Commission, Two White Flint North, Room 17 T2B1, 11545 Rockville Pike, at 8:30 a.m., Michael 18 Corradini, Chairman, presiding.
19 COMMITTEE MEMBERS:
20 MICHAEL CORRADINI, Chairman 21 RONALD G. BALLINGER, Member 22 DENNIS C. BLEY, Member 23 WALTER L. KIRCHNER, Member 24 JOSE MARCH-LEUBA, Member 25 HAROLD B. RAY, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 JOY L. REMPE, Member 2 PETER C. RICCARDELLA, Member 3 GORDON R. SKILLMAN, Member 4 MATTHEW SUNSERI, Member 5
6 DESIGNATED FEDERAL OFFICIAL:
7 ZENA ABDULLAHI 8
9 ALSO PRESENT:
10 BRUCE BAVOL, NRO 11 MICHAEL BRADBURY, NuScale*
12 ALLYSON CALLAWAY, NuScale*
13 TIM DRZEWIECKI, NRO 14 AZAT GALIMOV, NuScale*
15 ROBERT GAMBLE, NuScale*
16 JOSH KAIZER, NRR 17 BECKY KARAS, NRO 18 ANDY LINGENFELTER, NuScale*
19 JEFF SCHMIDT, NRO 20 JIM TOMKINS, NuScale*
21 JENNIE WIKE, NuScale*
22 23 *Present via telephone 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 P R O C E E D I N G S 2 8:30 a.m.
3 CHAIR CORRADINI: Okay, the meeting will 4 come to order. This is the second day of the 654th 5 meeting of the Advisory Committee on Reactor 6 Safeguards. During today's meeting, the Committee 7 will consider the following. NuScale topical reports 8 on codes and method qualification and critical heat 9 flux, future ACRS activities on, of the Planning and 10 Procedures Subcommittee, and preparation of ACRS 11 reports.
12 The ACRS was established by statute in, 13 and is governed by the Federal Advisory Committee Act, 14 or FACA. As such, this meeting is being conducted in 15 accordance with the provisions of FACA. That means 16 that the Committee can only speak through its 17 published letter reports. We hold meeting to gather 18 information to support our deliberations.
19 Interested parties who wish to provide 20 comments can contact our offices requesting time after 21 the Federal Register Notice describing the meeting is 22 published. That said, we also set aside ten minutes 23 for extemporaneous comments from members of the public 24 attending or listening in our meetings.
25 Written comments are also welcome. Ms.
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4 1 Zena Abdullahi is the Designated Federal Official for 2 the initial portion of this meeting. Portions of the 3 session of the NuScale topical reports on codes and 4 methods qualification and critical influx may be 5 closed in order to discuss and protect information 6 designated as proprietary.
7 The ACRS section of the US NRC's public 8 website provides our charter bylaws, letter reports, 9 and full transcripts of all full and Subcommittee 10 meetings, including all slides presented at those 11 meetings. We received no written comments or requests 12 to make oral statements from members of the public 13 regarding today's session.
14 There will be a phone bridge line, and to 15 preclude interruption of the meeting, the phone bridge 16 line will be placed in a listen-in only mode during 17 the presentations and Committee discussion. A 18 transcript of portions of the meeting is being kept, 19 and it is requested that speakers use one of the 20 microphones, identify themselves, and speak with 21 sufficient clarity and volume so they can be readily 22 heard.
23 I'll note that everybody please check your 24 appliances, make sure that everything's turned off so 25 we don't get beeps or buzzes. And so we'll turn to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 1 our first topic, which will be NuScale's topical 2 reports, codes, methods, and qualifications in 3 critical heat flux with Walt Kirchner.
4 I will note that Member Rempe and Member 5 Ricardella have conflicts in particular parts of these 6 topical reports and will only listen and give 7 information as requested. Walt.
8 MEMBER KIRCHNER: Thank you, Chairman, 9 I'll keep my remarks brief. We had a meeting of the 10 Subcommittee on May 15. That was a good meeting. And 11 with that, there were some questions and they will be 12 addressed today from members. I'm going to just turn 13 it to Bruce. Bruce.
