ML20093G897
| ML20093G897 | |
| Person / Time | |
|---|---|
| Issue date: | 03/05/2020 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Burkhart, L, ACRS | |
| References | |
| NRC-0839 | |
| Download: ML20093G897 (243) | |
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, March 5, 2020 Work Order No.: NRC-0839 Pages 1-117 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 671ST MEETING 5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 6 (ACRS) 7 OPEN SESSION 8 + + + + +
9 THURSDAY 10 MARCH 5, 2020 11 + + + + +
12 ROCKVILLE, MARYLAND 13 + + + + +
14 The Advisory Committee met at the Nuclear 15 Regulatory Commission, Two White Flint North, Room 16 T2D10, 11545 Rockville Pike, at 8:30 a.m., Matthew W.
17 Sunseri, Chairman, presiding.
18 19 COMMITTEE MEMBERS:
20 MATTHEW W. SUNSERI, Chairman 21 JOY L. REMPE, Vice Chairman 22 WALTER L. KIRCHNER, Member-at-Large 23 RONALD G. BALLINGER, Member 24 DENNIS BLEY, Member 25 CHARLES H. BROWN, JR., Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 VESNA B. DIMITRIJEVIC, Member 2 JOSE MARCH-LEUBA, Member 3 DAVID PETTI, Member 4 PETER RICCARDELLA, Member 5
6 ACRS CONSULTANTS:
7 STEPHEN SCHULTZ 8
9 DESIGNATED FEDERAL OFFICIAL:
10 MIKE SNODDERLY 11 12 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 C-O-N-T-E-N-T-S 2 Opening Remarks by the ACRS Chairman . . . . . . 4 3 NuScale Areas of Focus: Steam Generator 4 Design, Containment Evacuation System and 5 Hydrogen & Oxygen Monitoring . . . . . . . . . . 8 6 NuScale Topical Reports: Loss of Coolant Accident 7 (LOCA), Non-LOCA and Rod Ejection Accident 8 Methodologies . . . . . . . . . . . . . . . . . . 74 9
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4 1 P R O C E E D I N G S 2 (8:30 a.m.)
3 CHAIRMAN SUNSERI: The meeting will now 4 come to order. This is the first day of the 671st 5 meeting of the Advisory Committee on Reactor 6 Safeguards.
7 I am Matthew Sunseri, the Chair of the 8 ACRS. Members in attendance today are Pete 9 Riccardella, Ron Ballinger, Dave Petti, Joy Rempe, 10 Walt Kirchner, Jose March-Leuba, Charlie Brown.
11 Dennis Bley is here. He'll be stepping in 12 a minute and Vesna Dimitrijevic. We also have our 13 consultant, Steve Schultz present as well. And I note 14 that we have a quorum.
15 The ACRS was established by the Atomic 16 Energy Act and it's governed by the Federal Advisory 17 Committee Act.
18 The ACRS section of the U.S. NRC public 19 website provides information about the history of the 20 ACRS and provides documents such as our charter, 21 bylaws, Federal Register notices for meetings, letter 22 reports and transcripts of all full and subcommittee 23 meetings, including slides presented at the meetings.
24 The Committee provides its advice on 25 safety matters to the Commission through its publicly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 1 available letter reports. The Federal Register notice 2 announcing this meeting was published on February 21, 3 2020, and provides an agenda and instructions for 4 interested parties to provide written documents or 5 request opportunity to address the Committee.
6 The Designated Federal Official for this 7 meeting is Mr. Mike Snodderly. During today's meeting 8 the Committee will consider the following.
9 NuScale Area of Focus: Steam Generator 10 Design, Containment Evacuation System and Hydrogen and 11 Oxygen monitoring and number two, NuScale Topical 12 Reports: Loss of Coolant Accident (LOCA), Non-LOCA 13 and Rod Ejection Accident Methodology.
14 Following those presentations the ACRS 15 will engage in preparation of reports. As reflected 16 in our agenda, portions of the NuScale session may be 17 closed in order to discuss and protect information 18 designated as sensitive or proprietary. And I will 19 say there will be closed sessions today.
20 A phone bridge line has been opened to 21 allow members of the public to listen in on the 22 presentations and Committee discussion. We have 23 received no written comments or requests to make oral 24 statements from members of the public regarding 25 today's session.
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6 1 There will be an opportunity for public 2 comment and we have set aside time in the agenda for 3 comments from members of the public attending or 4 listening to our meetings. Written comments may be 5 forwarded to Mr. Mike Snodderly, the Designated 6 Federal Official.
7 A transcript of the open portion of the 8 meeting is being kept and it is requested that 9 speakers use one of the microphones, identify 10 themselves and speak with sufficient clarity and 11 volume so that they may readily be heard.
12 For the people that will be presenting 13 today, I ask that you consider the following. We've 14 seen a lot of the material. And in most of the 15 subcommittee meetings on these topics we've had full 16 committee membership participation.
17 So, please feel free to progress smartly 18 through, you know, maybe the background material and 19 stuff that we've seen before and focus your detail on 20 the things that you've been briefed on as important to 21 us because we know you know what topics are important 22 to us.
23 If we need to slow you down we will slow 24 you down. So, let us control the pace.
25 Just one thing before we get into the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 1 presentations. I do have an item of interest that I 2 want to make public. Today in the Federal Register 3 notice a notice was published that we are seeking 4 qualified candidates for membership on the ACRS.
5 The ACRS is seeking two members, one with 6 nuclear power plant experience and a second one 7 regarding, with risk analysis and the consideration of 8 uncertainty in decision making. So, those positions 9 fill out vacant and soon to be vacant with retirement 10 the positions.
11 And any interested candidates should 12 follow the instructions on the Federal Register 13 notice. We will now begin the presentations with 14 NuScale.
15 And I'll turn to staff to see if they have 16 any remarks that you want to make before the NuScale 17 presentation. Who is, Rebecca, are you over there?
18 MS. PATTON: No. We just thank the 19 Committee for their time and hope for a productive 20 dialogue.
21 CHAIRMAN SUNSERI: Okay, thank you. And 22 now, Marty, the floor is yours for the NuScale.
23 MR. BRYAN: Okay, thanks, Matt. I'm Marty 24 Bryan. I'm the licensing project manager for Chapter 25 3. I've got with me Bob Houser, Kevin Spencer, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 Matthew Presson and also Brian Wolf will be joining us 2 on the phone for part of the presentation.
3 So, today in open session it's fairly 4 brief. We're going to get into more of the feedback 5 we received in the closed session. But certainly ask 6 questions if something comes up.
7 So, we're going to do just a brief 8 overview of Steam Generator Design and then talk a 9 little bit about the proposed DCA revisions that we 10 intend to include in the errata for Rev 4. So, I'll 11 turn it over to Kevin.
12 MR. SPENCER: So, I'm Kevin Spencer. I'll 13 be doing a brief overview of the Steam Generator 14 Design this morning. This was previously presented so 15 I'll try to -- I'll make it fairly high level.
16 Each NuScale power module has two steam 17 generators. On the shell side we have the primary 18 fluid. On the tube side we have the secondary fluid.
19 We have about 1,380 tubes overall. They 20 range in length from 74 to 86 feet. It is a helical 21 coil design. Each tube is made out of Alloy 690, 22 thermally treated material.
23 I have brought with me this morning a 24 little, a plastic prototype of the steam generator 25 tubes and how they interact with the steam generator NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 supports. I'll pass this around.
2 Feel free, it does come apart. If it 3 falls apart you can put it back together easily. But 4 it will allow you to take a look at how the helical 5 coil tubes interact with the tube supports. So, I'll 6 pass this around.
7 MEMBER MARCH-LEUBA: While you still have 8 it in your hand, what's the length of the straight 9 shot on the tube? When does it start curving because 10 you're going to put the other thing, the metal thing 11 inside it, right?
12 MR. SPENCER: Yes. So, the helical coil 13 this is, the supports have the, work on the helical 14 coil section of it.
15 But at the, where it intersects with the 16 steam and feedwater plenum you can kind of see on the 17 drawing on the left-hand side here there is a straight 18 section, a straight leg section.
19 MEMBER MARCH-LEUBA: You need to look at 20 the microphone or he can't hear you.
21 MR. SPENCER: Okay. There is a straight 22 leg section down at the feedwater plenum and at the 23 steam plenum. That's a transition from the helical 24 coil to a straight tube.
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10 1 it's typically on the order of 20 to 30 inches at 2 least on the feedwater side.
3 MEMBER MARCH-LEUBA: So, you have like 20 4 inches of straight?
5 MR. SPENCER: Yes.
6 MEMBER MARCH-LEUBA: Good. That's good 7 information to have.
8 MR. SPENCER: Yes. And I do want to note, 9 we can actually just probably go to the next slide and 10 I'll do the IFR.
11 I did bring a prototype inlet flow 12 restrictor as well. Now this one is a prototype so 13 it's a little bit longer than the one you'll see on 14 the screen which is representative of the actual 15 design.
16 Notably, this has eight sections and the 17 actual design has five sections. This also doesn't 18 have the threaded connection that will thread it onto 19 the plate.
20 But it is kind of -- it's prototypical so 21 you it would allow you to get a feel for it.
22 MEMBER MARCH-LEUBA: That's not 23 proprietary, the design?
24 MR. SPENCER: No. Not in this form 25 without dimensions and such.
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11 1 MEMBER MARCH-LEUBA: The dimensions are 2 proprietary. But the number of stages is not 3 proprietary.
4 MR. SPENCER: Right, right.
5 MEMBER MARCH-LEUBA: Okay.
6 MEMBER RICCARDELLA: I note that from this 7 model that tubes can slide axially. Is that true in 8 the actual model?
9 MR. SPENCER: That won't be necessarily 10 true in the actual model because the helical coil will 11 be constrained on all sides.
12 But what I did want to mention here with 13 the five, with the set of five expansions you'll 14 notice that the IFR is contained within the actual 15 tube sheet.
16 So, it doesn't extend out past the, it 17 doesn't extend past the tube sheet into the heated 18 area.
19 MEMBER MARCH-LEUBA: What is the tube 20 sheet?
21 MR. SPENCER: Yes. So, it's not as long, 22 the --
23 MEMBER MARCH-LEUBA: So, this is outside 24 of the primary? It's not in contact with the primary 25 fluid?
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12 1 MR. SPENCER: That's correct. That's 2 correct.
3 CHAIRMAN SUNSERI: So, when you said the 4 tube straight piece is 30 inches or so is that 5 including the length through the tube sheet or after 6 it passes through the tube sheet?
7 MR. SPENCER: The straight section from 8 the feedwater transition plenum is probably on the 9 range from 20 to maybe 35 inches overall. And then 10 that does include the length of tube which is, which 11 passes through the tube sheet and is welded on the 12 secondary face of the tube sheet.
13 I think I can say that it's probably not 14 proprietary to say that. That's on the order of six 15 inches is the thickness of the tube sheet.
16 MEMBER MARCH-LEUBA: So, just so I can 17 visualize it. Where the IFR is inserted that is not 18 a tube but is a stronger piece of material?
19 MR. SPENCER: It is, it's a tube that's 20 passed through a hole. So, there's a six inch thick 21 metal plate.
22 MEMBER MARCH-LEUBA: So, it's a thick 23 metal plate with drills.
24 MR. SPENCER: Yes, with the appropriate --
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13 1 is inserted into the tube sheet. It's hydraulically 2 expanded.
3 So, it's pushed out with force up against 4 those walls. And then it's, there's a fillet weld on 5 the end of the tube on the secondary face.
6 MEMBER MARCH-LEUBA: So, it's welded at 7 the bottom?
8 MR. SPENCER: So, in this drawing here it 9 would be welded in between the IFR mounting plate.
10 And you'll see there's clouding on that second side.
11 That's to allow a similar metal weld.
12 MEMBER MARCH-LEUBA: The IFR is held in 13 place from the back on the, with a screw?
14 MR. SPENCER: Yes. So, there's an IFR 15 plate that all these, each IFR is inserted into the 16 plate. It's mounted through a threaded section 17 through the plate.
18 Ideally that's going to be a loose design 19 when it's inserted into the tube so that it will allow 20 each IFR to be seated into the tubes. That plate will 21 be mounted through various mounting studs to the 22 actual tube sheet.
23 That will prevent any sort of bowing or 24 flexure of that plate. And then once all that is in 25 position then those IFR, then the IFR threads NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 1 themselves will be tightened up and preloaded.
2 MEMBER MARCH-LEUBA: And you do this every 3 refueling, to load it?
4 MR. SPENCER: Yes. This will be --
5 MEMBER MARCH-LEUBA: So, you loosen the 6 screw in the back for every one of them and then put 7 them in?
8 MR. SPENCER: Yes. Not for every 9 refueling but for every inspection.
10 MEMBER MARCH-LEUBA: Yes, right. Every 11 time you take it apart.
12 MR. SPENCER: Yes. And it may be during 13 a steam generator inspection you may be doing 100 14 percent inspection of the tubes. You may also be 15 inspecting some smaller number of the tubes based on 16 the steam generator program that the utility sits on.
17 MEMBER MARCH-LEUBA: But all the IFRs are 18 on the same plate?
19 MR. SPENCER: I'm sorry.
20 MEMBER MARCH-LEUBA: All of the IFRs are 21 on the same plate --
22 MR. SPENCER: Yes.
23 MEMBER MARCH-LEUBA: -- for each entrance?
24 MR. SPENCER: Yes.
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15 1 them.
2 MEMBER BALLINGER: What is the orientation 3 of the flow restrictor, is the left-hand end the 4 furthest end of the tube sheet?
5 MR. SPENCER: The furthest end of the tube 6 sheet is the tip, yes.
7 MEMBER BALLINGER: Yes. So, is there any 8 concern about vibration there? It's a very short, 9 it's a sharp V on the thing.
10 Is there any concern that you might have 11 a wear problem on that point there because that's on 12 the hydraulically expanded part?
13 MR. SPENCER: Yes.
14 MEMBER BALLINGER: So, is there 15 possibility of this thing doing this?
16 MR. SPENCER: So, we've done a significant 17 amount of testing with respect to forward flows, flows 18 in the nominal direction from the feedwater into the 19 tube at velocities, we've done prototypic testing 20 where we're looking at Reynolds numbers that are much 21 higher that we would expect and the turbulent buffing 22 that we've looked at and any sort of vibration that 23 we've looked at has not been a cause for concern for 24 the IFR.
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16 1 no oscillations, no flow oscillations, correct?
2 MR. SPENCER: That's, so, yes. That 3 explicitly has been forward flow on the IFR.
4 MEMBER MARCH-LEUBA: For 100, 120 percent 5 nominal flow, not 300 percent nominal flow?
6 MR. SPENCER: I want to say that we've 7 gone up to like maybe 800 percent flow in our testing.
8 MEMBER MARCH-LEUBA: On the --
9 MR. SPENCER: In the forward direction.
10 MEMBER MARCH-LEUBA: Vibration testing?
11 MR. SPENCER: Yes, prototypically. Not at 12 temperature and pressure. But --
13 MEMBER MARCH-LEUBA: And this thing is 14 screwed into a plate on the back, right?
15 MR. SPENCER: Yes.
16 MEMBER MARCH-LEUBA: Yes, a Phillips 17 screwdriver. Hopefully you torque it the right 18 position, you don't do it like I do?
19 MR. SPENCER: Yes. Well, it will be a 20 hardware design that will prevent loose parts. So, we 21 wouldn't want to have loose parts from this. But it 22 will be, so it will be --
23 MEMBER MARCH-LEUBA: You have 1,200 of 24 these. One of them after ten years is not going to 25 get a little loose and go ping, ping?
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17 1 MR. SPENCER: Well, so again these would 2 be removed and, these would be considered to be a part 3 of the Steam Generator Program. So, they will be 4 inspected at the same frequency at which the tubes 5 would be inspected as a part of that Steam Generator 6 Program.
7 So, when the IFRs are removed they will, 8 you know, any time that you return a threaded part to 9 service part of your procedure in doing that is to 10 look at the condition of the threads, at the condition 11 of the mounting hardware to ensure that it can be put 12 back into service safely.
13 MEMBER MARCH-LEUBA: And I assume you look 14 inside the tube sheet to look for wear?
15 MR. SPENCER: Yes. So, there's 100 16 percent volumetric inspection of the tubes from the 17 inside. So --
18 MEMBER MARCH-LEUBA: I'm told from the 19 people that know about this that this particular alloy 20 scratches easily. Is that correct?
21 MEMBER BALLINGER: I don't know about 22 scratch easily. But its wear characteristics are much 23 different than Alloy 600.
24 MEMBER MARCH-LEUBA: It creates oxide, you 25 scratch the oxide, it creates oxide, you scratch the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 1 oxide.
2 MEMBER BALLINGER: Now I have one more 3 question. Is there any thought to having an ejection 4 collar on one of those things?
5 What I'm saying is it would be a pretty 6 bad hair day if the nut on the outside, if it were to 7 fracture there and this thing ended up going into the 8 tube.
9 But if it was designed so that there was 10 a diameter change in the plate if the nut cracked it 11 wouldn't be possible to send that thing into the tube.
12 MR. SPENCER: Yes, yes. So, we've done 13 some preliminary test analysis. I guess, I mean the 14 current design that we're here to present today is the 15 current design for the DCA.
16 You know, we do -- as we change operation, 17 if we change operationally in the future we're going 18 to also be required to change this as a function of 19 that to ensure that we have the same characteristics 20 to prevent DWO that the inlet flow restrictor is 21 designed to do.
22 So, if we change the operation that 23 affects the design and that allows us to reexamine the 24 design. But the current design that we're presenting 25 today doesn't include that feature.