14 MR. BAVOL: Yes.
15 MEMBER KIRCHNER: Go from here.
16 MR. BAVOL: Okay, my name is Bruce Bavol, 17 I'm a project manager in the Office of New Reactors.
18 We're going to go over briefly in the public portion 19 of the session here just a quick overview of the scope 20 and the timeline for our review.
21 And again, this is the topic TR-0616-22 48793, Revision 0, Nuclear Analysis Codes and Methods; 23 and Topical Report 0116-21012, Revision 1, NuScale 24 Power Critical Heat Flux Correlations.
25 Today, in the public portion like I said, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 we're going to be going over the staff review 2 timeline, the scope of the review, and the staff 3 conclusion.
4 The technical contributors today, to my 5 right, Tim Drzewiecki, he was the lead reviewer for 6 this, both topical reports. We also have Becky Karas, 7 who's the Branch Chief of Reactor Systems; Jeff 8 Schmidt, the lead in that branch; and additional 9 support, Josh Kaizer's also here.
10 For the timeline for the nuclear codes and 11 methods qualification, that topical report came in 12 August 30, 2016, was docketed in our ADAMS, Agencywide 13 Document Access Management System. We only had one 14 request for additional information, RAI 8807. That 15 RAI was dispositioned in the safety evaluation that 16 was provided to the ACRS members.
17 The staff does plan to issue, if 18 everything goes according to plan, the final SER in 19 late July 2018 and publish a -A approved version of 20 the topical report in October 2018. And with that, 21 I'll let Tim, turn it over to Tim to go over the 22 scope.
23 MR. DRZEWIECKI: Okay. So in terms of how 24 these reviews were done, this slide pertains to the 25 codes and methods review, is that findings made for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 1 this topical is that you can use this methodology, 2 which is then going to be used later, or is used as 3 part of a design certification application, to make 4 findings with several other GDCs.
5 We requested the use of the CMS5 software.
6 Some base values, the NRFs, which are used to account 7 for the uncertainty in some of the reactor physics 8 inputs, as well as an update methodology to those 9 NRFs. Things that the staff had looked at was to 10 ensure the code could actually capture the appropriate 11 geometry material and the physics.
12 We ensured that there was, on the 13 appropriate, on the assessment of the CMS5 code suite, 14 as well as we looked at the update methodology. We 15 also verified that there was a need to manage any 16 uncertainty during the operation. The surveillances, 17 things like that.
18 Things that the staff had looked at in 19 more detail was to ensure that there was no impact 20 from having several modules that are close together.
21 We ensured the physics modeling was appropriate, as 22 this was a lower flow reactor than we had seen it 23 before. So we ensured the thermal-hydraulics 24 modeling, which is inside of SIMULATE5, was 25 appropriate, and that they had appropriate validation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 of analyses for these high leakage cores and heavy 2 reflectors.
3 Things the staff had looked at was of 4 course this topical report itself, but we had also 5 looked at other information that was provided by 6 Studsvik as part of a generic CMS5, the code review.
7 And then of all of the manuals, which we've seen some 8 of that done here today.
9 We also looked at other reviews that were 10 done in this area. So things that we have found as 11 part of this review is that the CMS5 code suite does 12 apply to the NuScale design.
13 We had one limitation such that if they 14 were to make any kind of fuel changes or something 15 like that, it would have to be done within the 16 limitations of certain materials, which are bounded 17 and have been reviewed as part of the CMS5 code suite 18 review.
19 We found the base NRFs are acceptable.
20 For the record, all the NRFs are larger than what, or 21 they account for more uncertainty than what is 22 currently out there, the base NRFs for the entire op 23 fleet.
24 And as well as the NRF methodology, we 25 imposed a limitation and one condition on this in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 order to ensure that there are certain features inside 2 of the code suite that are kind of fixed and that you 3 really can't change without the aid of a code vendor.
4 And so they are kind of limited to those.
5 MEMBER MARCH-LEUBA: Tim, hold it. I've 6 been looking ahead at your closed session slides, and 7 I realized that you are going to agree with some 8 members' suggestions to add a fourth condition.