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19 1 But we have done some preliminary stress 2 results. We'll present those in the closed session a 3 little bit to show that, you know, we think we have 4 sufficient margin to, any sort of ASME, you know, any 5 sort of ASME analysis on the thread or on the bolt or 6 anything like that.
7 I think I've presented this slide kind of 8 overall. If you have any questions about it otherwise 9 I suggest we move on.
10 CHAIRMAN SUNSERI: You just move on and 11 we'll stop you.
12 MR. BRYAN: One thing that is different 13 from the last time we were here, we got a lot of 14 feedback. We went back and evaluated it.
15 And we are now proposing a COL item to 16 address the evaluation methodology. And so, I'll 17 pause there just a minute and let you read the COL 18 item.
19 But this is what we proposed to address 20 developing a methodology that would evaluate the 21 secondary side instabilities including reverse flow.
22 MEMBER MARCH-LEUBA: If I'm reading 23 correctly you will ensure you have a validated tool 24 that will be able to predict instabilities and what 25 happens during them and how then to calculate the ASME NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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20 1 loads if they should happen. Is that what you're 2 saying?
3 MR. BRYAN: Yes, correct.
4 MEMBER MARCH-LEUBA: And that will be a 5 COL item?
6 MR. BRYAN: Correct.
7 MEMBER MARCH-LEUBA: Can we say carveout 8 in the open session?
9 MEMBER KIRCHNER: That has a different 10 meaning.
11 MEMBER MARCH-LEUBA: I know but, okay, 12 maybe we'll wait for the -- yes, but can we talk about 13 that?
14 CHAIRMAN SUNSERI: Why don't you wait 15 until the staff --
16 MEMBER MARCH-LEUBA: All right. I wanted 17 to see what the difference is. But we'll wait for the 18 staff to tell us what the difference is.
19 CHAIRMAN SUNSERI: Would you envision that 20 this is, this methodology be documented on a technical 21 report or a topical report or something? I'm just 22 trying to think of what, how that would get looked at.
23 MR. HOUSER: Yes, it would be. We would 24 develop something that's very. Yes. It would be 25 documented and available.
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21 1 It would be much like the methodologies 2 that were developed for the LOCA and non-LOCA topical 3 reports in terms of content. We can get into that in 4 a little bit more detail in the closed session.
5 MEMBER MARCH-LEUBA: You would issue a 6 topical or a technical report?
7 MR. BRYAN: It would be technical, I 8 think.
9 MEMBER MARCH-LEUBA: Yes, that would be 10 more likely.
11 MEMBER BROWN: But you all developed the 12 other reports. Now you're pushing this off to the COL 13 who has no background in this design other than they 14 have chosen you all as the design document, the design 15 whatever you want to call it.
16 It's kind of hard to see this guy walks in 17 cold and has to develop all this analysis technology 18 and methodology for a design that they haven't even 19 seen until they decided to go with you. Maybe I'm 20 speaking out of turn.
21 This just seems to be kind of complicated 22 when you all have spent several years developing your 23 own evaluations and design analyses and topical 24 reports, that's all.
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22 1 forward with development of that ASME scale, that 2 methodology.
3 MEMBER BROWN: So, whey the COL if you're 4 all doing it and you're all not going to provide it 5 yourself?
6 MEMBER RICCARDELLA: The timing.
7 MEMBER BROWN: I understand that. But 8 that's, time is nice. But I'm looking at it from the 9 technical standpoint and the ability to get a, I guess 10 a methodology that it's truly representative of what, 11 you know, the design and what density wave 12 oscillation.
13 I'm not a thermal hydraulic guy, okay.
14 But I know that's not good.
15 MEMBER RICCARDELLA: But it's not 16 realistic to assume that there's going to be a COL guy 17 and NuScale is just going to walk away and this COL 18 applicant is going to build the plant all by himself.
19 Come on, Charlie.
20 MR. HOUSER: That will not happen.
21 MEMBER RICCARDELLA: That's absurd.
22 MEMBER BLEY: There's another thing here.
23 Correct me if I'm wrong. If you issue it as a 24 technical, I assume you'll just move into this and 25 you're working on it.
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23 1 If you finish it and it's a technical 2 report it won't come to the staff or to us until 3 there's a COL applicant. If you issued it as a 4 topical it might come right away for approval.
5 Am I correct in that assumption of how 6 things could progress?
7 MR. BRYAN: Yes. In terms of technicals 8 and topicals that's correct.
9 MEMBER MARCH-LEUBA: But it's not 10 necessary. I mean, you could send a technical ahead 11 of time.
12 Given the visibility that it has already 13 had you likely will or you will have a visit in 14 Corvallis to go see it, I think.
15 MR. MELTON: I want to say, it's Mike 16 Melton with NuScale. So, the COL items will be 17 addressed with, you know, people that are technically 18 qualified.
19 You know, all resources will be applied to 20 make sure that methodologies or NuScale's involvement.
21 I don't think we need to be concerned 22 about because the design expertise, analysis, you know 23 consultants we'll have the right workforce to make 24 sure that this gets done properly as with all our COL 25 items.
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24 1 I just want to assure the Committee.
2 We'll make sure it gets done properly. And I think 3 Marty is correct. This does sound like a technical 4 report but I don't know if we've made that internal 5 decision.
6 But because of the applicability it's 7 probably that direction. If it does go in the form of 8 a topical report it will follow the process.
9 MR. PRESSON: And in terms of process 10 it's, you know, we have the ITAAC which is tagged to 11 the COL. But this would ensure that methodology is 12 reviewed prior to the ITAAC process. So, it would be 13 captured in the FSAR portion of that.
14 MR. DUDEK: And just to add, this is 15 Michael Dudek, the Branch Chief for Nuclear Reactors.
16 The COL versus the carveout is really, as you said, a 17 timing issue.
18 We have not seen or evaluated fully the 19 proposed COL item. Previous to that we had identified 20 a technical open item and that's where we proposed not 21 giving them finality in the role which is AKA the 22 carveout.
23 So, as we evaluate and go forward we may 24 take that off the table. But as of now it's still an 25 open item and we propose not giving finality through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 1 the rulemaking.
2 MEMBER MARCH-LEUBA: So, that carveout 3 would be a way to address this technical review, if 4 it's a technical report.
5 MR. DUDEK: What they do could suffice and 6 take that open off the table. But we have yet to 7 reach that conclusion.
8 MR. BRYAN: Okay. So, just to wrap up.
9 There is again, we got a lot of feedback. We heard 10 the feedback. We went back for both the staff and the 11 Committee and we revised both 3.9 to include the COL 12 item.
13 And we also clarified the language in 5.4.
14 There was a lot of discussion about the use of RELAP 15 there. So, we took that discussion out and replaced 16 it.
17 We thought you would have the errata by 18 now. But that got held up, that you would have seen 19 it before this meeting. But that will be forthcoming 20 in the errata letter to clean up some of the 5.4 21 language.
22 So, that's really all we had planned to 23 cover in the open session. We'll get into some more 24 of the details in the closed session.
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26 1 carveout from our perspective. As Matthew said, by 2 having successful completion of the ITAAC we have a 3 COL item, we believe this constitutes the basis for 4 NRC determination to allow operation of the facility 5 certified under 10 CFR 52.
6 MEMBER MARCH-LEUBA: You say operation, 7 you mean certification under 52, right?
8 MR. BRYAN: Yes.
9 CHAIRMAN SUNSERI: Okay, thank you.
10 Members, any questions for the presenters?
11 MEMBER MARCH-LEUBA: I just wanted to put 12 on the record that this is good. I'm happy that 13 you're taking it seriously and we are going to follow 14 through instead of trying to avoid it. So, today I'm 15 happier than I was yesterday.
16 CHAIRMAN SUNSERI: Well, that's a 17 milestone. Okay. All right, thank you. Let's bring 18 up the staff now.
19 And as you all are taking the table I 20 would remind you once again this is open. And if we 21 ask any questions that drive us to proprietary 22 information just refrain and we'll address those in 23 the closed session later.
24 MS. JOHNSON: Good morning, everyone. My 25 name is Marieliz Johnson. I'm the project, not yet.
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27 1 (Off-microphone comments.)
2 MS. JOHNSON: Do you hear me better now?
3 So, I'm Marieliz, sorry, Marieliz Johnson, project 4 manager for NuScale the certification application.
5 Today we're going to present the NRC 6 review of the NuScale steam generator. For the agenda 7 we have the NRC staff review team. We have a brief 8 summary of the review of the steam generator.
9 And we will go through a summary of the 10 steam generator design issues that are not resolved by 11 the, by certification, by the design certification 12 application. Here's a list of the review team.
13 And then I'm going to turn it over to Greg 14 Makar to continue.
15 MR. MAKAR: I'm Greg Makar from the 16 Corrosion and Steam Generator Branch. And I want to 17 briefly review our -- is that better?
18 I'm Greg Makar of Corrosion and Steam 19 Generator Branch and I want to briefly review our 20 findings on the topics for steam generator materials 21 and Steam Generator Program. And then I'll turn our 22 attention to the incomplete topic of secondary side 23 flow stability.
24 We found in most cases, except for that 25 one, we found the materials area acceptable. That NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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28 1 includes material selection and the associated 2 requirements, things like the application of the ASME 3 code and fabrication, cleaning, inspection 4 requirements.
5 The design limits the crevices along the 6 tubes and enables flow along the tubes and we found 7 that important, degradation mechanisms associated with 8 crevices.
9 The materials will be compatible with the 10 planned primary and secondary environments. And the 11 design provides for primary and secondary side access 12 for inspection, cleaning, foreign object search and 13 retrieval.
14 Next slide, please. Steam Generator 15 Program we found to be acceptable. It is consistent 16 with the standard tech specs and the industry 17 guidelines.
18 We say appropriately acceptable because 19 there are some differences in terminology and other 20 aspects of the tech specs that are different for 21 NuScale.
22 And the inspection program, it's a 23 performance based framework that has some prescriptive 24 elements and it defines tube integrity in terms of the 25 structural and or describes the performance criteria NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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29 1 in terms of the structural and leakage integrity of 2 the tubes.
3 They have provided a generic tube plug in 4 criterion which is the amount of through wall loss of 5 the tube that you can have before you have to take a 6 tube out of service.
7 And the COL applicant will submit the and 8 prepare the steam generator inspection program and 9 implement that plan and provide any site specific 10 information which includes their own degradation 11 assessment, their own plug in criterion and timing and 12 so forth. Next slide.
13 MEMBER BALLINGER: I just had something 14 pop into my head. The standard tube integrity 15 inspection technique is bobbin coil or something like 16 that.
17 But usually the, it's on the primary side, 18 goes up the primary side. In this case you're going 19 to have to go up the secondary side.
20 And if the criteria is 40 percent through 21 wall volumetric, right, that's one of the criteria for 22 tube plugging, that volumetric will be on the inside 23 not the outside of the tube. So, is there, that going 24 to work out okay?
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30 1 looking for any kind of degradation that you could 2 expect according to your degradation assessment for 3 that particular plant.
4 Some degradation has come from the inside 5 of the tube, some secondary, some --
6 MEMBER BALLINGER: Cracking is not an 7 issue. But I'm talking about removal of material, 8 volumetric defect on the inside of the tube where the 9 bobbin coil or pancake or whatever you're using goes 10 up.
11 That's a little bit different, I think, 12 then what you would find in a recirculating or once 13 received generator like in a PWR.
14 MR. MAKAR: Well, the inspection will be 15 able to detect volumetric on the inside or the 16 outside.
17 MEMBER BALLINGER: Okay.
18 MR. MAKAR: As it does now.
19 MEMBER BALLINGER: Okay.
20 PARTICIPANT: Are you worried about the 21 coil getting caught up?
22 MEMBER BALLINGER: You know, I'm not a 23 coil expert. But if it's all of a sudden now you have 24 a 40 percent volumetric defect once you have removed 25 material.
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31 1 MR. MAKAR: And that, still the most 2 likely place for that is on the outside of the tubes 3 at support structures. But it could be that this flow 4 restrictor if that, you know, we talked about that.
5 MEMBER BALLINGER: Corrosion on the 6 outside of that type doesn't concern me. You're not 7 going to get any kind of thing because it's on the 8 primary side.
9 MR. MAKAR: But the support structures are 10 on the, are also on the outside.
11 MEMBER BALLINGER: Yes, okay.
12 MR. MAKAR: So, we still need to look for 13 anything they expect on both the inside and outside.
14 MEMBER BLEY: I hadn't thought about it 15 and it's not an issue here. But in the current 16 designs where the primary is on the inside when you go 17 in to work you've got a lot of streaming coming out of 18 those tubes, radiation streaming.
19 I wonder if that's going to be different 20 or better this way around. Go ahead, I'm just 21 wondering.
22 MEMBER KIRCHNER: Proximity to the core 23 of the tube sheets is going to make for a much 24 different situation. In the current fleet the 25 inspection of the PWRs is, like you said, it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 1 whatever, particulate corrosion, whatever inside the 2 tubes.
3 This one you're much closer, the structure 4 has been sitting much closer to the core. So, I 5 wonder what activation --
6 The core is about what, ten feet lower 7 than the start of the steam generator? But that's the 8 difference I see in terms of personnel exposure. They 9 take this, put it in the dry dock and then inspect it.
10 It may be hotter, the material.
11 MEMBER BLEY: Kind of in general. But in 12 the current ones you have it on the inside of the 13 tubes and you really get a beam kind of coming out of 14 it.
15 MEMBER MARCH-LEUBA: Activation is neutron 16 flux and very few neutrons are going to make it 17 through 20 feet of water. So, there will be a gamma 18 flux.
19 But the gamma doesn't activate. In 20 inspection the core will be in a different place.
21 MEMBER BALLINGER: That's about 20 tenth 22 value layers.
23 MEMBER KIRCHNER: The other thing is that 24 assuming they keep doing water chemistry, but these 25 are low flows. So, if stuff is going to accumulate on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 1 the primary side it's going to be around the tube 2 sheet entrance on the primary side, if there's crud in 3 other things.
4 MEMBER BALLINGER: The first inspection 5 will be interesting.
6 MR. MAKAR: All right, next slide, please.
7 We have determined that this, we have this issue of 8 structural leakage integrity that has not been fully 9 demonstrated.
10 And that's related to the effective 11 density wave oscillations on tube integrity and also 12 for the method of analysis for the secondary side, 13 thermal hydraulic conditions and associated loads.
14 NuScale is working to address that topic.
15 And if there are no, unless there are other questions 16 about our Chapter 5 review I'm going to turn this over 17 to Tom Scarbrough to talk more about the secondary 18 flow instability topic.
19 MR. SCARBROUGH: Thank you, Greg. I'm Tom 20 Scarbrough with the Mechanical Engineer Branch. We 21 had quite a bit of discussions over the past few weeks 22 regarding the steam generator tubes and their 23 integrity.
24 And after quite a bit of significant 25 interactions, you know, among all the technical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 reviewers. There are a number of technical reviewers.
2 You know, there are several chapters that are involved 3 here of this.
4 And so, after a lot of deliberation we 5 decided that at this point we're going to propose that 6 we specify the structural integrity and leakage, the 7 structural and leakage integrity of the steam 8 generator tubes are not resolved and not receiving 9 finality in the NRC draft proposed rule for design 10 certification.
11 MEMBER BLEY: I would just interject here.
12 We've had concerns about wear which could lead to two 13 failures.
14 The PRA certainly has not reflected 15 anything about this phenomena if it exists.
16 MR. SCARBROUGH: Yes. And you brought 17 that point and that was one of our concerns that we've 18 talked about quite a bit over the last few weeks.
19 And so, we're going to talk about the 20 specific details of the technical reasons why in the 21 next couple slides. But I'm just kind of telling you 22 what the process is right now.
23 MEMBER RICCARDELLA: What you said, does 24 that mean carveout?
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35 1 sir. I didn't use carveout --
2 MEMBER RICCARDELLA: That's different than 3 what the licensee, the licensee was talking about a 4 COL item and an ITAAC and you're talking about a 5 carveout.
6 MEMBER BLEY: This is a carveout. They've 7 been working independently on this.
8 MR. SCARBROUGH: Yes. They've been trying 9 to resolve the issue themselves. And they proposed a 10 COL item. We looked at the COL item.
11 We don't have a technical concern with the 12 COL item. We actually think it's a good thing. But 13 in terms of whether or not we could certify the 14 specific aspects, and this is focused right, it's 15 focused on the steam generator tube integrity.
16 And, you know, it's not the whole steam 17 generators. And so, but in this focused area we do 18 not feel we had confidence that we could decide on 19 finality for this particular aspect.
20 MEMBER MARCH-LEUBA: So, from the way you 21 envision the certificate is to have a carveout and a 22 COL. Is that correct?
23 MR. SCARBROUGH: Yes, yes. In discussions 24 when we had first seen their proposed COL item we 25 said, you know, we have our own process for, you know, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 going on carveout.
2 And the response we received, they felt 3 like the COL item was a good thing, right. It was a 4 benefit to their design in terms of what, how they 5 presented their design certification application.
6 And we agreed. But it doesn't 7 specifically affect what we're trying to do here with 8 the finality.
9 MEMBER MARCH-LEUBA: I was going to make 10 another momentous announcement in the fact that I'm 11 happier with the applicant's proposal than with yours.
12 CHAIRMAN SUNSERI: So, let me ask a 13 question. So, let me so, I guess this doesn't make a 14 big difference.
15 But would you envision that when a COL 16 applicant comes in and does what the applicant is 17 saying in the COL item for this activity, would that 18 information be sufficient to resolve the carveout?
19 They would have, I know they would have to 20 license amendment or something like that to get it 21 approved. But is the work that they plan to do for 22 the COL the work that needs to be done to address your 23 safety concerns?