9 MR. DRZEWIECKI: That's right, yeah.
10 MEMBER MARCH-LEUBA: Can that be addressed 11 in the open session, or it that --
12 MR. DRZEWIECKI: I think it can be.
13 MEMBER MARCH-LEUBA: In addition, right?
14 MR. DRZEWIECKI: Yeah. And so I will do 15 that now. Let me make sure I got my language correct 16 here. Okay, so there was a comment in terms of the 17 branch structure, because it was not clear that they 18 had an appropriate branch structure in order to 19 capture things at high void and things like that.
20 So you know, we had gone back, there will 21 be a change made into the SE. Because the staff did 22 look at the branching structure, and we verified it 23 was appropriate for use and it can capture all the 24 normal and off-normal conditions.
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10 1 captures the void fraction, as you can see, especially 2 during something like an ECCS actuation. So we're 3 going to add that language.
4 In addition, all of the cases that are 5 supporting this code suite were done using a case 6 matrix which is developed by Studsvik. And what that 7 case matrix is is the histories and the branch 8 structures used to make all the cross sections. And 9 that is the S5C case matrix.
10 And so we're going to add a limitation 11 saying that you have to use this case matrix because 12 that was used in all their cases.
13 CHAIR CORRADINI: And we can get into 14 details in the closed session.
15 MR. DRZEWIECKI: Yeah.
16 CHAIR CORRADINI: Okay.
17 MR. DRZEWIECKI: I think I can say that 18 off in the closed session.
19 CHAIR CORRADINI: Okay, thank you.
20 MR. DRZEWIECKI: Thank you.
21 MR. BAVOL: Okay, moving on to the 22 critical heat flux topical report. The original 23 Revision 0 was submitted in October 2016, and in 24 November 30, a Revision 1 was provided to the NRC and 25 the revised submittal implements an additional NSP4 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 1 critical heat flux correlation and incorporates some 2 of the changes that the NRC requested and requests for 3 addition information.
4 There were two RAIs that were submitted 5 under this review for this particular topical report, 6 and those two responses and evaluations were 7 dispositioned in the safety evaluation. Again, the 8 staff plans to provide a final SER in late July 2018 9 with the codes and methods at the same time, and then 10 publish the -A approved version early October 2018.
11 MR. DRZEWIECKI: Okay, so in terms of the 12 staff review, again, on the CHF review itself, you 13 don't make any findings associated with any of GDCs, 14 but it used to support findings associated with things 15 like GDC10, 12, as well as any kind of dose 16 consequences analysis. Because this is used to verify 17 that, you know, you may or may not have a fuel damage.
18 They had requested the approval for the 19 NSP2 and NSP4 CHF correlations with very specific 20 requirements. It's only for new fuel, HTP2 fuel, over 21 a specified range of application, as well as certain 22 CHFR limits. So that's the approval that was being 23 sought.
24 Key areas of the review, there's really 25 three large ones, and that's the data itself, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 model generation, and a large emphasis was the model 2 error, which is actually used to develop the CHFR 3 limits.
4 So areas in which the staff spent a little 5 more time was really on the development of the 95/95 6 limits. We also wanted to make sure that we 7 understood some of the correlation behavior. As 8 discussed before, traditional PWR on a DMVR 9 correlation you would expect to see at a higher 10 pressure that you would gain from a margin.
11 In other words, if you had, you know, say 12 like, an event in which pressure was reduced, you 13 would usually see less thermal margin. And you'd be 14 closer to getting a DMVR on the event. This 15 correlation shows a different trend.
16 So as you increase pressure in your 17 system, you get less thermal margin, or you're closer 18 to CHFR. But they had data to back that up. And 19 that's also been seen in other correlations and other 20 data, historical data.
21 Things that the staff had looked at was 22 obviously this correlation itself, as well as there 23 was an inspection at Stern Laboratories, that was some 24 of the early testing that was done for a fuel type 25 that was not the prototype, which is currently being NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 1 used in the plant, but for an early prototype for the 2 NuScale design.
3 There was a test report that was out in 4 2014, as well as staff did an audit of the testing at 5 the KATHY facility in Germany. And we also built an 6 appendix as part of the CHF SE, which was based on 7 information obtained from the review of other CHF 8 correlations.
9 MEMBER SKILLMAN: Tim, back to your 10 explanation of margin change for increasing pressure.
11 Is there a brief explanation for why that phenomenon 12 occurs the way it does?
13 MEMBER BLEY: The physics, not the data, 14 the physics.
15 MR. DRZEWIECKI: So, okay, well, we -- so 16 I will give you an explanation of why I think it is.
17 Now I could say, on the basis for the finding, it was 18 the fact that they had data. That was the real basis.
19 But what I think it is is that as you --
20 so in terms of, you know, of thermodynamics, and you 21 know, please step in if you disagree with me, Mr.
22 Cardini, but as I get to a lower pressure, my latent 23 heat of vaporization gets larger.
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14 1 that layer. And so that's why I think it is. So does 2 that make sense? Like if you think of, like a TS 3 diagram in thermodynamics, there's that dome, and as 4 you go to higher and higher pressures, it takes less 5 energy to boil water. And I think that's what it is.
6 MEMBER SKILLMAN: Except that the pin 7 temperature must be at or above a certain value to 8 drive the heat across that vapor. So it's the pin 9 temperature, specifically the clad temperature I would 10 think, that is of greatest concern.
11 So in order to drive that heat across that 12 film, you've got to have a much higher fuel 13 temperature, and across the gap inside the pen, and 14 then across the clad.
15 So this kind of ties into our discussion 16 yesterday about how the NsScale design sometimes acts 17 like a P and sometimes acts like B. And what you're 18 pointing to here is a case where as you higher in 19 pressure, the thermal margin decreases. It just 20 almost, to me it's counterintuitive. It may be 21 absolutely accurate and based on data, but it sure 22 seems backwards to me.
23 MR. DRZEWIECKI: Well, that was the first 24 impression that we had when we first saw this 25 correlation. Because we had heard of that phenomena NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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15 1 before we had gone to the audit at KATHY. And that --
2 CHAIR CORRADINI: I'm sorry.
3 MR. DRZEWIECKI: Oh, please, if you want 4 to interject.
5 CHAIR CORRADINI: No, no.
6 MEMBER KIRCHNER: Actually, I do. So I 7 pondered this too, because it is somewhat 8 counterintuitive, especially for in comparison to the 9 many of the CHF correlations that have been presented 10 and audited and developed in the past.
11 I think that what's also different here is 12 the flow rates are significantly less than we'd see.
13 So we've got a regime change in how they're 14 approaching the critical, the crisis. And it's more 15 in a dryout mode almost, because I think it's higher 16 local quality in general as well. And so my sense is 17 that's what we're seeing.
18 Whereas if you look at higher pressure and 19 much higher flow rate data, then you by and large get 20 more margin with pressure, and that's typical of most 21 the PWR correlations we've seen in the past.
22 MR. DRZEWIECKI: That's right, I agree 23 with that. And I do want to state, so, yeah, so in 24 order to get an understanding of that, we had gone 25 out, that was one of the drivers for us doing an audit NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 1 at the KATHY facility, was for us to get an early look 2 at that data, but to make sure we understood what was 3 going on.
4 As a secondary check on that as well, 5 there is these, there's this, it's the Groeneveld CHF 6 look-up tables, which are like an historical, you 7 know, database. And so we had looked for those trends 8 in that table and found them there, at the same flow 9 rates, at the same you know, regimes, we saw the same 10 behavior. So that gave us more comfort.
11 MEMBER SKILLMAN: Thank you, that was 12 helpful for me. Thank you very much, thanks.
13 MR. DRZEWIECKI: Okay, what so the staff 14 found is that the NSP2 CHF correlation is acceptable 15 for use of the safety analysis of NuScale power 16 module, but new fuel NSP2 fuel, with a CHF limit of 17 1.17 over a range which is specified in that topical, 18 that's table 8.2, subject to two limitations.