24 MR. SCARBROUGH: Yes. They're very 25 similar because they, if they plan to demonstrate that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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37 1 they are not going to have issues with the potential 2 DWO and the reverse flow the first step from this 3 perspective is to develop a methodology that would 4 predict that reliably.
5 And so, then they would use that. And 6 then, you know, as through and we'll talk a little 7 more about the sections that we have a concern with.
8 But in the design certification they need to have a 9 methodology listed right for all of the various 10 aspects of the design.
11 And this methodology is not ready yet.
12 And so, once they are ready they will use it to, in 13 combination with probably the ANSYS model to show that 14 the stress and the wear on the tubes are not 15 significantly impacted by the DWO and reverse flow.
16 So, again that's the first step that the 17 COL applicant would come in and say here is the 18 methodology and this is how we're going to use it to 19 show that we do not have significant wear on the tubes 20 or damage the IFRs.
21 MEMBER BLEY: I'd like to try something 22 because we haven't dealt with carveouts as such before 23 this time around. It seems to me what we have the 24 applicant has a COL item which will have to be met 25 during the COL.
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38 1 What you're saying is they're saying what 2 they're going to do. You're just saying we haven't 3 reviewed this yet. We have to review it at the COL 4 time.
5 MR. SCARBROUGH: Yes, yes. And that's 6 basically what a carveout is. It's, if we did not 7 have a carveout and we didn't mention it at all 8 officially we have finality on all the aspects of a 9 steam generator.
10 We really don't have the authority to 11 question the steam generator tubes anymore. And so, 12 we're not ready there. We're not there yet, you know.
13 We still want to review the COL item and 14 make sure the methodology is proper.
15 MS. PATTON: I think Mike has something to 16 add.
17 MR. DUDEK: And, Mr. Chairman, just to 18 dovetail into Tom's response is that the COL item is 19 only one small piece of the carveout. I think you'll 20 see that in the upcoming slides is that, yes, they can 21 include the COL item.
22 And it may address one small piece of the 23 carveout. But that doesn't resolve the larger picture 24 of all of the open items that are included. And 25 you'll see they are included in the carveout.
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39 1 CHAIRMAN SUNSERI: Yes, I get it. But it 2 does outline the methodology that would take them 3 there, right?
4 MR. DUDEK: You'll see that's only one 5 small piece.
6 CHAIRMAN SUNSERI: Yes, okay. All right.
7 MR. DUDEK: It's like a first step.
8 MS. PATTON: There is also a little 9 difference in the legal definition between like a 10 carveout versus a COL item. And a carveout makes it 11 very clear that has to be done by the COL.
12 You know, you can rely on a carveout in 13 making the findings and it's a little bit more limited 14 how much reliance we can place on a COL item. And so, 15 we're still working through that and some of the 16 questions on COL item versus carveout.
17 So, I don't want to get ahead of that.
18 But some of those differences are what's being 19 considered in this as well.
20 CHAIRMAN SUNSERI: As Dennis said, we're 21 still learning on this. But when it comes to carveout 22 and I don't like using that vernacular.
23 But is there any timing issues regarding 24 when a license then would be issued or when a licensee 25 would be able to start operating the plant regarding NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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40 1 a carveout or anything?
2 MR. SCARBROUGH: Well, they would need to 3 come in -- the COL applicant would come in and address 4 this aspect of the design that did not reach finality 5 as part of design certification.
6 MEMBER RICCARDELLA: So, doesn't that, I 7 mean a carveout automatically implies a COL item, 8 right? I mean you have to resolve the carveout 9 because it hasn't been, that aspect of the design 10 hasn't been approved.
11 MR. SCARBROUGH: In words or not, right, 12 of course. And so, the COL item that NuScale is 13 proposing is that first step to resolve this issue 14 that's been carved out, exactly.
15 MEMBER MARCH-LEUBA: So, while you're 16 making the presentation can you address my bias. I 17 see opposite to what you said. I see that their COL 18 proposal is broader than your very limited carveout.
19 MEMBER RICCARDELLA: Maybe we need to see 20 the remaining, the additional slides.
21 MEMBER KIRCHNER: All they have proposed 22 is a methodology. You still have to do all of the 23 analysis and have to do the ASME code case, et cetera, 24 et cetera. It's much more.
25 MEMBER BLEY: We've only seen their first NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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41 1 slide. Maybe we could look at some more.
2 MS. PATTON: There are two more slides on 3 the carveout.
4 MEMBER MARCH-LEUBA: I want you to address 5 my biases while you present it because what I hear 6 here is as long as you satisfy the ASME code, the tube 7 doesn't break, we're perfectly okay with it.
8 MR. SCARBROUGH: No. We're only on the 9 first bullet on the first slide.
10 MEMBER MARCH-LEUBA: I have a bias of 11 controllability and moisture in the steam line.
12 MR. SCARBROUGH: Exactly, yes. We're 13 going to get there. So, this is just --
14 VICE CHAIRMAN REMPE: To follow up on 15 Matt's question about how to fix things. I can 16 remember with Vogtle that there was, they let them go 17 ahead and pour concrete for some things but not some 18 nuclear construction.
19 And that was a fuzzy line. When does the 20 carveout have to be addressed? Does it affect what 21 can be done in the construction for a COL applicant?
22 MR. SCARBROUGH: In this case and we are 23 fortunate we had actually two OGC lawyers helping us 24 with this, right, and so, because this is new ground 25 for me too. This is carved out.
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42 1 And this only affects the steam generator 2 tube aspect of the design. Everything else goes 3 forward the way it is supposed to go forward.
4 VICE CHAIRMAN REMPE: Okay. And that 5 would be true for the other carveout too?
6 MR. SCARBROUGH: Yes.
7 VICE CHAIRMAN REMPE: It's just limited on 8 that thing, thank you.
9 MEMBER BLEY: But at the COL stage an 10 applicant could not get a license until these 11 carveouts were fulfilled, reviewed and approved?
12 MR. SCARBROUGH: Yes. This aspect has to 13 be completed, you know, for the COL applicant to 14 receive the COL.
15 MS. PATTON: Right. It basically just 16 identifies the portion of the design that wasn't 17 granted finality through the rule, right.
18 So, it's basically takes a piece that 19 would normally be in a design certification and says 20 the COL when they apply has to provide this additional 21 piece.
22 MEMBER BLEY: But since this is new to us, 23 one last question. Assuming the Commission issues a 24 design certification that rule would then say the 25 following aspects have not yet been evaluated or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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43 1 something.
2 MR. SCARBROUGH: Exactly. We are working 3 with OGC on the exact words. And we're going to sort 4 of show you the words that we're working with OGC to 5 put into the rule itself that will indicate that this 6 specific aspect of the steam generator tubes does not 7 receive finality yet as to OGC license.
8 MS. PATTON: Right. There's a few lines 9 that actually go directly into the rule and carve it 10 out.
11 MEMBER BLEY: We have one or more other 12 carveouts that are going on.
13 MR. SCARBROUGH: I believe there's two 14 other carveouts on different topics.
15 MS. PATTON: That's why I said, there's a 16 little difference in legal definition between like a 17 carveout and a COL items and a carveout, you know, 18 makes it very clear within the rule that needs to be 19 provided.
20 CHAIRMAN SUNSERI: Thank you, thanks for 21 taking us on this little detour of the regulatory 22 practice here. Let's get back into the technical 23 presentation. Go ahead, Mike.
24 MR. DUDEK: Just one more side note.
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44 1 according to OGC has evolved for a COL item.
2 The COL items is being used now is more 3 interpreted as just an information tracking item. It 4 doesn't have any legal gumption or enforcement in the 5 COL going forward. So, it's more of an information 6 tracker versus an enforcement item.
7 MEMBER BLEY: I could get an operating 8 license without fulfilling the COL item?
9 MS. PATTON: We would have to probably 10 have an attorney answer that.
11 MEMBER BLEY: I think so. That really 12 sounds bizarre.
13 MS. PATTON: My understanding is that, my 14 little bit of understanding and, Mike, you can chime 15 in is that there is, more like there could be a 16 potential fight about that a little bit.
17 And this, a carveout makes it, gives it 18 the force of law.
19 MEMBER BLEY: Is the authority here. It 20 brings the strength.
21 MS. PATTON: It's stronger than a COL 22 item.
23 MEMBER BLEY: We've supported a number of 24 design certs under the assumption all COL items --
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45 1 that topic at a different meeting. Okay, thanks.
2 Tom, go ahead.
3 MR. SCARBROUGH: Okay. So, Appendix G is 4 going to be the portion of Part 52 which is the 5 NuScale design certification rule.
6 And so, there's a section, Section 6, I'll 7 call it issue resolution which will talk about the 8 steam generator tube integrity issue and indicate that 9 it's not resolved within the meaning of 5263 Alpha 5.
10 And that, I went back and pulled that out.
11 That has to do with all matters all resolved except 12 for 10 CFR 2.335 which has to do with petitions.
13 So, that's what that has -- basically it's 14 saying that this issue has not been resolved yet for 15 finality for the design certification.
16 And then there is another section that 17 will be in Appendix G, which is Section 4 which talks 18 about what is the COL applicant responsible for. And 19 it will talk about the fact that the COL applicant 20 needs to provide the design information to address the 21 steam generator tube integrity.
22 And so, those sort of two sections that we 23 are working with OGC now to get the words just right 24 from the legal perspective to make sure we carve it 25 out to cover the issues but also, you know, it's only NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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46 1 the steam generator tube area aspect that's being 2 carved out.
3 And so, the rule now, the proposed rule 4 language it's with OGC right now and they're working 5 on it to have it ready for Commission approval. So, 6 that's where we are right now.
7 So, now Becky is going to walk us through 8 -- there is two specific sort of parts to this 9 carveout. And, but we talk about them separately just 10 because it's easier to keep track of.
11 So, Becky is going to talk about the first 12 part.
13 MS. PATTON: Okay. So, currently in the 14 FSAR that NuScale submitted, Section 3912 there's a 15 listing of the computer programs that are used by 16 NuScale for the dynamic and static analyses and for 17 the hydraulic transient load analyses.
18 So, you know, if you look in that 19 currently it will list, you know, NRELAP, for example, 20 as one of those codes. And then, you know, points you 21 over to 1502 for the code description and the V&V.
22 And so, you know, my branch in Reactor 23 Systems assisted, you know, with the review of NRELAP 24 for those, you know, mechanical, those blow down 25 loads.
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47 1 Currently in the FSAR in Chapter 5 it also 2 lists NRELAP as being used for determining the 3 pressure drop in the IFR design to ensure acceptable 4 mass flow fluctuations for power levels, et cetera, et 5 cetera.
6 Our understanding is that, you know, 7 NuScale has plans to, you know, modify that to clarify 8 that. But basically, that's listing currently of 9 NRELAP in 391 is intended for blow down loads 10 currently.
11 That's what the staff had reviewed. We 12 hadn't reviewed it for, you know, other loading 13 conditions potentially for DWO.
14 So, this would be a portion of the 15 carveout to say that 3912 with DWO loads being a 16 potential loading condition you would need to list a 17 method of analysis into 3912 for those loading 18 conditions.
19 And those presently are not there. So, 20 the carveout would specify that in demonstrating steam 21 generator tube integrity a COL applicant would need to 22 provide information to demonstrate that GDC 4 is met 23 for the method of analysis to predict thermal 24 hydraulic conditions of the steam generator fluid 25 system and the resulting load stresses and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 1 deformations from DWO.
2 So, our understanding is that NuScale is 3 planning on, you know, adding some, you know, a COL 4 item for one to this section to specify that would be 5 done in the future. We would still, you know, the 6 current plan is that we would still maintain this as 7 part of the carveout.
8 But that's, basically the first portion 9 would be that method, you know, hasn't been specified 10 and it's integral to the finding in that section made 11 by the Mechanical Branch that all those methods are 12 listed.
13 MR. SCARBROUGH: Right, exactly. So, 14 that's the first part. So, that would -- that's the 15 COL item sort of section.
16 Now the other part is the actual steam 17 generator tube integrity issue. And that sort of has 18 been, I've been to the meetings in the past couple 19 months of the ACRS and heard a lot about that.
20 But the bottom line is NuScale has not 21 provided reasonable assurance that the flow 22 oscillations that occur in the steam generator 23 secondary fluid system will not cause damage to the 24 steam generator tubes directly from DWO or reverse 25 flow or indirectly by possible damage from the inlet NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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49 1 flow restrictors, IFRs where they might vibrate and 2 such.
3 As you saw, they're kind of a cantilevered 4 process. And NuScale talked about their forward flow 5 testing.
6 But they really haven't done really much 7 in the other direction to see if there was something 8 that might cause these to have some issues in the 9 opposite reverse flow direction. And so, that's what, 10 the concern we have there.
11 So, and it sort of -- this issue sort of 12 grew over time because, you know, if you go back to 13 the original Rev 2 of the DCA it indicated in Section 14 5412 that the flow restriction devices would preclude 15 DWO.
16 And then there was Rev 3 which came out 17 that said well, there will be oscillations but they 18 will be within acceptable limits. And as we've gotten 19 more interaction with NuScale in terms of what that 20 really meant and what the information was we 21 determined that we weren't comfortable with the amount 22 of degradation that might occur from reverse flow from 23 DWO and such.
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50 1 was catastrophic failure of a number of tubes could it 2 interfere with the natural circulation process because 3 everything in this reactor relies on natural 4 circulation for cooling.
5 And so, if you had a significant break of 6 a number of tubes you could disrupt natural 7 circulation cooling either from ECCS system which we 8 talked a lot about this week and also the decay 9 removal system.
10 You know, both of those are natural 11 circulation processes. So, that was our concern.
12 Until we are comfortable that there won't be this 13 potential for catastrophic failure because there is 14 GDC 4 which is dynamic effects and vibrations and 15 such.
16 And then there's also GDC 31 which is the 17 fracture prevention of the reactor coolant pressure 18 boundary. And so, and that GDC talks about the fact 19 that you need to have capability to ensure that you do 20 not have a rapid, propagating failure of the reactor 21 coolant pressure boundary.
22 And if you had a number of these IFRs come 23 loose and go through these tubes you might have a 24 number of tubes that fail at the same time. So, we 25 did not feel comfortable that we had enough NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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51 1 information to be able to say that, yes, this issue 2 can have finality.
3 And so, as part of this carveout is a 4 specification, and this would be in the rule itself 5 that a COL applicant will need to provide information 6 demonstrating that 10 CFR Part 100, Appendix A, which 7 is the seismic capability aspect and also Part 50 8 Appendix A, GDC 4 and 31 are met with respect to 9 structural and leakage integrity for the steam 10 generator tubes that might be compromised by these 11 adverse effects from DWO and the secondary fluid 12 system.
13 But we're going to be very clear in the 14 carveout that these are the areas that we're carving 15 out. You know, we're not carving out the entire steam 16 generators and that sort of thing because we have to 17 make sure that we focus it on what the concern was and 18 what is not receiving finality.
19 And that's what is happening right now 20 with the rule that OGC is helping us with. So, that 21 is the two sort of technical issues.
22 So, there's no question, now I was going 23 to have Yuken go through and kind of describe the DWO 24 phenomenon and what's going on with that.
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52 1 had performed the TF-2 tests mainly for thermal 2 hydraulic performance of the steam generators. These 3 tests are also used for flow induced vibration 4 purpose.
5 The TF-2 specimen had five columns of 6 tubes with 250 tubes in total. And one column of tube 7 with 52 tubes was used for the density wave 8 oscillation tests.
9 Density wave oscillation was observed 10 during the TF-2 testing with temperature and flow 11 oscillations in the secondary cooling. The DWO 12 frequency was low and will not excite the steam 13 generator tube structural resonances. Based on the TF-14 2 strength gauge measurements, the staff estimate that 15 the alternating stress intensities will be below the 16 ASME fatigue endurance limits.
17 However, any differences such as geometry, 18 material and operating conditions between the TF-2 and 19 the actual as built steam generators have not been 20 evaluated.
21 As discussed on the next slides the staff 22 is concerned about the potential impact of the density 23 wave oscillation on the steam generator tubes directly 24 and indirectly by the inlet flow restrictors.
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53 1 ask, those strain gauges that you show on the previous 2 slide, were they on the inside or the outside of the 3 tube?
4 MR. WONG: They are on the outside of the 5 tubes.
6 MEMBER RICCARDELLA: Okay. So, they would 7 pick up pressure oscillations. But if there were any 8 thermal gradient effects that would only occur on the 9 inside. It might not, you might not see it on the 10 outside, right?
11 MR. WONG: They pick up the strains as 12 well.
13 MEMBER RICCARDELLA: Not if there was a, 14 if there was a thermal gradient and thus a strain 15 gradient through the thickness of the tube it would 16 not, you know, when you do a thermal shock on a 17 component you get higher stresses on the inside than 18 on the outside.
19 That's a fairly thin tube. But you still 20 might have some through wall gradient.
21 MR. WONG: The tubes are very thin. And 22 from the --
23 MEMBER RICCARDELLA: I understand.
24 MR. WONG: -- what the data indicates it 25 does pick up the strain in this subset. They suspect NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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54 1 some of the strains at the thermal oscillations.
2 The steam generator in the flow 3 restrictors are designed to provide the necessary 4 pressure drop to limit density wave oscillation in the 5 tubes.
6 As explained earlier, the flow restrictors 7 are mounted on the mounting plate and inserted into 8 the steam generator tubes. NuScale performed in the 9 flow restrictor, excuse me, leakage flow instability 10 tests for the conceptual design of the inlet flow 11 restrictors.