19 And similarly, the NSP4 is also acceptable 20 for use with a higher limit over a range which is 21 specified in a different table. And subject to the 22 same limitations and MOC limitations are under Tong 23 Factor or the non-uniform flux factor is always a 24 penalty, it's never a credit. So it's always greater 25 than one.
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17 1 And that they have, and that, but the use 2 of these correlations have to be performed in 3 accordance with a frozen model, because all the data 4 reduction was done using a frozen model, which is 5 described in the subchannel analysis methodology.
6 MEMBER KIRCHNER: Tim, that de factor 7 implies that they have to use the VIPRE 01.
8 MR. DRZEWIECKI: That's right.
9 MEMBER KIRCHNER: Code.
10 MR. DRZEWIECKI: That's right. With a 11 certain set of correlations, yeah.
12 MEMBER KIRCHNER: I would just say in 13 plain English, so to speak, as part of, expand on 14 that. Otherwise, people go to the reference and only 15 define that it's the VIPRE 01 code that they have to 16 use, right. Can we put that in the SE? Well, that's 17 your choice, but.
18 MR. DRZEWIECKI: You know what, it's --
19 you think it's we have to identify? I mean, they 20 identify the TR, but.
21 MEMBER KIRCHNER: No, well, they use that 22 same code to reduce the data. So.
23 MR. DRZEWIECKI: Yeah.
24 MEMBER KIRCHNER: But it does beg the 25 question how good is the correlation outside of using NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 1 that code. You're putting a limitation on for 2 NuScale, but in a generic sense, this -- of course, 3 it's tied to that new fuel, HTPTM.
4 MR. DRZEWIECKI: Yeah.
5 MEMBER KIRCHNER: But just you know, is it 6 really dependent on the VIPRE code?
7 MR. DRZEWIECKI: Well, I will say --
8 MEMBER KIRCHNER: Or it's just a 9 limitation you're putting on NuScale's application of 10 this report?
11 MR. DRZEWIECKI: Yeah, so we're putting in 12 on there because in terms of, you know, of all the 13 error quantification was done using that code with 14 these tools. So if you use, yeah, so if they use some 15 other tool, it is possible, but they'd have to do more 16 work.
17 MEMBER KIRCHNER: Do more work, okay.
18 Thank you.
19 MR. BAVOL: So this concludes our public 20 portion presentation.
21 MEMBER KIRCHNER: Do you want me to ask 22 for public comment?
23 CHAIR CORRADINI: I think yes.
24 MEMBER KIRCHNER: Okay. Is there anyone 25 in the audience who wishes to make a comment while NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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19 1 we're waiting for the phone line to be open? Okay.
2 CHAIR CORRADINI: We'll wait for the 3 characteristic crackle. Anybody on the phone line?
4 MEMBER KIRCHNER: Is there anyone on the 5 phone line who wishes to make a comment?
6 (Off record comments.)
7 MEMBER KIRCHNER: Once again, anyone on 8 the open line wishing to make a comment? Mr.
9 Chairman, hearing none.
10 CHAIR CORRADINI: Okay, so Zena, please 11 close it. And also, I think there will be NuScale 12 folks that may be on the closed line for the closed 13 session.
14 So give us a moment to change over our 15 technology. And if you're on the closed line, please 16 acknowledge that you're out there. I think we're 17 expecting some subject matter experts from NuScale.
18 Is that correct? But you guys, staff does not expect 19 anybody on the line.
20 MR. BAVOL: That's correct.
21 CHAIR CORRADINI: Okay. So the closed 22 line is open now. So is anybody on the line, please?
23 Okay.
24 MR. BRADBURY: Good morning, this is Mike 25 Bradbury from NuScale.
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20 1 CHAIR CORRADINI: Okay, anybody else?