12 The staff did not identify any concerns 13 for the test for the normal flow or forward flow.
14 However, these tests did not include density wave 15 oscillation conditions as the forward flow.
16 NuScale has selected a final inlet flow 17 restrictor design that is similar to one of the tested 18 designs. And NuScale will perform validation testing 19 for the final inlet flow restrictor design after 20 design certification.
21 Next slide, please. Unstable density wave 22 oscillation can cause reverse flow to the inlet flow 23 restrictors including subcooled liquid from modest 24 density wave oscillation or slug and two-phase flow 25 for strong density wave oscillation.
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55 1 NuScale has not yet evaluated potential 2 impacts on steam generator tubes and inlet flow 3 restrictors for reverse flow such as fatigue of bolted 4 joints and loose inlet flow restrictors.
5 The concerns due to leakage flow 6 instability cantilever the inlet flow restrictors 7 unless stable under reverse flow conditions. Also, 8 due to cyclic pressure drops and high speed turbulent 9 two-phase flow through the inlet flow restrictors.
10 The concern also includes cavitation 11 erosion of the steam generator tube walls and wear of 12 inlet flow restrictors and the tube walls that can 13 further worsen density wave oscillation.
14 MEMBER BLEY: Excuse me. Two related 15 questions. When you say they're less, the flow 16 restrictors are less stable under reverse flow 17 conditions, what do you mean by that?
18 And my second question is I'm envisioning 19 this thing maybe going back and forth a little bit.
20 And can these screws back out? I've seen screws back 21 out in vibrating situations.
22 And if they do I guess that flow 23 restrictor is free to either flow out or go forward.
24 MR. WONG: Literature indicates when a 25 cantilever structure, when the flow is going from the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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56 1 support end to the free end it's more stable.
2 MEMBER BLEY: So, it's this kind of 3 vibration that you're talking about?
4 MR. WONG: Correct.
5 MEMBER BLEY: Yes, that makes sense.
6 MEMBER MARCH-LEUBA: When the flow is 7 going forward you're pulling. When you're pushing.
8 The pushing is much more -- when you're pulling it 9 straightens out.
10 When you're pushing it moves towards the 11 wall, right.
12 MEMBER BLEY: That makes sense if that is 13 what you're talking about.
14 MR. WONG: Yes, yes. And if the screws --
15 MEMBER BLEY: Let me, I've looked at these 16 things and I kind of assume that you've got a lot of 17 turns on that screw that hold it in place. But that 18 screw is long enough to go through that plate.
19 I don't know how many turns you get. So, 20 I'm -- the idea that a screw could back out might not 21 be crazy.
22 MEMBER RICCARDELLA: It's preloaded, you 23 know.
24 MEMBER BLEY: Yes. I know it's preloaded.
25 But now you're jerking it back and forth.
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57 1 MEMBER RICCARDELLA: Yes, but, you know, 2 theoretically if the preload sustain you don't get 3 oscillatory loads on a preloaded bolt. That's why you 4 preload bolts.
5 MEMBER BLEY: But you preload them under 6 assumptions and this assumption wasn't there.
7 MEMBER RICCARDELLA: Yes. You preload and 8 there's also, typically there's something that keeps 9 it from backing out like in LWR internals they use 10 some sort of retainer device or something to keep it 11 from unscrewing.
12 CHAIRMAN SUNSERI: NuScale said that there 13 would be, you know, loose parts prevention measures 14 applied, right. So, if that's what you're talking 15 about.
16 MEMBER RICCARDELLA: Yes. But that 17 doesn't, if you just contain it as a loose part like 18 you would put a cap over it that doesn't keep it, that 19 doesn't ensure that the preload is maintained. It 20 could still lose preload.
21 It, you know, they're going to be doing a 22 lot of work in this area obviously. That's detailed 23 design work that has to be done.
24 MR. SCARBROUGH: Right. And that's, 25 they're going to have to finish, you know, the design, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 1 pick the final design and then qualify the design.
2 So, there's still quite a bit of work to 3 do to address your issues that you're raising.
4 MEMBER RICCARDELLA: But theoretically, if 5 it's properly preloaded you won't see oscillatory 6 loads.
7 MEMBER BLEY: And one would think after 8 this testing and analysis a consideration of reverse 9 flow would be part of that preloads.
10 MEMBER RICCARDELLA: Yes, for sure.
11 MR. SCARBROUGH: Okay. Next slide, 12 please. So, where do we go from here, okay?
13 Assuming that the design certification 14 rule is issued, the COL applicant will be responsible 15 to address the steam generator tube integrity in its 16 COL application and it has these sort of two parts 17 that we talked about.
18 One is the method of analysis that they 19 have a COL item that's going to make sure the COL 20 applicant knows they have to submit that. And then 21 the second part will be demonstrating that the tubes 22 will not be damaged by DWO directly or by, or 23 indirectly by the IFRs vibrating and things of that 24 nature causing some damage.
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59 1 for demonstrating that in the process of receiving its 2 COL. So, that's going to be a review that the staff 3 will do.
4 And this will all come back to the ACRS 5 for you all to take a look at as well. And then 6 assuming that COL is issued, the next step will be a 7 COL holder.
8 And there's a number of aspects that the 9 COL holder is responsible for. There are ITAAC 10 related to the ASME Boiler and Pressure Code, Section 11 3 requirements.
12 But there also, in addition to that there 13 is the Comprehensive Vibration Assessment Program, the 14 CVAP which Yuken reviews quite a bit in terms of the 15 review for applicants.
16 And there's specific aspects. There is 17 some additional testing. The TF-3 referred to as TF-3 18 testing that has to be done. There's also vibration 19 testing that's specified in Tier 2 in Table 14.272 20 that had to do.
21 So, they have that to do. And plus 22 they're going to have some instrumentation on, for the 23 initial start of a steam generator.
24 So, the COL holder has quite a bit of work 25 to do as well after that phase of receiving the COL.
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60 1 So, that's the process after the design certification 2 to make sure this issue is fully reviewed as part of 3 the next step after design certification.
4 And then Becky is going to talk about next 5 steps.
6 MS. PATTON: Sure. NuScale is currently 7 preparing errata to the Revision 4 of the DCA. And 8 you saw part of that with their proposed COL item that 9 they presented earlier.
10 They are also, you know, preparing some 11 other changes potentially to clarify some of the steam 12 generator secondary fluid flow issues that could 13 impact the tubes, the IFRs, some of the various 14 statements, you know, made in the associated chapters.
15 So, we have prepared drafts for the 16 proposed rule. And it discussed the steam generator 17 tube integrity, the issue as a whole. It includes the 18 method of analysis and as Tom mentioned, the portion 19 of the carveout related to integrity of the IFR and 20 the tubes.
21 So, the draft proposed rule would exclude 22 both aspects of that issue from finality and will, 23 basically what will happen is a COL applicant would 24 have to provide those portions when they apply for the 25 COL and then that's when the NRC staff would perform, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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61 1 you know, that review.
2 Except, I think as noted they could, you 3 know, put a topical report or something together on 4 the method, you know, that could come in ahead of 5 time. If it's a technical report it would typically 6 come with the COL.
7 But either way the COL would, you know, 8 fulfill that by either referencing like an approved 9 topical report, you know, or providing the technical 10 report.
11 So, other aspects of the steam generator 12 design are considered acceptable to staff. Those 13 would be granted finality but not the ones 14 specifically identified in the carveout.
15 MEMBER MARCH-LEUBA: Okay. And that's 16 where my earlier comment was. Apparently the staff is 17 not concerned about controllability and operability of 18 the steam generator?
19 MR. SCARBROUGH: That issue is, we 20 consider, we separated. The design certification 21 focuses on the reactor aspects. The COL applicant 22 still will need to come in and talk about the 23 secondary side, control and things of that nature.
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62 1 on the reactor safety. And our concern was that if 2 there was catastrophic failure of a number of tubes 3 that could affect reactor safety.
4 And so, that's how we separate it. We 5 haven't, it's not that we're not concerned about it.
6 We just have put that over into the COL application 7 review.
8 MS. PATTON: Right. The carveouts are 9 linked to what the findings are that the staff has to 10 make at the design certification stage specifically.
11 So, you know, you can as a finding right 12 that he has to make on that IFR, for example, you 13 know, show it doesn't fall apart and somehow impact 14 the integrity of the tubes or fail to perform its 15 function and therefore you could, you know, have 16 oscillations impacting that.
17 The controllability of the plant, whether 18 or not there are any issues with that, you know, I 19 think if I remember correctly I believe the control 20 system like gets, you know, that gets designed later.
21 I think there's a COL item on some aspects 22 of the MPS control system. So, those are things that 23 would be looked at, you know, at the COL stage.
24 You don't need a, you don't use a carveout 25 for that.
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63 1 MEMBER MARCH-LEUBA: Control and 2 protection system will determine, you're still dumping 3 moisture in the steam line that becomes an issue too.
4 MS. PATTON: Right. But, so some issues 5 have to be, you know, looked at as part of the design, 6 their design, the findings that need to be made, you 7 know, under the regulations.
8 And so, those where you can't make them 9 it's a carveout.
10 MEMBER MARCH-LEUBA: So, you have not made 11 any finding about the controllability and operability 12 of the secondary side?
13 MS. PATTON: No. The control system is 14 part of --
15 MR. SCARBROUGH: That would be a COL item, 16 COL application review not for design certification.
17 MEMBER DIMITRIJEVIC: Okay. Up to now I 18 look at this as operability issue. I did not think it 19 was a safety concern because of your putting, they 20 don't call it steam generator tube rupture but steam 21 generator tube failure. Now when you bring the safety 22 concern isn't that too big to carveout because you 23 cannot even make conclusion that this plant meets 24 safety goal?
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64 1 conclusion in Chapter 19 that this plant is meeting 2 safety goal.
3 MR. SCARBROUGH: Well, we're carving out 4 just the aspect of the steam generator tube integrity 5 aspect.
6 MEMBER DIMITRIJEVIC: Yes. But this is a 7 risk steam generator to fail. That leads to loss.
8 So, you are carving safety concern which can impact 9 conclusions about safety of this plant. How can you 10 do that?
11 So, by making it a, well by making it a 12 carveout for one you're putting it directly in the 13 rule. So, the COL applicant will have to demonstrate 14 that IFR, you know, does remain intact, doesn't, you 15 know, cause damage to the tubes, right, performs its 16 function.
17 That is ensured to have to be demonstrated 18 by the COL applicant by carving that out specifically.
19 So, that's what we would expect.
20 MEMBER DIMITRIJEVIC: But then your 21 Section 19 cannot make conclusions that this plant 22 meets safety goal until that's proved. Just, I just 23 want to say that.
24 Until this is proved by COL applicant we 25 don't know that this plant meets safety goals.
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65 1 MR. SCARBROUGH: The COL applicant will 2 have to demonstrate this to be able to receive 3 permission to load fuel. So, they're going to have to 4 --
5 MEMBER DIMITRIJEVIC: No, no. I 6 understand. But I just say the second sentence in 7 Chapter 19 is this plant meets safety goals with 8 badging, blah, blah, blah.
9 That's not true anymore. It won't be true 10 until they prove that in the COL.
11 MR. SCARBROUGH: We have interacted with 12 OGC on how this process works. And according to their 13 legal opinion you sort of carve that, this very narrow 14 focus out when you make that decision.
15 So, we're going through the process of OGC 16 of what carveouts work. And so far they've indicated 17 that this focused carveout is acceptable from the 18 perspective of you can proceed with design 19 certification with this carveout.
20 So, that's sort of where we are with the 21 process.
22 MEMBER DIMITRIJEVIC: You know, if you 23 think that this is safety concern, you know, it would 24 be tough to agree with that, that you can proceed 25 having such a big safety concern.
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66 1 MR. SCARBROUGH: Okay, well thank you.
2 I'll relay that back to OGC and make sure we're on 3 good legal ground. Thank you.
4 CHAIRMAN SUNSERI: Okay. Any other Member 5 comments?
6 VICE CHAIRMAN REMPE: Okay. Real quick, 7 this has changed in the last few weeks. It's been 8 changed again because we might have done a letter this 9 week and how confident are we in the material that 10 we've only seen in slides?
11 MR. SCARBROUGH: Well, in terms of the 12 carveout I think we're pretty comfortable. We have 13 OGC agreement on how the carveout works and how it's 14 very focused on this specific aspect.
15 So, we're comfortable with this aspect.
16 We don't plan to, this has to go to the Commission of 17 course and they have to, you know, sign out the rule.
18 But we do not plan to have any changes at this point 19 in terms of how the carveout.
20 And it's very consistent with the slides 21 you've seen in terms of the wording. The discussion 22 in the rule is very short.
23 It's very similar to what is in the slides 24 because OGC says you just have to focus it and make 25 sure you that you carve out a very narrow, specific NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 1 concern that you have.
2 So, we don't anticipate any changes. But 3 it does have to go to the Commission for their 4 approval.
5 MS. PATTON: Right. I mean, the 6 Commission, you know, review of the proposed rule 7 always happens afterwards anyway.
8 VICE CHAIRMAN REMPE: So, just trusting 9 and wanted to kick the tires and make sure. Thank 10 you.
11 MS. PATTON: Right. I mean obviously feel 12 free to weigh in one way or another because you're 13 always before the Commission. Bob had --
14 MEMBER BROWN: Yes. You zipped right 15 through something where you said changes in the MPS, 16 Module Protection System. What --
17 MS. PATTON: No, I believe that's, I'm 18 sorry I may have misspoke.
19 MEMBER BROWN: I was hoping you were, 20 okay.
21 MS. PATTON: I believe it's the control 22 system.
23 MEMBER BROWN: Okay. You're talking about 24 the control system for like feedwater control or 25 something like that.
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68 1 MS. PATTON: But my understanding was the 2 control system actually has like a COL item on it.
3 MEMBER MARCH-LEUBA: I said trip but I 4 meant protection of equipment and protection of --
5 MEMBER BROWN: She used the words, the 6 acronym MPS when she zipped right through a comment 7 earlier.
8 MS. PATTON: Yes. I meant to say control, 9 MCS.
10 MEMBER BROWN: Module Protection System 11 is, has nothing to do with this.
12 MS. PATTON: No.
13 MEMBER BROWN: Thank you for the 14 clarification.
15 MR. CALDWELL: This is Bob Caldwell. I'm 16 the deputy director of DNRL. I just want to make 17 sure. But we cannot make a safety finding based on a 18 COL item.
19 We can't say the design is good or bad 20 based on the COL item. It is a tracking item.
21 However, COL items must be addressed during the COL 22 application where we do a review, basically the same 23 SRP type review of what's actually being built with 24 all the final design details in it.
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69 1 will ever be built for that. So, a carveout is very 2 specific. It's very focused. It's on one of the 3 findings.
4 We have multiple findings during our DC 5 review and the certification. So, they are findings 6 by regulation. I'm not familiar that we ever make a 7 finding that the plant is safe.
8 We say that the plant meets the 9 regulations and that all the regulations are satisfied 10 with the exception of an aspect of a regulation. So, 11 we're very comfortable with the COL carveout, excuse 12 me, the carveout process.
13 We're also very comfortable with the COL 14 items. But we can't make a safety finding that the 15 regulations are met based on a COL item.
16 MEMBER BROWN: So, you're confirming 17 Member Dimitrijevic's comment that you can't give a 18 firm basis that it meets the safety goal until, that's 19 why you're saying later? That's what I heard you just 20 say.
21 I'm sorry, I didn't talk to the mic.
22 Vesna noted that how can you give a, say you meet the 23 safety goals, I forgot what the words are, okay, as 24 part of this rulemaking.
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70 1 because of this until the COL applicant completes 2 whatever is necessary on the steam generator design 3 issue. And you all will be reviewing it at that time.
4 You made a comment you can't make a firm 5 commitment that it meets it until you finish this and 6 that's going to be delayed. I'm just trying to 7 confirm what Vesna said that I got it, that first of 8 all they kind of waved their hands.
9 And you're saying well, she's really kind 10 of right. That's the way I --
11 MEMBER RICCARDELLA: I'm not a policy 12 person. But the rulemaking says hey, it meets the 13 safety goals in everything except for these specific 14 areas in which are carved out.
15 MR. CALDWELL: That's correct.
16 MEMBER RICCARDELLA: That's not a big 17 deal.
18 MEMBER BROWN: I didn't say the rule was 19 --
20 MEMBER DIMITRIJEVIC: There is three 21 things core damage large release and conditional 22 containment which this will impact significantly. So, 23 those are three safety goals that come from the PRA 24 perspective.
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71 1 would be me. And, you know, this is not a carveout 2 for the hydrogen, you know line. You are carving out 3 a big part of the thing.
4 I mean, you know, it's not really small 5 item like we were discussing yesterday the hydrogen 6 and, you know, line. So, I mean, I really, you know, 7 I am really, I am not comfortable with this.
8 MR. SCARBROUGH: Okay, well thank you.
9 We'll go back and talk to OGC and make sure that we're 10 on --
11 MR. CALDWELL: Let me just make it clear.
12 Excuse me, this is Bob Caldwell again. For the items 13 of which we determine finality they meet the safety 14 goals.
15 For the items that we have not reached 16 finality on we do not say one way or the other. But 17 for everything that we have reached finality on we 18 have, we believe we meet the Commission's safety 19 goals.
20 MEMBER BROWN: But you won't have finality 21 on this?
22 MR. CALDWELL: We won't have that on that 23 before we actually get the review on the COL for that 24 one aspect.
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72 1 operate until they do.
2 MEMBER BROWN: So, that part I understand.
3 But you need to know how to say, no.
4 CHAIRMAN SUNSERI: All right, Members, any 5 other, I'm sorry, Tom, anything else?
6 MR. SCARBROUGH: No, we're good. Thank 7 you.
8 CHAIRMAN SUNSERI: Members, any other 9 comments or questions for staff while we're in the 10 open session?
11 MEMBER MARCH-LEUBA: Let me put something 12 on the open session. Certainly I like better the 13 approach of the applicant than your approach in the 14 sense that I believe, and this is a belief of religion 15 if you want, that the output of that process will be 16 ending up more validated so we will know for sure 17 whether we are unstable or not.
18 And we will make the changes that will be 19 necessary to the plant so that we won't be unstable at 20 100 percent flow. That's what I believe the output of 21 the COL process will be.
22 And I love it. As I said before, I'm 23 getting tired of winning. So, thank you very much.
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73 1 like to make a statement please come up to the mic and 2 do so.
3 While we're doing this, Mike, can you get 4 the public line open?
5 MR. DUDEK: I will. But just to clarify 6 for members of the public that are on the line we are 7 going to closed session in order to protect 8 proprietary information to the NuScale design is the 9 reason that we're going as announced earlier in the 10 meeting that we can go to closed session to protect 11 proprietary information.
12 We will reopen the line for public 13 discussion or for the public to participate at 1:00 14 p.m. this afternoon when the open session will begin 15 again. Thanks.
16 CHAIRMAN SUNSERI: Anybody in the room?
17 MR. DUDEK: This is Michael Dudek. I just 18 have one additional comment to add on to what you 19 said, Jose. It's not one or the other.
20 I think you're going to get both. So, I 21 think you're going to get NuScale's proposed design 22 fixes and you're going to get the carveout. So, 23 that's just the extra regulatory assurance.
24 MEMBER MARCH-LEUBA: Well, let me 25 reiterate, I'm happier today than I was yesterday.
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74 1 CHAIRMAN SUNSERI: No comments from the 2 room. So, I'll turn to the phone line. Any member of 3 the public on the phone line that wishes to make a 4 statement please state your name and your comment.
5 All right. We're going to close the phone 6 line. And at this point we have reached the end of 7 the open session. We're going to take a 15 minute 8 break.
9 We're going to reconvene at 10 after ten 10 in a closed session with NuScale presenting first. We 11 are recessed until 10:10.
12 (Whereupon, the above-entitled matter went 13 off the record at 9:53 a.m. and resumed at 1:03 p.m.)
14 CHAIRMAN SUNSERI: All right, we are 15 reconvening the meeting now. We will start with 16 NuScale in open discussion to begin with the -- lost 17 my -- rod ejection accident.
18 MR. PRESSON: Matt Sunseri?
19 CHAIRMAN SUNSERI: Matthew, you all are 20 ready to go?
21 MR. PRESSON: Yeah, thank you, and good 22 afternoon. Appreciate you all taking the time to hear 23 from us on these topical reports today. I'm Matthew 24 Presson, Licensing Project Manager for NuScale Power.
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75 1 methodologies for rod ejection accidents, loss of 2 coolant accidents, and non-loss of coolant accidents.
3 The presentations provided today are 4 identical to the presentations we gave to the 5 Subcommittee on February 19, so we'll be moving 6 through them at a pretty quick summary level today.
7 But for interested members of the public, when the 8 transcripts for that February 19 meeting come out, 9 there will be a fair amount more detail there.
10 That being said, while we'll be giving a 11 summary, if you have any questions, feel free to 12 interrupt. And we have our engineers listening in on 13 the phone or one of them here at Rockville, so let us 14 know.
15 CHAIRMAN SUNSERI: All right, thank you.
16 MR. PRESSON: Next slide. All right, so 17 slide 2. For our first presentation on the rod 18 ejection method, it'll by myself up here, and Kenny 19 Anderson is supporting from Corvallis as our Nuclear 20 Fuels Analyst. Next slide.
21 For slide 3, I did want to spend a minute 22 on this just to re-scope, given our week of discussing 23 DCA and FSAR topics here. This slide provides us with 24 a high level map of the technical and topical reports, 25 which develop the methods needed for Chapter 15 and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 1 other related thermohydraulic sections.
2 Today we will be looking specifically at 3 rod ejection LOCA and non-LOCA. And while these do 4 support the NuScale FSAR, the results of these as 5 applied to the FSAR design are presented in Chapter 6 15. Our discussions today will be focused on the 7 separate licensing submittals for these methods.
8 All right, for our agenda, our 9 presentation will cover a quick summary of the event, 10 our acceptance criteria, our expectations against 11 future reg guides, especially DG-1327, a flow chart of 12 the method, how we initialize and evaluate our events.
13 And then a quick summary of that method again.
14 For slide 5, we discuss why we look at a 15 separate method for rod ejection and for meeting our 16 GDC-28 commitments. And it provides a couple of 17 examples on why it's unique insofar as Chapter 15 18 events, such as its focus on nuclear physics instead 19 of thermohydraulics, where that spatial focus is.
20 Postulated causes, and definitely acceptance criteria, 21 which we will also discuss on slide 6.
22 This slide 6 is another summary table 23 providing information on which acceptance criteria are 24 more unique to the rod ejection event than the rest of 25 Chapter 15 events. For the NuScale method, most of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 those acceptance criteria are covered by our method 2 requirements to preclude fuel failure, there we go, 3 and that's part of that footnote down at the bottom of 4 the table.
5 Our next slide, while not applicable to 6 the current method or FSAR DCA design, discusses why 7 we feel pretty comfortable in meeting future proposed 8 criteria for pellet clad interaction. As it is not 9 current criteria, we do not have a full evaluation 10 showing this. But as no exposure is credited in our 11 rod ejection method and as M5 cladding is less 12 susceptible to those interactions in general, we are 13 confident that we won't be challenged when those 14 criteria are revised.
15 So for slide 8, we are looking at a flow 16 chart that shows an overview of our method, how we 17 moved from SIMULATE5 to SIM-3K. And then eventually 18 split it out to look at our peak RCS pressure, our 19 MCHFR, and our fuel temperature and enthalpy 20 requirements.
21 For slide 9, that's a very summary 22 discussion, but it does provide some of the 23 information for how we initiate and set up our steady 24 state assumptions and evaluations. We use SIMULATE5 25 to set up the core response. SIMULATE5 is covered in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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78 1 our nuclear analysis codes and methods qualification.
2 And our design does include the assumption 3 bounding potential for an ejected assembly to damage 4 adjacent assemblies, which has been discussed in terms 5 of our FSAR design. I believe, if my notes are 6 correct, that we and the NRC intend to follow up with 7 that during DCA discussion in the April full committee 8 insofar as the DCA design. For the scope of this 9 method, it is simply an assumption that is built into 10 those initial conditions.
11 Slide 10. Slide 10 shows how we build on 12 from that steady state initialization and move into 13 our dynamic response. SIM-3K is used to model the 14 transient and what's benchmarked to demonstrate a 15 combined neutronic, thermohydraulic and fuel time 16 modeling capabilities. So the slide also lists some 17 of the primary uncertainties that were applied for the 18 simulations.
19 Slide 11 discusses how we move into our 20 CHF evaluation, where we use VIPRE-01. This was 21 originally demonstrated to be appropriate for our 22 design in our subchannel analysis methodology.
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79 1 power distributions, couple of the other bullets seen 2 there. And to that point, we evaluated additional 3 sensitivities to holistically justify those changes.
4 For slide 12, to insure against our fuel 5 heat-up criteria, we include a hand calc, which takes 6 a adiabatic approach, including total energy generated 7 by a SIM-3K, and runs that through either as a 8 temperature or energy increase. Those values are 9 compared against NRC-developed acceptance criteria.
10 And some example values are included in the 11 Subcommittee closed session slides from February 19.
12 Slide 13 looks at the first side of our 13 dynamic system response. So we covered CHF in the 14 previous slide. Our first type of dynamic response 15 that we look at is our CHF evaluation. It takes a 16 transient response and provides those system 17 thermohydraulic conditions over to VIPRE for a 18 subchannel evaluation.
19 Next slide, 14, discusses a quick summary 20 of our second dynamic system response, which is 21 looking for pressurization. For that we, it's a 22 little bit different scenario. We are looking for 23 something that raises the power quickly up to just 24 below those high power and high power rate trip 25 setpoints, and let it go for as long as it takes NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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80 1 before it trips the core.
2 So then from there we calculate the peak 3 system pressure and compare that against our 4 acceptance criteria.
5 So a very quick presentation, but in 6 summary, we have a conservative analysis method for 7 our unique rod ejection accident, at least in terms of 8 Chapter 15 events.
9 And the topical report provides details 10 and justification for software tools and acceptance 11 criteria used, the applicability of the method and 12 those tools, the appropriate treatment of 13 uncertainties, and the results of this application of 14 the method by input to our DCA FSAR Chapter 15. So.
15 MEMBER KIRCHNER: I do have one question.
16 I didn't bring the slides from the previous 17 Subcommittee meeting, but I thought on the slide for 18 fuel that shows the figure fuel enthalpy rise versus 19 oxide wall thickness, you drew a box in within the 20 lefthand figure that you were using for your 21 acceptance criteria.
22 You mention the next-to-the-last bullet, 23 the upper limit that you were using, so I think I can 24 say that. I was curious, I don't remember how you 25 chose a point on the abscissa on oxide wall thickness.
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81 1 Is that a proprietary number?
2 MR. PRESSON: I'll have to ask Kenny if 3 that was a proprietary value.
4 MEMBER KIRCHNER: Yeah, I --
5 MR. PRESSON: But it was based on not 6 needing to basically ever take credit or advantage of 7 any of the space after you pass that point, so.
8 MEMBER KIRCHNER: So that was a box that 9 you drew as your acceptance criteria.
10 MR. PRESSON: Yeah, that's correct.
11 MEMBER KIRCHNER: For the actual NPM, 12 right?
13 MR. PRESSON: Yup.
14 MEMBER KIRCHNER: I'll go back and check 15 on whether that was an open slide or a closed. But 16 again, the basis for that was that that was the 17 estimated maximum oxide oxidation you would see?
18 MR. PRESSON: Correct. And Kenny, if 19 you're available --
20 MEMBER BALLINGER: That's number's a 21 widely used number.
22 MEMBER KIRCHNER: Okay.
23 MEMBER BALLINGER: The one that they use, 24 so.
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82 1 needed to declare that I might have a conflict of 2 interest in certain aspects of this discussion on this 3 particular methodology and limit my participation in 4 such discussions and deliberations.
5 CHAIRMAN SUNSERI: Noted.
6 MR. PRESSON: Kenny, are you available to 7 chat? Because I do believe that value is open 8 information, it just didn't show up on the slide.
9 MEMBER KIRCHNER: Okay.
10 MR. PRESSON: Yeah, you can talk right 11 now.
12 MR. ANDERSON: Hi, this is Kenny in 13 Corvallis. Yes, that number comes from our assumed 14 or calculated maximum corrosion. And it, I think it 15 is on the slide, but perhaps it's not showing up in 16 the presentation.
17 MEMBER KIRCHNER: Yeah, okay. Thank you.
18 MR. PRESSON: Yeah, I'm 99% sure it's not, 19 so. All right, that is the end of our presentation, 20 so if there are any questions.
21 CHAIRMAN SUNSERI: Any members, comments, 22 questions on rod ejection? All right, then we're 23 done. Did that one.
24 MR. PRESSON: All right. Are we 25 presenting this? Yeah. Good?
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83 1 CHAIRMAN SUNSERI: Yes, when you're ready.
2 MR. PRESSON: All right, so next 3 presentation will be a similar summary fashion. And 4 again, same slides as before. This is the, our 5 presentation on our NuScale topical report, loss of 6 coolant accident evaluation model.
7 So here we have myself, Matthew Presson, 8 on the line we have Dr. Pravin Sawant, a Supervisor of 9 Code Validation and Methods. We also have Dr. Selim 10 Kuran, who is our Thermohydraulic Analyst. And Ben 11 Bristol, our Supervisor of System Thermohydraulics.
12 Slide 3 provides a quick overview of our 13 agenda. We describe a very summary version of our 14 methodology, provide a reference slide for our NPM 15 safety systems. There were the four elements of our 16 LOCA topical report and the PIRT, our assessment base.
17 The evaluation model for NRELAP5, and our 18 applicability evaluation. And we discuss how we 19 extend the LOCA evaluation to an IORV event and end 20 with conclusions.
21 So slide 4, little bit of background on 22 the NPM and the LOCA. Some of the unique features 23 involve our integrated design, which eliminates a lot 24 of piping and limits potential breaks. Coolant is 25 captured completely in containment, cooled and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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84 1 returned to the reactor pressure vessel using a large 2 pool.
3 Our regulatory requirements that we use to 4 build our method are, well, that we used to make sure 5 our method met, was the 10 CFR 50.46 acceptance 6 criteria. And we looked to maintain maximum PCT at 7 steady state with no clad heat-up. To meet those for 8 our evaluation method, we used conservative LOCA 9 acceptance criteria. These are figures of merit that 10 the core remains covered, and therefore it collapsed 11 liquid level over the top of active fuel.
12 Our MCHFR is greater than our CHFR limit 13 of 1.29, and our containment pressure and temperature 14 are below the design limit.
15 For slide 6, this provides kind of a 16 roadmap for how we take those acceptance criteria and 17 develop them out into a method. So we start with our 18 10 CFR 50.46 requirements. We then process that using 19 Reg Guide 1.203. And we develop that into our LOCA 20 PIRT Element 1. Use that to develop our assessment 21 base for separate effects testing and integral effects 22 testing.
23 Move on to Element 3, where we developed 24 the evaluation model. And finally, with Element 4, we 25 use all the prior elements to assess that adequacy.
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85 1 Slide 7 is just a quick picture for 2 reference in case any information is needed, but it 3 provides a lot of information about our safety 4 systems, kind of how they're oriented. All right.
5 Slides 8 and 9 get us into Element 1 of 6 our PIRT process. So there we assessed our relative 7 importance of phenomena. We would recognize experts 8 and NuScale subject matter experts in our PIRT panel.
9 And we targeted those figures of merit, CHF, collapsed 10 level above top of active fuel, and containment 11 pressure and temperature. That when we used rankings 12 in importance and knowledge to see where we needed to 13 focus our, any evaluations on.
14 It was a result of that for slide 10, we 15 developed this understanding of phases, Phase 1a 16 blowdown, Phase 1b ECCS actuation, and Phase 2 flow 17 reversal at RRVs. For LOCA, we focus on Phase 1a and 18 1b. We move onto long-term cooling for Phase 2.
19 All right, yeah, for slide 12, it goes 20 into how we develop our NRELAP5 code. We use RELAP5 21 3D, version 4.1.3, as the baseline code. We maintain 22 a code configuration control and development 23 consistent with NuScale's NQA-1 2008 and 2009 NQA 24 program.
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86 1 made for NRELAP5 were to consider NuScale specific 2 components such as our helical coil steam generator.
3 Make sure that we met those regulatory requirements 4 from earlier and apply error corrections as they're 5 determined.
6 Slide 13 is a very high level, but we did 7 want to point out that we have a fair number of tests 8 spanning our integral effects testings and separate 9 effects testing. And for slide 14, we present our 10 NIST-1 facility, where a large portion of those tests 11 took place. It's the primary source of our NuScale-12 specific test data, and it includes a good number of 13 design features that look to scale and provide 14 information for our LOCA and non-LOCA events.
15 All right, so for our NuScale LOCA model 16 overview, we look into the analysis and justifications 17 of why we use NRELAP5, what we need for time-step 18 controls, how we set up those boundary conditions, and 19 how we maintain and treat setpoints and trips. We 20 also take a look at the LOCA break spectrum and dig 21 into the methodology of sensitivity calculations.
22 Those are required by Appendix K, they are 23 phenomena-specific, and we use them to establish a 24 conservative bias.
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87 1 evaluation, we took both the bottom-up and top-down 2 approach. For the bottom-up approach, we identified 3 the dominant models and correlations for the hydraulic 4 phenomena, it's in table 8-1 of the topical report.
5 Identified a lot of key parameters and reviewed those 6 models and correlations. Again, a lot of that is in 7 Chapter 8.
8 For the integral performance, the top-down 9 portion of it, we reviewed the codes and evaluated the 10 integral performance of those codes using those 11 integral effects test data. And we compared that test 12 data to NRELAP5 scalability via scaling and distortion 13 analysis. And we note those differences and 14 distortions between the NPM and NIST and look to see 15 how we can account for them using NRELAP5.
16 So our conclusions for the LOCA method is 17 that there are a number of conservatisms built into 18 it. We have both as much from 10 CFR 50, Appendix K, 19 as is applicable to the NuScale design. And we look 20 to make sure that those other unique considerations 21 are considered by other methodology conservatisms.
22 We developed this using the cycle 23 independent bounding LOCA analysis. It is supported 24 by an extensive experimental database. A lot of those 25 new to NuScale using this one, as well as several NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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88 1 others. Applicability evaluation is consistent with 2 Reg Guide 1.203, and we maintain -- we look to 3 maintain those figures of merit.
4 So CHF is not challenged, our collapse 5 level in the reactor remains above the top of active 6 fuel. There is no clad or fuel heat-up, and our 7 pressure and temperature remain below design limits.
8 And the next slide, slide 20, really slide 9 21, go into how we kind of extend our LOCA into IORV 10 space. So we're looking to kind of evaluate liquid 11 space, RRV and steam space, RVV and RSV discharge.
12 And these are fairly similar transients to the LOCA.
13 From that, we followed a very similar 14 process as our LOCA, developing the method. And yeah, 15 next slide. On slide 22, we account for a couple of 16 the differences. The main difference is our key 17 acceptance criteria, our MCHFR limit moves to 1.13 and 18 1.37.
19 And our conservatisms are the same as 20 LOCA, but with the following exceptions. That we 21 remove an additional 15% bias in fuel. We have our 22 limiting axial power shapes and radial peaking based 23 on subchannel analysis. The Moody choked flow model 24 for two phase is applied to the initiating valve, and 25 the initial conditions are biased to minimize MCHFR.