2 MR. LINGENFELTER: Andy Lingenfelter from 3 NuScale.
4 CHAIR CORRADINI: Okay, anybody else?
5 MR. GAMBLE: Robert Gamble, NuScale.
6 CHAIR CORRADINI: So all right, we've got 7 three NuScale engineers online, anybody else? Okay, 8 if you guys could mute your lines on your side, so 9 there's no kind of behind-the-scenes chatter, and 10 we'll keep the closed line open. If we need you, 11 we'll inquire.
12 Walt. So we'll go into closed.
13 MEMBER KIRCHNER: Yes.
14 CHAIR CORRADINI: We'll go into closed 15 session.
16 (Whereupon, the above-entitled matter went 17 off the record at 8:54 a.m.)
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Presentation to the ACRS Full Committee Staff Review of NuScale Topical Reports TR-0616-48793, REV 0, NUCLEAR ANALYSIS CODES AND METHODS QUALIFICATION TR-0116-21012, REV 1 NUSCALE POWER CRITICAL HEAT FLUX CORRELATIONS Presenters:
Bruce Bavol - Project Manager, Office of New Reactors Timothy Drzewiecki, Ph.D.- Reactor Systems Engineer, Office of New Reactors June 7, 2018 (Open Session)
Non-Proprietary 1
NRC Technical Review Areas/Contributors Reactor Systems NRO/DSRA/SRSB:
Jeffrey Schmidt Timothy Drzewiecki Additional Support on Nuclear Analysis Methodology Boyce Travis, NRO/DSRA/SCVB Additional Support on CHF:
Joshua Kaizer, NRR/DSS/SNPB David Heeszel, NRO/DSEA/RGS 2 Non-Proprietary
Staff Review Timeline TR-0616-48793, Nuclear Analysis Codes and Methods Qualification NuScale submitted its Topical Report (TR) TR-0616-48793, Nuclear Analysis Codes and Methods Qualification, on August 30, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16243A517).
Staff issued request for additional information (RAI 8807) on May 9, 2017 NuScale responded to RAI 8807 on July 6, 2017 (ADAMS Accession No. ML17187B239)
Staff plans to issue its final SER in late July 2018 Staff plans to publish the -A (approved) version of the TR in early October 2018 3 Non-Proprietary
Scope of the Staff Review
- Regulatory Basis
- An approved nuclear analysis methodology is used to establish a partial basis for compliance with several general design criteria (GDC) of 10 CFR 50, Appendix A
- Applicant requested approval of:
- 1. Applicability of Studsvik Scandpower, Inc. (Studsvik), Core Management Software, Version 5 suite (CMS5) to the NuScale reactor core design (RXC)
- 2. Base nuclear reliability factors (NRFs)
- 3. NRF update methodology
- Key areas of NRC staff review:
- Geometric, Material, and Physics modeling capabilities of CMS5 (applicability to NuScale RXC)
- Assessment of CMS5 suite (code-to-code and empirical data)
- NRF development and update methodology
- Uncertainty management during operation 4 Non-Proprietary
Scope of the Staff Review (cont)
- Areas requiring additional review (due to unique design features)
- Geometry Modeling: Multi-module effects and modeling of heavy reflector
- Physics Modeling: Thermal-hydraulic modeling capabilities in SIMULATE5
- Validation analyses to address high leakage cores and heavy reflectors
- Information considered by NRC staff:
- TR-0616-48793, Nuclear Analysis Codes and Methods Qualification, and supporting information obtained via RAIs and audit of supporting calculations
- Documentation provided by Studsvik in support of generic CMS5 topical report:
- SSP-14/P01-028-TR-NP, Generic Application of Studsvik Scandpower Core Management System to Pressurized Water Reactors (ML15355A285)
- CASMO5, SIMULATE5, CMSLINK5 manuals
- Similar topical reports/safety evaluations (Duke Power Company Nuclear Design Methodology, Arizona Public Service Company PWR Reactor Physics Methodology) 5 Non-Proprietary
Staff SER Conclusions
- 1. CMS5 code suite is applicable to the NuScale RXC design, subject to Limitation 1:
- Limitation 1: Application of TR-0616-48793 is limited to the materials identified in the safety evaluation for the generic CMS5 methodology
- 2. The base NRFs proposed in TR-0616-48793 are acceptable
- 3. The NRF update methodology proposed in TR-0616-48793 is acceptable, subject to Limitation 2 and Condition 1
- Limitation 2: Updates to any delayed neutron parameter NRF cannot reduce the magnitude of the NRF below 5 percent
- Condition 1: Updates to the pin peaking NRFs (FH and FQ) must include the pin-to-box bias and fixed in-core detector bias in accordance with the generic CMS5 topical report.