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89 1 So on slide 23 we come to similar 2 conclusions. IORV as an extension of the LOCA method.
3 Maintains its own PIRT assessment and applicability 4 within the LOCA. The minor method differences mainly 5 account for the AOO classification of that.
6 And MCHFR occurs early within that 7 transient, and then rapidly rises, given the flow-to-8 power ratio. So our primary concern there, that the 9 collapsed liquid in the RPV does remain above the top 10 of active fuel.
11 MEMBER MARCH-LEUBA: The MCHFR occurs 12 early but does not violate the limit, right?
13 MR. PRESSON: Correct.
14 MEMBER MARCH-LEUBA: Because the way you 15 have it written, I said, wait a moment.
16 MR. PRESSON: Yeah, well, and it is still 17 the minimum, or maximum but it does not violate --
18 MEMBER MARCH-LEUBA: I know exactly what 19 you mean, it can be misinterpreted.
20 MR. PRESSON: Yup. And that is our LOCA 21 presentation.
22 CHAIRMAN SUNSERI: Members, any comments 23 or questions for NuScale? No? All right. So you may 24 proceed with the non-LOCA, 25 MEMBER BLEY: It just strikes me that if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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90 1 I were listening in, I would think we have no 2 interest. But we had a Subcommittee meeting on this 3 where we delved into the associated issues in great 4 detail.
5 CHAIRMAN SUNSERI: That's a good point, 6 and we had good, full Committee participation at those 7 subcommittees as well.
8 MEMBER MARCH-LEUBA: And it was two days 9 ago.
10 CHAIRMAN SUNSERI: Yes.
11 MEMBER MARCH-LEUBA: So that's why we're 12 so quiet, because this is just a pro forma 13 presentation.
14 MR. PRESSON: Yeah, two days ago for 15 Chapter 15 and two weeks ago for the original 16 Subcommittee for this. But those transcripts aren't 17 up yet.
18 CHAIRMAN SUNSERI: But it's important to 19 get it on the record for public --
20 MR. PRESSON: Yeah. There was a good full 21 day of conversation on this.
22 VICE CHAIRMAN REMPE: Good, huh?
23 MR. PRESSON: I would say so, yeah. Hey, 24 it's nuclear industry, we value a questioning 25 attitude.
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91 1 All right, and our final presentation for 2 this afternoon is on the Non-Loss of Coolant Accident 3 Topical Report. Again, presenters are myself up here, 4 as well as Megan McCloskey, who is our Thermohydraulic 5 Analyst. We have Ben Bristol on the line, who is our 6 Supervisor of System Thermohydraulics, and Paul 7 Infanger, our Licensing Specialist is in the audience 8 as needed.
9 So for slide 3, we go over our outline 10 where we, just the outline of the presentation. We 11 give a scope of the non-LOCA LTR as compared to other 12 Chapter 15 events, as well as other FSAR events. We 13 discuss those non-LOCA events that are covered in the 14 method. We discuss the development of our non-LOCA 15 method and give a general overview of how we perform 16 those analyses and look at a couple of specific 17 events.
18 So slide 4, discussing scope. Our non-19 LOCA method does look at NRELAP5 system transient 20 analysis of non-LOCA events. It looks at that 21 interface to subchannel and accident radiological 22 analysis. And goes over the short-term transient 23 progression with DHRS cooling. So what is out of 24 scope for the non-LOCA method is the SAFDLs, which are 25 evaluated and downstream subchannel analysis, with its NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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92 1 own topical report.
2 All of these out-of-scope items are either 3 captured in topical reports or technical reports.
4 Also includes accident radiological dose analysis, 5 control rod ejection, which we already covered, as 6 well as LOCA, and those IORV events. Peak containment 7 pressure has its own technical report. And the long-8 term transient is covered in the long-term cooling 9 technical report.
10 So our non-LOCA evaluation method is 11 applicable to the following events. We covered 12 cooldown events, heat-up events, reactivity events, 13 inventory increase and inventory decrease. Most of 14 these are fairly standard events for Chapter 15, but 15 a couple of unique ones for NuScale giving our design 16 our loss of containment vacuum and containment 17 flooding. As well as the heat-up event of an 18 inadvertent operation of DHRS.
19 A quick overview of non-LOCA event 20 acceptance criteria. This table presents those 21 criteria in general, so for the minimum critical heat 22 flux ratio and the maximum fuel center line 23 temperature, you'll note that both of those point to 24 the Footnote 1, where we, that was pretty much as 25 collapsed down to the same AOO acceptance criteria.
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93 1 And here we have a zoom-in on how our non-2 LOCA method interacts with those other topical 3 reports. We have our, you developed a design, you 4 look at the events. Our non-LOCA methods covers the 5 system thermohydraulic response. That then passes 6 that information on to VIPRE for subchannel analysis, 7 looking at CHF.
8 And then mass and energy releases from the 9 thermohydraulic response and other inputs are looked 10 at in our accident radiological analysis, which is 11 bounded by our accident source term topical report.
12 MEMBER MARCH-LEUBA: And this might be 13 relevant for some other topic, but not every single 14 transient evaluated within RELAP gets evaluated with 15 VIPRE.
16 MR. PRESSON: Correct.
17 PARTICIPANT: You use screening criteria.
18 MR. PRESSON: Yup.
19 MEMBER MARCH-LEUBA: Say two words about 20 it?
21 MR. PRESSON: Yeah, we'll actually cover 22 that on a later slide, but that is correct, yeah. For 23 slide 8, we look at our margin to acceptance criteria.
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94 1 release, and establishing those safe, stable 2 conditions to pass on down.
3 For slide 9, the evaluation method 4 development follows a fairly similar path as our LOCA.
5 Followed the same Reg Guide 1.203 process in 6 developing the graded approach. Element 1 is looking 7 at establishing the applicable transients and 8 acceptance criteria and to create that non-LOCA PIRT.
9 Elements 2, 3, and 4 leverage a fair 10 amount of information from LOCA, but it definitely 11 does focus on the differences between high ranked 12 phenomenon, well, the differences between the LOCA and 13 non-LOCA high ranked phenomena, make sure that we have 14 additional NRELAP5 code validation performed to focus 15 on, for example, DHRS and the integral non-LOCA 16 response.
17 Slide 10 covers the results and what was 18 considered in our non-LOCA PIRT, including the general 19 categories of event types, the SSCs that were 20 considered, as well as the phases that are part of our 21 non-LOCA, our pre-trip transient, our post-trip 22 transition, and finally Phase 3 of stable natural 23 circulation.
24 Slide 11 gives a quick summary of 25 NRELAP5's applicability for non-LOCA. As mentioned NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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95 1 before, there was a KATHY analysis performed to 2 determine how to address those high ranked phenomena, 3 looking to see what validation was still applicable, 4 as taken from the LOCA evaluation model and adding 5 additional validation and benchmarks for non-LOCA.
6 That also looked to our conservative 7 inputs and make sure that we had suitable subchannel 8 analysis established. Sorry, yeah.
9 Overall conclusion is that the NRELAP5 10 code with the NPM system model is applicable for 11 calculation of the NPM non-LOCA system response, so.
12 Slide 12 goes over that analysis process.
13 Topical report section 4, where we develop that plant 14 base model. We adapt it as needed for the specific 15 events. You perform you steady state and transient 16 calculations within RELAP5, and you evaluate those.
17 You confirm your margins to RCS pressure acceptance, 18 steam generator pressure acceptance criteria.
19 And you, this kind of goes to your point 20 earlier, you identify the cases that you look to 21 examine further with subchannel analysis and extract 22 the boundary conditions as applicable. So we're 23 looking conservative bias directions of maximal 24 reactor power, core exit pressure, core inlet 25 pressure, minimum RCS flow rate.
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96 1 And the NRELAP5 CHF calculations for non-2 LOCA may be used as a screening tool to assist 3 analysts in determining limiting cases to be evaluated 4 in that downstream subchannel analysis of that CHF.
5 It's not itself used for those non-LOCA events.
6 So, and 6, you look to identify if any 7 applicable radiological analysis needs to be 8 performed.
9 MEMBER MARCH-LEUBA: How do you identify 10 the step 6, what do you use as criteria?
11 MS. McCLOSKEY: For the events with 12 downstream radiological analysis, we look at the 13 system transient response and which cases have the 14 maximum mass release, which would carry the 15 radioactivity and increase the dose. And the maximum 16 iodine spiking time between reactor trip and isolation 17 of the break.
18 MEMBER MARCH-LEUBA: But if all your 19 analyses show no clad damage, what do you do?
20 MS. McCLOSKEY: Is the question why do we 21 do it, or what do we do?
22 MEMBER MARCH-LEUBA: What do you do if you 23 run all of your transients and none of them results in 24 clad damage? So your core is intact.
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97 1 conditions to the radiological analysis, and they use 2 an appropriate source term based on, I think, and I am 3 not radiological analysis analyst, you've got tech 4 spec limits on fuel failure rates and normal operating 5 coolant that can be --
6 MEMBER MARCH-LEUBA: So you assume normal 7 operation failure rates, and that is what is gives you 8 the source term.
9 MS. McCLOSKEY: Again, I'm not an expert 10 on the radiological analysis of what they used for the 11 source term.
12 MEMBER MARCH-LEUBA: Okay, I don't 13 remember, but that sounds familiar.
14 MS. McCLOSKEY: But there are source terms 15 that are evaluated.
16 MEMBER MARCH-LEUBA: It was like one --
17 yeah.
18 MR. PRESSON: From tech spec 19 concentration, just got a note, so.
20 Slide 13 looks at our general methodology 21 and event-specific methodology. In general we're 22 looking at steady state conditions, our treatment of 23 plant controls, loss of power, single failure, making 24 sure we have bounding reactivity parameter input. And 25 then bias the other parameters as needed. And we also NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 1 look at operator action as needed.
2 For the event-specific methodology, we 3 then dive a little deeper into the description of the 4 event initiation and progression. And we make sure we 5 appropriately scope for the acceptance criteria of 6 interest and target limiting single failures, the loss 7 of power scenarios, and whether or not we need 8 additional sensitivity calculations. The initial 9 condition biases and conservatisms that already 10 existent, or if we need, again, to perform more 11 sensitivities.
12 And then tabulated representative results 13 of those sensitivity calculations. So, and those 14 sample analysis results are provided in Section 8 of 15 the non-LOCA method.
16 So, for conclusions, slide 14. Our non-17 LOCA system transient evaluation model is developed 18 following that graded approach we discussed in 19 accordance with guidance provided in Reg Guide 1.203.
20 It applies to NPM-type plant design, natural 21 circulation water reactors with helical coil steam 22 generators and an integral pressurizer.
23 NRELAP5 is used to simulate those systems 24 thermohydraulic responses to demonstrate primary and 25 secondary pressure acceptance criteria are met, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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99 1 that safe and stable conditions are achieved. And 2 system transient results provide the boundary 3 conditions that are then passed down to our subchannel 4 methods and radiological analyses.
5 And that concludes our non-LOCA.
6 CHAIRMAN SUNSERI: Members, any questions 7 or comments for NuScale?
8 MEMBER MARCH-LEUBA: Not today.
9 CHAIRMAN SUNSERI: Okay, well, good, we 10 appreciate the recap and the presentation. So at this 11 time we can transition over the staff for their 12 comments.
13 So as the presenters are taking their 14 seats, I'll turn to Rebecca and ask if you have any 15 overarching remarks that you want to make at this 16 point.
17 MS. PATTON: No, just thank you.
18 CHAIRMAN SUNSERI: Because I skipped you 19 earlier today. Okay, so Bruce, are you the lead here?
20 All right, well, whenever you're ready.
21 MR. BAVOL: All right, good afternoon, 22 everybody, my name is Bruce Bavol, I'm the Project 23 Manager on the NuScale project. This afternoon from 24 the NRC staff we're going to be talking several 25 topical reports, the first being rod ejection, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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100 1 second being loss of coolant accident analysis, and 2 the third non-loss of coolant analysis.
3 To my right, Chris Van Wert will be 4 leading the rod ejection. We're, since we've talked 5 a lot about these topics, I'm just going to move right 6 into the staff review and turn it over to Chris.
7 CHAIRMAN SUNSERI: Yeah, and I think 8 similar kind of comments. I mean, there's not, you 9 don't have to read every bullet on the slide, we're 10 well versed in the topic to hit the high points and 11 the important message that you want to leave us with.
12 MR. VAN WERT: All right, good afternoon, 13 this is Chris Van Wert. And since we're jumping here 14 into the review, just want to point out that what is 15 included and not included within the review, we did 16 look at the criteria and the methodology as a whole, 17 as well as the assumptions that went into it.
18 And it's worth noting that the analysis 19 itself for the DCA is not part of this review, that is 20 handled separately under the Chapter 5 staff 21 evaluation report. It's also worth noting that the 22 staff did audit calculations and other supporting 23 information during its review.
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101 1 core coolability, and fission product inventory. And 2 we did determine that they either followed the 3 guidance provided in SRP 4.2's Appendix B, or were 4 conservative compared to it.
5 And as we discussed during the 6 Subcommittee, it was also not part of the staff's 7 review, but we were cognizant of the draft guidance 8 that's out there in terms of revised guidance for rod 9 ejection accidents.
10 And we did compare the two to see where 11 NuScale fell within it. But again, since that's draft 12 guidance, that wasn't a criteria that they had to 13 follow. But they were conservative in regards to 14 either criteria.
15 So next was the evaluation of the code 16 suite. In terms of rod ejection, they used CASMO5 to 17 SIMULATE5, you know, SIMULATE-3K and RELAP5 and VIPRE.
18 Most of those, with the exclusion of SIMULATE-3K, were 19 already reviewed and approved as part of another 20 topical report, the nuclear analysis codes and 21 methods, so that was not part of this review.
22 However, SIMULATE-3K was unique to this 23 and the validation was contained within it, so the 24 staff's review did cover it. And we did determine 25 that they successfully demonstrated that they could NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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102 1 use it properly and get accurate results.
2 MEMBER MARCH-LEUBA: Has SIMULATE-3K been 3 licensed by any other vendor or facility?
4 MR. VAN WERT: So SIMULATE-3K has been 5 used in licensing actions and has been reviewed by the 6 staff. It has not been submitted by Studsvik as the 7 standalone methodology topic report. So there's no 8 generic, yeah.
9 MEMBER MARCH-LEUBA: But any licensee or 10 vendor?
11 MR. VAN WERT: Licensees have submitted.
12 MEMBER MARCH-LEUBA: Some licensees use 13 it?
14 MR. VAN WERT: Yeah.
15 MEMBER MARCH-LEUBA: Okay, good.
16 MR. VAN WERT: For plant cycle, this 17 attribute did include plant cycle assumptions used by 18 NuScale. And in general, they included ranges and 19 power and cycle time and range of operating conditions 20 and show that they used limiting conditions.
21 The staff also agreed that the assumptions 22 in terms of the automatic systems response of non-23 safety systems were conservative, and that the 24 methodology regarding timing of loss of AC power 25 conservatively biases the RCS pressure evaluation.
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103 1 The staff reviewed the methodology itself, 2 including how information is passed between the 3 different codes, the uncertainties, the modeling 4 assumptions, and the handling of reactor trips. And 5 in conclusion, the staff determined that they were 6 conservative and that the methods were acceptable for 7 demonstrating compliance with the acceptance with 8 acceptance criteria.
9 And in conclusion overall, the staff 10 concludes that the criteria used for evaluating REA 11 either follows or is more conservative than the staff 12 guidance, and that the methodology accounts for 13 various potential operating conditions in time in life 14 and conservatively addresses uncertainties in plant 15 conditions.
16 The staff therefore finds the use of this 17 topical report acceptable for evaluating reactivity-18 initiated accidents from the NuScale plant design.
19 And if there are any questions? And if 20 not, pass it on to Shanlai Liu.
21 CHAIRMAN SUNSERI: Members? No, all 22 right. Continue on.
23 DR. LU: Okay, Shanlai Lu from the staff, 24 NRR.
25 Okay, right away jumping to the -- okay, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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104 1 the review team might have skipped that, so you 2 already talked about that one.
3 So the design features of course that you 4 guys have already gone through one. Very simple 5 design, there are three reactor vent valves on top of 6 the reactor vessel, two reactor, you know, return 7 valves. And then containment functions as a part of 8 ECCS.
9 So the scope of this topical report of 10 course is number one, it's to underline the LOCA. And 11 then as a part doing part of the review process, they 12 extended this topical report to cover the IORV 13 methodology. And as part of it, it also supports the 14 peak containment pressure and non-LOCA topical report 15 and non-term cooling analysis models.
16 Applicable regulation for LOCA of course 17 10 CFR 50.46. They decided to use Appendix K, which 18 does give them some flexibility to reduce the number 19 of runoffs that don't have to do the best estimate a 20 whole bunch of statistical sampling. Okay, next 21 slide.
22 The review approach, and we did take an 23 early engagement and, so that we can -- we conducted 24 extensive audits, all the way to, you know, a couple 25 months before this some presentation. And because of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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105 1 that effort, and then we only identified a total 2 number of 13 RAIs, which is 45 RAI -- but through the 3 process we resolved 210 other issues.
4 And those were really resolved based on 5 extensive staff sensitivity studies and based on 6 NRELAP5 confirmatory analysis with TRACE, thanks for 7 our research support.
8 And the primary and, you know, scope of 9 this review is a focus on LOCA and a non-LOCA too.
10 And related to IORV. So the review area number one is 11 PIRT. And based on the staff's review, we conclude 12 that the PIRT process they had followed the CSAU 13 methodology.