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Staff Review Timeline TR-0116-21012, Revision 1, Critical Heat Flux Correlations NuScale submitted its Topical Report (TR) TR-0116-21012, Revision 1, Critical Heat Flux Correlations, on November 30, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17335A089), this revision replaced the submittal dated October 5, 2016 (ADAMS Accession No. ML16279A363), NuScale Power Critical Heat Flux Correlation NSP2.
The revised submittal implements an additional NSP4 critical heat flux (CHF) correlation and incorporates changes associated with NRC requests for additional information (RAIs).
Staff issued two requests for additional information (RAI 8795 on May 8, 2017 - ADAMS Accession No. ML17128A468) and (RAI 8931 on August 21, 2017 - ADAMS Accession No. ML17233A127)
NuScale responded to RAI 8795 on (July 7, 2017 - ADAMS Accession No. ML17188A461) and RAI 8931 on (September 25, 2017 - ADAMS Accession No. ML17268A385)
Staff plans to issue its final SER in late July 2018 Staff plans to publish the -A (approved) version of the TR in early October 2018 7 Non-Proprietary
Scope of the Staff Review
- Regulatory Basis
- An approved critical heat flux (CHF) correlation is used to establish a partial basis for compliance with several regulations including the general design criteria (GDC) of 10 CFR 50, Appendix A:
- GDC 10 and GDC 12 (margin to fuel design limits), 10 CFR 52.47(a)(2)(iv) and GDC 19 (radiological consequences)
- Applicant requested approval of:
- Use of NSP2 and NPS4 CHF correlations for the safety analysis of the NuScale Power Module
- NuFuel-HTP2TM fuel
- Over a specified range of applicability
- With specified CHF ratio (CHFR) limits
- Key areas of NRC staff review:
- Collection of experimental data
- CHF model generation
- Quantification of CHF model error (development of CHFR limits) 8 Non-Proprietary
Scope of the Staff Review (cont)
- Areas requiring additional review
- Statistical Analysis - Development of the 95/95 CHFR limits
- Correlation behavior differs from trends observed in previously reviewed CHF correlations for pressurized water reactors
- Information considered by NRC staff:
- TR-0116-21012, Revision 1, NuScale Power Critical Heat Flux Correlations, and supporting information obtained via RAIs and audit of supporting calculations
- Inspection report for CHF testing at Stern Laboratories, April 18, 2013
- NuScale CHF test program technical report, January 24, 2014
- Audit report for CHF testing at KATHY, August 12, 2016
- Information obtained from previously reviewed CHF correlations as documented in Appendix A of the staffs SE 9 Non-Proprietary
Staff SER Conclusions
- NSP2 CHF correlation is acceptable for use in performing safety analyses of the NPM with NuFuel-HTPTM fuel
- With a CHFR limit of 1.17
- Over the range of applicability provided in Table 8-2 of TR-0116-21012, Rev. 1
- Subject to Limitation 1 and Limitation 2
- NSP4 CHF correlation is acceptable for use in performing safety analyses of the NPM with NuFuel-HTPTM fuel
- With a CHFR limit of 1.21
- Over the range of applicability provided in Table 8-4 of TR-0116-21012, Rev. 1
- Subject to Limitation 1 and Limitation 2
- Limitations
- Limitation 1: Nonuniform flux factors must always be greater than or equal to 1
- Limitation 2: CHF analyses using the NSP2 and NSP4 correlations must be performed in accordance with TR-0915-17564, Subchannel Analysis Methodology 10 Non-Proprietary
Questions/comments from members of the public before the closed session starts?
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