14 And we used NRELAP5 code, which is a 15 derivative of NRELAP-3D, which has been used 16 extensively before. But they did add additional NPM 17 special features. We went through all each features 18 before.
19 And in order to confirm and then benchmark 20 the code, they conduct the extensive testing which 21 lasted a very long time, actually, more than ten 22 years. And then they also performed the scaling 23 distortion analysis. We reviewed that one, identified 24 the issues, and they did additional testing. And, 25 which resolved the issues there too.
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106 1 And as part of IORV analysis methodology, 2 and then we dived very deep into the actual CHF 3 correlations by our staff. And what is used for low 4 and low flow and high flow conditions, including the 5 STERN and the KATHY facility specific fuel databases 6 they used for AOOs. So those are review areas we 7 covered. Next slide.
8 And as I mentioned that we did extensive 9 staff confirmatory analysis, which covers the separate 10 event test and the integral effect test, extensively 11 on the NIST models itself. And we used both TRACE and 12 a RELAP5 code, and more than 55 sets of calculation 13 were performed.
14 And because of all the effort, we were be 15 able to resolve a lot of the, you know, audit issues.
16 So we can zoom in to the RAIs, like total questions 17 are only 45. Those are the confirmatory analysis.
18 Based on the review, we concluded at the 19 end NuScale LOCA EM model. And RELAP5 version 1.4 20 approved for determining critical heat flux and 21 collapsed liquid level for NuScale NPM in compliance 22 with 10 CFR 50.46 key requirements.
23 And the code is, can be used to determine 24 the peak containment pressure, but with the limitation 25 that they have to apply certain specific peak NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 containment pressure analysis criteria there. And the 2 CHF model is approved, subject to limitations and the 3 conditions for low flow and the high flow conditions.
4 So with that, that's the conclusion of 5 staff's presentation on LOCA topical report. All 6 right.
7 CHAIRMAN SUNSERI: Comments from members?
8 All right, Alex, your turn.
9 MS. SIWY: Is this the one that doesn't 10 work?
11 CHAIRMAN SUNSERI: Yeah, I'm sorry, use 12 the one to your right.
13 MS. SIWY: Okay, all right. My name is 14 Alex Siwy and I'm a Technical Review in the Reactor 15 Systems Branch in NRR. To provide a basic summary of 16 the staff's review process, we conducted our review of 17 the non-LOCA topical report in accordance with the 18 applicable NRC regulations and guidance. Our SER is 19 based on Revision 2 of the topical report.
20 The staff conducted audits similar to what 21 was done for LOCA, two audits in four different phases 22 that covered different topics. We examined about 140 23 different issues as part of the audits, and overall, 24 the audits really helped to confirm the staff's 25 understanding of the docketed information and to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 1 inform RAIs.
2 In total, we issued 33 RAI questions, and 3 to date all of these have been resolved and responses 4 have been incorporated into the topical report as 5 appropriate.
6 So this slide covers the scope of the non-7 LOCA methodology, which NuScale covered well in their 8 presentation. I think the thing that I would 9 highlight here is that some of the items that are 10 discussed in the topical report the staff is not 11 making conclusions on as part of the topical report 12 review, because we feel that those items are more 13 appropriate for a design-specific application of the 14 methodology. These include items like the limiting 15 loss of power assumptions and single failures.
16 One of the major areas of staff review 17 were the key design features and models that would be 18 particularly relevant for non-LOCA event analysis.
19 The staff reviewed things like the natural circulation 20 design, the helical coil steam generator models, the 21 DHRS modeling, and the fact that the evacuated 22 containment vessel produces the potential for a new 23 type of event.
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109 1 transient analyses. As the applicant discussed, they 2 developed the non-LOCA EM based on the LOCA EM using 3 a grade approach. The staff reviewed the applicant's 4 non-LOCA PIRT to ensure that the important phenomena 5 were identified and appropriately captured in the non-6 LOCA topical report.
7 And the staff reviewed how the applicant 8 addressed each of the highly ranked non-LOCA 9 phenomena, which included methods such as separate and 10 integral effects tests, code-to-code benchmark, use of 11 bounding input values, as well as other analysis 12 methodologies.
13 Related to this topic was one significant 14 issue that we encountered as part of our review. In 15 particular, the staff requested additional 16 justification for how multidimensional flow effects in 17 the RCS and thermal stratification in the reactor pool 18 are addressed as part of the non-LOCA EM. The staff's 19 major concerns on this topic were the potential for 20 reduced RCS flow rates, as well as degradation in DHRS 21 performance.
22 To summarize, the applicant's RAI response 23 resolved the issue, as was confirmed by the staff 24 audit of underlying calculation notes, as well as 25 audit discussions with the applicant.
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110 1 The staff reviewed each of the NRELAP5 2 assessments against test data presented in the non-3 LOCA topical report, as well as a couple that were 4 presented as part of the LOCA topical report. And 5 overall, the staff finds that the KAIST, the NIST HP-6 03 and HP-04 tests served to validate the NRELAP5 DHRS 7 models.
8 The SIET TF-1 test validated the steam 9 generator secondary side phenomena, but the staff had 10 some concerns about the ability of the SIET TF-2 test 11 to fully validate primary to secondary heat transfer.
12 The NLT2A, 2B, and 15P2 integral effects 13 test together demonstrate the applicability of NRELAP5 14 to evaluate non-LOCA transients. And the benchmark 15 against RETRAN-3D provides confidence that the NRELAP5 16 point kinetics model with the thermohydraulic feedback 17 produces results that are consistent with those of an 18 NRC-approved code.
19 There were a couple of significant review 20 issues related to the assessment against NRELAP5, or 21 assessments of NRELAP5 against test data.
22 In particular, the applicant removed steam 23 generator and DHRS heat transfer biases from the 24 methodology in response to staff questions about the 25 steam generator heat transfer uncertainty based on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 1 SIET TF-2 concerns that I mentioned on the previous 2 slide. And this was associated with the DCA Chapter 3 15 UOI, as well as concerns about DHRS nodalization.
4 To address these concerns, the applicant 5 provided justification that non-LOCA figures of merit 6 are not sensitive to these biases. And based on its 7 review of the justification, as well as audits of the 8 underlying calculations, the staff finds that the 9 removal of the DHRS and steam generator heat transfer 10 biases is supported for NPM model Revision 2.
11 But we did impose a related limitation and 12 condition because some of the sensitivities were 13 specific to the particular design at hand.
14 The staff also reviewed the general and 15 event-specific non-LOCA methodology. Overall, the 16 process for analyzing non-LOCA events, including the 17 interfaces with other methodologies, provides an 18 acceptable analysis framework. The staff also finds 19 that the deterministic approach using conservative or 20 bounding inputs, initial conditions, and assumptions 21 is acceptable for conservative calculations of non-22 LOCA events.
23 In addition, the staff reviewed each of 24 the event-specific methodologies and concluded that 25 the application of those methodologies will ensure NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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112 1 conservative results.
2 And finally, the staff reviewed the 3 representative non-LOCA event calculations in Section 4 8 of the topical report and concludes that they 5 adequately illustrate how the non-LOCA methodology can 6 be applied to conservative transient analyses.
7 This slide just summarizes the limitations 8 and conditions found in the staff SER. I won't go 9 through them line by line, but there are six different 10 limitations and conditions.
11 And in conclusion, the staff finds that 12 all technical issues from the course of the review 13 have been resolved and that the use of NRELAP5 with 14 the non-LOCA methodology described in the topical 15 report is acceptable for the non-LOCA safety analyses 16 of the NuScale NPM design, subject to the specified 17 limitations and conditions.
18 CHAIRMAN SUNSERI: Very good, thank you.
19 Members, any questions or comments?
20 MEMBER KIRCHNER: I just would like to 21 thank NuScale and the staff for their very good 22 presentations during our February Subcommittee 23 meetings and their excellent short summaries today.
24 Thank you.
25 CHAIRMAN SUNSERI: Any other comments?
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113 1 All right, so we'll ask if there are any members in 2 the room that would like to make a comment. And while 3 we're doing that, if we can open up the phone lines 4 for public comment.
5 MR. PRESSON: Hey, Matthew Presson with 6 NuScale. I wanted to confirm for you that the 100 mil 7 corrosion limit is indeed non-proprietary. So good to 8 use.
9 MEMBER BALLINGER: It's only a 100 -- it's 10 a 100, not 80?
11 MR. PRESSON: That is what was emailed to 12 me, yes.
13 MEMBER BALLINGER: Okay. All right.
14 Eighty has been around for the last 15 years or 20 15 years.
16 CHAIRMAN SUNSERI: All right, there's no 17 comments from the room, so we'll turn to the phone 18 line. If there is a member of the public that is on 19 the phone line that wishes to make a comment, now is 20 your opportunity. Please state your name and provide 21 your comment.
22 MR. LEWIS: Marvin Lewis.
23 CHAIRMAN SUNSERI: Okay, Marvin, we'll 24 take yours first.
25 MR. LEWIS: Wonderful, thank you. Look, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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114 1 it sounds reasonable -- saying that it's going to mean 2 that the reactor will operate without problems. But 3 at least the verbiage sounds good. I do have a 4 question, mainly about density waves fluctuate --
5 density wave oscillations.
6 When you get the water hammer and 7 everybody runs out of the nuclear power plant, how do 8 you know there's going to be enough people left to 9 handle the -- resume without emergencies, thank you.
10 CHAIRMAN SUNSERI: Thank you for your 11 comment. Ms. Fields, I think you're next.
12 MS. FIELDS: Yes, this is Sarah Fields.
13 I brought this up at the NuScale Subcommittee meeting 14 a few days ago. I do not understand how the NRC will 15 be finalizing the draft rule and submitting it to the 16 Commission on March 19, which is two weeks from now.
17 And then the NRC intends to publish rulemaking effort 18 on June 1.
19 There's still a few things to iron out 20 between the ACRS, NuScale and the NRC that have been 21 discussed over the past few days. The ACRS won't 22 finalize their -- or submit their final letter until 23 June 23, I believe. And then the NRC staff won't 24 finalize the SER until November. And yet the NRC 25 appears to be going ahead with this rulemaking as if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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115 1 all the T's have been crossed and the I's have been 2 dotted, which they haven't.
3 So I think the NRC's schedule for this 4 rulemaking is rather premature. Also, there really is 5 no rush. The prospective COL applicant, the only 6 prospective applicant, is the Utah Associated 7 Municipal Power Systems, or UAMPS.
8 The type of reactor that UAMPS intends to 9 construct and operate would have 25 more percent power 10 than the current NuScale design. Therefore, UAMPS 11 must wait until the NuScale -- after NuScale submits 12 its standard design approval application, which would 13 include that 25% power increase, before they could 14 submit their COL application to the NRC. And the 15 NuScale SDA application's not expected until the 16 latter part of 2021.
17 So basically, there really is no COL 18 applicant out there who will be submitting an 19 application specifically referencing this design 20 certification. So I just wanted to put that out 21 there. I think that the public should be able to wait 22 until all ACRS and NRC staff documents related to this 23 design certification are complete before the 24 rulemaking. Thank you.
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116 1 members of the public on the phone line that wish to 2 make a statement? Okay, we will close the phone line 3 at this point, thank you. And we're at a transition 4 point here. Let me poll the Committee here. Do we 5 see the need for a closed session to talk to staff or 6 NuScale about any proprietary information?
7 MEMBER MARCH-LEUBA: I recommend that we 8 go into closed session to read the letters for 9 proprietary content, so NuScale can tell us they're 10 not proprietary. And then we go back to open session 11 to discuss them.
12 VICE CHAIRMAN REMPE: But we should be all 13 done with the transcriber.
14 CHAIRMAN SUNSERI: Yeah, we can do that --
15 MEMBER MARCH-LEUBA: Off the transcript.
16 CHAIRMAN SUNSERI: Yeah, off the. Well, 17 we're going off the record anyway at this point in 18 time. So I think we'll proceed along that, those 19 lines. Walt, is that okay with you?
20 All right, so we are going to go off the 21 record at this point in time. The next time we will 22 be on is at 10:45 tomorrow morning when we'll look at 23 the biannual review of the Nuclear Safety Research 24 Program.
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117 1 report writing --
2 VICE CHAIRMAN REMPE: Matt, say again what 3 you said. We're not going to have any more 4 transcribers, right, for the rest of this session or 5 this meeting? Because we're not going to need a 6 transcriber for that or for P&P. P&P's public, but --
7 CHAIRMAN SUNSERI: Well, I don't know 8 about transcribers, I'm just talking about open 9 session.
10 MEMBER MARCH-LEUBA: We'll stay have a 11 transcriber. You need to put your microphone on.
12 VICE CHAIRMAN REMPE: P&P is open.
13 CHAIRMAN SUNSERI: Okay, we are going 14 closed.
15 (Whereupon, the above-entitled matter went 16 off the record at 2:07 p.m.)
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LO-0220-69052 February 28, 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 Titled ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, 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 Full Committee Meeting on March 5, 2020. The materials support NuScales presentation of the NuScale steam generator design.
The enclosure to this letter is the nonproprietary presentation titled ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Marty Bryan at 541-452-7172 or at mbryan@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 Gregory Cranston, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 Bruce Bavol, NRC, OWFN-8H12
Enclosure:
ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, 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-0220-69052
Enclosure:
ACRS Full Committee Presentation: NuScale - Steam Generator Design, PM-0220-69051, 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 Full Committee Presentation NuScale Steam Generator Design March 5, 2020 1
PM-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presenters Kevin Spencer Engineer, NSSS Engineering Bob Houser Manager, Testing and Code Development Brian Wolf Supervisor, Code Development Marty Bryan Licensing Project Manager 2
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Agenda
- Steam Generator Design
- DCA Revisions 3
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Steam Generator Design 4
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Steam Generator Inlet Flow Restrictor Inlet Flow Restrictor (IFR)
IFR in Tubesheet 5
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Steam Generator Design
- Integral Helical Coil SG Design features
- Shell side is primary side - Tube side is secondary side
- Alloy 690 TT (1380 tubes, 74 - 86 ft long, 5/8 OD)
- Low flow in primary (~1ft/sec)
- Tube wall degradation allowance (0.010 > ASME min wall)
- Support 100% volumetric inspection
- Normal access to shell side of tubes from below during refueling
- Steam Generator Program and In-service Inspections
- Follow guidance of NEI 97-06 & EPRI (COL Item 5.4-1: Develop and implement a SG Program)
- SG is designed with a flow restrictor at tube inlet to reduce the potential for density wave oscillations (DWO) 6 PM-0220-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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DCA Revisions
- An Action Item has been established for the Combined License applicant (COL Item 3.9-14)
A COL applicant that references the NuScale Power Plant design certification will develop an evaluation methodology for the analysis of secondary-side instabilities in the steam generator design. This methodology will address the identification of potential density wave oscillations in the steam generator tubes, and qualification of the applicable portions of the reactor coolant system integral reactor pressure vessel and steam generator given the occurrence of density wave oscillations, including the effects of reverse fluid flows within the tubes.
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DCA Revisions (contd)
- FSAR Section 3.9 has been revised and establishes a COL Item for development of an evaluation methodology for analysis of secondary side instabilities.
- FSAR Section 5.4 clarifies language related to secondary side instabilities.
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NuScale Conclusion
- The successful completion of ITAAC and the COL Item constitutes the basis for the NRC determination to allow operation of a facility certified under 10 CFR 52 9
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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 10 PM-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Backup Material 11 PM-0220-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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ITAAC Closure Path for DWO
- Resolution of DWO is to be achieved through ITAAC activities related to the steam generator
- Tier 1 Table 2.1-2 defines the NuScale Power Module (NPM) ASME Code Class 1, 2, 3, and CS components that comply with ASME Code Section III requirements including:
Equipment Name ASME Code Section III RCS Integral RPV/SG/Pressurizer 1
- Number 02.01.01 specifies that each ASME Code Class 1, 2, and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550 12 PM-0220-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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ITAAC Closure Path for DWO (continued)
- An ITAAC inspection is performed of the NuScale Power Module ASME Code Class 1, 2, 3, and CS as-built component Design Reports to verify that the requirements of ASME Code Section III are met
- From Subsection NCA of the 2013 Edition of the ASME Code -
- NCA-2142.2 requires that Design Specifications identify all loadings (e.g.
pressure, temperature, mechanical loads, cycles, and/or transients) and the service limits a component will experience
- Transient (TH) loads are based on time history of design basis transients, described in DCA Section 3.9.1.
- NCA-3254 and 3255 provide additional information about design specifications
- NCA-3260 requires that the Design Report evaluate the loads as defined in the design specification 13 PM-0220-0220-69051 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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NRC Review of NuScale Steam Generator NuScale Design Certification Application ACRS Full Committee Meeting March 5, 2020 (Open Session)
March 5, 2020 1
Agenda
- NRC Staff Review Team
- Summary of Review of Steam Generator (SG) Materials, Design, and Inspection
- Safety Significance
- Method of Analysis
- Appendix G to 10 CFR Part 52 March 5, 2020 2
NRC Staff Review Team
- Technical Reviewers:
o Gregory Makar, materials engineering o Leslie Terry, materials engineering o Yuken Wong, mechanical engineering o Peter Yarsky, Office of Research o Raymond Skarda, Office of Research o Carl Thurston, reactor systems o Kaihwa Hsu, mechanical engineering o Steven Hambric (consultant)
- Project Management:
o Marieliz Johnson o Bruce Bavol
- Technical Management:
o Thomas Scarbrough, mechanical engineering o Rebecca Patton, reactor systems o Steven Bloom, materials engineering March 5, 2020 3
NuScale Steam Generator SER Sections 5.4.1 and 5.4.2 SG Materials, Design, and Inspection FINDING: SG Materials and SG Program meet applicable requirements for most review areas:
- Materials acceptable with respect to selection, fabrication, testing, and inspection
- Design limits crevice areas along tubes
- Primary and secondary water chemistry acceptable (based on industry guidelines)
- Design provides primary and secondary access for inspection and for removal of corrosion products and foreign objects March 5, 2020 4
SER Sections 5.4.1 and 5.4.2 SG Materials, Design, and Inspection FINDING: SG Materials and SG Program meet applicable requirements for most review areas:
(Continued)
- SG Program based on applicable industry guidelines and consistent with the Standard Technical Specifications
- Generic tube plugging criterion determined in accordance with applicable guidance
- Combined License (COL) applicant will develop and implement an SG Program and provide corresponding plant-specific information March 5, 2020 5
SER Sections 5.4.1 and 5.4.2 SG Materials, Design, and Inspection SG DESIGN - Secondary Flow Oscillations
- NRC staff considers design demonstration of structural and leakage integrity for SG tubes to be incomplete for DCA including:
- Ability of SG tubes to maintain structural and leakage integrity during density wave oscillation (DWO) in SG secondary fluid system
- Method of analysis to predict thermal-hydraulic conditions and loads of SG secondary fluid system
- NuScale is working to demonstrate SG tube integrity subsequent to design certification March 5, 2020 6
Regulatory Process for Incomplete SG Tube Integrity
- NRC staff is proposing to specify structural and leakage integrity of SG tubes as not resolved and not receiving finality in NRC draft proposed rule for NuScale design certification.
- Appendix G to 10 CFR Part 52, Section VI, Issue Resolution, is being proposed to clarify that SG tube integrity is not resolved within the meaning of §52.63(a)(5)
- Section IV, Additional Requirements and Restrictions, is being proposed to state that COL applicant is responsible for providing design information to address SG tube integrity.
- Draft proposed rule currently in concurrence process prior to being provided to the Commission for approval.
March 2020 Non-Proprietary 7
SG Secondary Fluid System Method of Analysis
- DCA Part 2, Tier 2, Section 3.9.1.2 states that it lists computer programs used by NuScale for dynamic and static analyses and hydraulic transient load analyses.
- Section 3.9.1.2 does not include the method of analysis to appropriately predict thermal-hydraulic conditions and loads of SG secondary fluid system.
- In demonstrating SG tube integrity, COL applicant will need to provide information demonstrating that 10 CFR Part 50, Appendix A, GDC 4, is met for the method of analysis to predict thermal-hydraulic conditions of SG secondary fluid system and resulting loads, stresses, and deformations from DWO.
March 5, 2020 8
Demonstration of SG Tube Integrity
- NuScale has not provided reasonable assurance that flow oscillations that occur in SG secondary fluid system will not cause damage to SG tubes directly from DWO or indirectly by inlet flow restrictors (IFRs).
- COL applicant will need to provide information demonstrating that 10 CFR Part 100 and Part 50, Appendix A, GDC 4 and 31, are met with respect to structural and leakage integrity of SG tubes that might be compromised by adverse effects from DWO in SG secondary fluid system.
March 5, 2020 9
DWO Phenomenon
- TF-2 testing involved a full scale mock-up of 252 tubes.
- DWO was observed during TF-2 testing with temperature and flow oscillations in the secondary coolant.
- DWO frequency during TF-2 testing did not excite SG tube structural resonances. Alloy 690 (Ni-Cr-Fe)
- TF-2 alternating stress intensities for instrumented TF-2 tubes were below fatigue endurance limit, although TF-2 geometry, materials, and operating conditions might not be conservative compared to as-built SG.
- As discussed on the next slides, the staff is concerned about the potential impact of DWO on the SG tubes directly and indirectly by the IFRs.
March 5, 2020 10
SG Inlet Flow Restrictor
- SG Inlet Flow Restrictor (IFR) designed to provide necessary pressure drop to limit DWO in the SG tubes.
- Staff evaluated leakage flow instability (LFI) between IFRs and SG tubes during forward flow test (separate from TF-2) and did not identify any concerns.
- However, testing did include DWO conditions.
- NuScale has not validated the final IFR design.
March 5, 2020 11
SG Inlet Flow Restrictor -
DWO Concerns
- Unstable DWO could cause reverse flow through IFRs
- Subcooled liquid for modest DWO
- Slug and two-phase flow for strong DWO
- NuScale has not yet evaluated the potential impacts on SG tubes and IFRs for reverse flow such as:
- Fatigue of bolted joints, and loose IFR parts
- LFI in that cantilevered IFRs are less stable under reverse flow
- Cyclic pressure drops
- High speed turbulent two-phase flow
- Cavitation erosion of SG tube walls
- Wear of IFRs and/or tube walls that could further worsen stability March 5, 2020 12
Post-Design Certification
o Provide validated SG secondary fluid system flow thermal-hydraulic method of analysis o Demonstrate that SG tubes will not be damaged by DWO directly or indirectly by IFRs
o Complete ITAAC on Tier 1 Table 2.1-4 (#1) to confirm that ASME BPV Code Class components designed to ASME BPV Code Section III o Implement Comprehensive Vibration Assessment Program (CVAP) -
COL Item 3.9-1 Satisfy Tier 1, TF-3 flow testing requirement, and Tier 2, Table 14.2-72 SG flow-induced vibration testing Instrument one tube in initial startup SG testing with strain gages at top, middle, and bottom, for FIV evaluation March 5, 2020 13
Next Steps
- NuScale is preparing errata for Revision 4 to DCA to clarify SG secondary fluid flow issues that could impact SG tubes and IFRs.
- NRC staff discusses SG tube integrity, including SG secondary flow method of analysis, in the draft proposed rule for NuScale design certification to be provided for Commission approval.
o Draft proposed rule excludes SG tube integrity from finality.
o NRC staff will address SG tube integrity as part of a NuScale COL application review.
- Other aspects of the NuScale SG design are acceptable to the NRC staff and would be granted finality.
March 5, 2020 14
Questions?
March 5, 2020 15
LO-0320-69151 March 4, 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 Full Committee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-0320-69146, 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 Full Committee Meeting on March 5, 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 Full Committee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-0320-69146, 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 Full Committee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-0320-69146, 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-0320-69151
Enclosure:
ACRS Full Committee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-0320-69146, 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 Full Committee Presentation NuScale Topical Report Rod Ejection Accident Methodology March 5, 2020 1
PM-0320-69146 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
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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
- LOCA EM assessment basis Extended leveraged for non-LOCA. DHRS cooling
- 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
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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
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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 7
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Unique Event Method (Flowchart) 8 PM-0320-69146 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-0320-69146 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-0320-69146 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-0320-69146 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-0320-69146 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-0320-69146 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-0320-69146 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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- CHF - Critical Heat Flux
- GDC - General Design Criteria
- HZP - Hot Zero 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-0320-69146 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 17 PM-0320-69146 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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LO-0320-69139 March 4, 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 Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, 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 Full Committee Meeting on March 5, 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 Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, 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 Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, 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-0320-69139
Enclosure:
ACRS Full Committee Presentation: NuScale Topical Report, Loss-of-Coolant Accident Evaluation Model, PM-0320-69138, 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 Full Committee Presentation NuScale Topical Report Loss-of-Coolant Accident Evaluation Model March 5, 2020 1
PM-0320-69138 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 Ben Bristol Supervisor System Thermal Hydraulics 2
PM-0320-69138 Copyright 2020 by NuScale Power, LLC.
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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
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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-0320-69138 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-0320-69138 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
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Element 1 PIRT 8
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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-0320-69138 Copyright 2020 by NuScale Power, LLC.
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Element 2 Assessment Base 11 PM-0320-69138 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-0320-69138 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-0320-69138 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-0320-69138 Copyright 2020 by NuScale Power, LLC.
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Element 3 NRELAP5 NPM LOCA 15 PM-0320-69138 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-0320-69138 Copyright 2020 by NuScale Power, LLC.
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Element 4 Applicability Evaluation 17 PM-0320-69138 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-0320-69138 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-0320-69138 Copyright 2020 by NuScale Power, LLC.
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Appendix B to LOCA LTR Extension to IORV Event 20 PM-0320-69138 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
- Element 3 (NRELAP5 Model) unique due to event classification 21 PM-0320-69138 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-0320-69138 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-0320-69138 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-0320-69138 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-0320-69138 Copyright 2020 by NuScale Power, LLC.
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LO-0320-69143 March 4, 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 Full Committee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0320-69141, 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 Full Committee Meeting on March 5, 2020. The materials support NuScales presentation of the Non-Loss-of-Coolant Accident topical report.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Full Committee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0320-69141, 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 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 Full Committee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0320-69141, 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-0320-69143
Enclosure:
ACRS Full Committee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0320-69141, 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 Full Committee Presentation NuScale Topical Report Non-Loss-of-Coolant Accident March 5, 2020 1
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Presenters Ben Bristol Supervisor, System Thermal Hydraulics Meghan McCloskey Thermal Hydraulic 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
- 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
- Short-term response divided into 3 generic phases with associated FoM 10 PM-0320-69141 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-0320-69141 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-0320-69141 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-0320-69141 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-0320-69141 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Presentation to the ACRS Full Committee Staff Review of NuScale Topical Report TR-0716-50350, Revision 1, Rod Ejection Accident Methodology TR-0516-49422, Loss-of-Coolant Accident Analysis Methodology TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology Presenters:
Chris Van Wert - Senior Reactor Systems Engineer, Office Nuclear Reactor Regulation Shanlai Lu - Senior Nuclear Engineer, Office Nuclear Reactor Regulation Alex Siwy - Reactor Systems Engineer, Office Nuclear Reactor Regulation March 5, 2020 (Open Session)
Non-Proprietary 1
Presentation to the ACRS Full Committee 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 Nuclear Reactor Regulation March 5, 2020 (Open Session)
Non-Proprietary 2
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 3 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 briefed advisory committee on reactor safeguards (ACRS) subcommittee on February 19, 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.
4 Non-Proprietary
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 Reactivity Initiated Accident (RIA) analysis for the NuScale Design Certification Application (DCA)
- Contained in Section 15.4.8 of the Safety Evaluation Report (SER) for the NuScale Design Certification
- During its review, staff audited calculations and other supporting information 5 Non-Proprietary
Analysis Criteria
- The staff reviewed the proposed analysis criteria
- Reactor Coolant System Pressure
- Fuel Cladding Failure
- Core Coolability
- Fission Product
- The staff concluded that the proposed criteria either followed or were conservative to the guidance provided in Standard Review Plan (SRP)
Section 4.2 Appendix B
- Staff also notes that DG-1327, Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents is currently being developed.
- Draft guidance is not staff requirements, but the staff notes that the more stringent internal limits imposed by NuScale would not exceed the draft guidance limits as they currently stand 6 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 7 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 reactor coolant system (RCS) pressure evaluation 8 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 9 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.
10 Non-Proprietary
Questions?
11 Non-Proprietary
Presentation to the ACRS Full Committee Staff Review of NuScale Topical Report TR-0516-49422 Loss-of-Coolant Accident Analysis Methodology Presenters:
Dr. Shanlai Lu - Senior Nuclear Engineer, Office of Nuclear Reactor Regulation March 5, 2020 (Open Session)
Non-Proprietary 12
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 13 Non-Proprietary
Design Features And Scope 3 RVVs 2 RRVs each its own IAB, trip valve and trip reset valve Containment functions as part of ECCS
- A methodology to analyze LOCA
- A methodology to analyze IORV
- Support Peak Containment Pressure, Non-LOCA TR and Long Term Cooling Analysis Applicable Regulation:
10CFR50.46 Appendix K 14 Non-Proprietary
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 15 Non-Proprietary
Review Areas Phenomena Identification and Ranking Table Following CSAU method, NuScale identified twenty-one phenomena as important to capture in the LOCA model NRELAP5 code is used to model NPM Steam Generator Model, Containment Wall Condensation Model, Critical Flow Model, CHF Correlations. NPM Model and Nodalization NIST Tests, Scaling and Distortion Analysis A new scaling analysis approach was used with distortion analysis to justify the applicability of NIST IETs IORV Analysis Methodology Two different sets of CHF correlations are used for low flow and high flow conditions. STERN and KATHY facilities provide specific fuel CHF databases 16 Non-Proprietary
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 17 Non-Proprietary
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 18 Non-Proprietary
Questions?
19 Non-Proprietary
Presentation to the ACRS Full Committee Staff Review of NuScale Topical Report TR-0516-49416 Non-Loss-of-Coolant Accident Analysis Methodology Presenters:
March 5, 2020 (Open Session)
Non-Proprietary 20
NRC Staff Review Team
- NRC Technical Reviewers:
- Consultants (Energy Research, Antonio Barrett, NRR Inc.):
Jeff Schmidt, NRR Mohsen Khatib-Rahbar Alex Siwy, NRR Walter Tauche (subcontractor)
Ray Skarda, RES Morgan Libby Peter Lien, RES Michael Zavisca Ron Harrington, RES Jason Thompson, RES 21 Non-Proprietary 21
Review Process Overview Staff conducted its review in accordance with applicable NRC regulations and guidance Safety evaluation report (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 requests for additional information (RAIs) 33 RAI questions issued
- All resolved and responses incorporated into TR-0516-49416, Revision 2, as appropriate 22 Non-Proprietary 22
Non-LOCA Methodology Scope
- Provides a methodology for performing system transient analysis of specified non-LOCA design-basis events for the NuScale Power Module (NPM)
- Evaluates primary and secondary pressure figures of merit
- Includes interfaces with other methodologies, both upstream and downstream
- 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
- The staff is evaluating some items discussed in the TR as part of a design-specific application of the methodology 23 Non-Proprietary 23
Key Design Features and Models for Non-LOCA Staff focused its review on several key features of the NuScale design and their representation in the NRELAP5 model:
- Natural circulation design
- Helical coil steam generators (SGs)
- Transfer heat from reactor coolant system (RCS) to feedwater
- Passive decay heat removal system (DHRS) condensers
- Transfer decay heat to reactor pool using the SGs
- Evacuated containment vessel 24 Non-Proprietary 24
Applicability of NRELAP5 to Non-LOCA Analysis
- The applicant developed the non-LOCA evaluation model (EM) from the LOCA EM using graded approach described in RG 1.203
- The staff reviewed the applicants non-LOCA phenomena identification and ranking table (PIRT) to ensure that important phenomena were identified and captured in the non-LOCA TR
- The staff reviewed how the applicant addressed highly ranked non-LOCA phenomena:
- Separate effects tests: NIST HP-03, HP-04, KAIST, and SIET
- Integral effects tests: NIST NLT-02a, NLT-02b, NLT-15p2
- Code-to-code benchmark against RETRAN-3D
- Use of bounding input values
- Other analysis methodologies (e.g., subchannel) 25 Non-Proprietary 25
Significant Review 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 9351, Question 15.00.02-31)
- Staffs major concerns were the potential for reduced RCS flow rates and degradation in DHRS performance
- The applicants RAI response resolved the issue, as supported by the staff audit of underlying calculation notes and audit discussions with the applicant 26 Non-Proprietary 26
NRELAP5 Assessments Against Test Data The staff finds that:
- The KAIST, NIST-1 HP-03, and NIST-1 HP-04 tests validate the NRELAP5 DHRS models
- The SIET TF-1 tests validated steam generator secondary side phenomena, but the staff had concerns about the ability of the SIET TF-2 tests to fully validate primary-to-secondary heat transfer
- The NLT-02a, NLT-02b, and NLT-15p2 integral effects tests together demonstrate applicability of NRELAP5 to evaluate non-LOCA transients
- The benchmark against RETRAN-3D provides confidence that the NRELAP5 point kinetics model produces results similar to those from an NRC-approved code 27 Non-Proprietary 27
Significant Review Issues - NRELAP5 Assessments
- The applicant removed steam generator and DHRS heat transfer biases from the methodology in response to staff questions about:
- Steam generator heat transfer uncertainty based on the SIET TF-2 tests, associated with DCA Chapter 15 Unclear Open Item 15.0.2-4 (RAI 9466, Question 15.00.02-6)
- DHRS nodalization (RAI 9374, Question 15.00.02-22)
- The applicant provided justification that non-LOCA figures of merit are not sensitive to these biases
- Based on its review of the justification and audits of underlying calculations, the staff finds that removal of the heat transfer biases is supported for NPM model Revision 2
- The staff imposed the associated Limitation/Condition 3 28 Non-Proprietary 28
General and Event-Specific Non-LOCA Methodology
- The staff reviewed the overall non-LOCA analysis process and finds that it provides an acceptable analysis framework
- The staff finds that the deterministic approach using conservative or bounding inputs, initial conditions, and assumptions is acceptable for conservative calculations of non-LOCA events
- The staff reviewed each event-specific methodology and ensured that they will ensure conservative results when implemented
- The staff reviewed the representative non-LOCA event calculations in the TR and concludes that they illustrate how the non-LOCA methodology can be used for conservative transient analyses 29 Non-Proprietary 29
Staff SER Limitation and Condition Summary I. Future changes to LOCA TR must be assessed for impacts to Non-LOCA EM II. 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 III. 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 IV. Any credit for secondary MSIVs (not safety-related) must be approved through design review V. Event-specific electrical power assumptions, single failures, and operator actions must be approved through design review VI. 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 30 Non-Proprietary 30
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 31 Non-Proprietary 31
- DCA design certification application
- DHRS decay heat removal system
- EM evaluation model
- LOCA loss-of-coolant accident
- NPM NuScale Power Module
- PIRT phenomena identification and ranking table
- RAI request for additional information
- RIA Reactivity Initiated Accident
- SER safety evaluation report
- TR topical report 32 Non-Proprietary 32