ML19211C779

Transcript of the Advisory Committee on Reactor Safeguards Nuscale Subcommittee Meeting - June 19, 2019

ML19211C779
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Issue date:	06/19/2019
From:	Michael Snodderly Advisory Committee on Reactor Safeguards
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Snodderly, M, ACRS
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NRC-0389
Download: ML19211C779 (421)

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Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION

Title:

Advisory Committee on Reactor Safeguards NuScale Subcommittee: Open Session Docket Number: (n/a)

Location: Rockville, Maryland Date: Wednesday, June 19, 2019 Work Order No.: NRC-0389 Pages 1-257 NEAL R. GROSS AND CO., INC.

Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.

Washington, D.C. 20005 (202) 234-4433

1 1

2 3

4 DISCLAIMER 5

6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9

10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.

16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.

20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +

4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5 (ACRS) 6 + + + + +

7 NuSCALE SUBCOMMITTEE 8 OPEN SESSION 9 + + + + +

10 WEDNESDAY 11 JUNE 19, 2019 12 + + + + +

13 ROCKVILLE, MARYLAND 14 + + + + +

15 The Subcommittee met at the Nuclear 16 Regulatory Commission, Two White Flint North, Room 17 T2D10, 11545 Rockville Pike, at 8:30 a.m., Jose March-18 Leuba, Chair, presiding.

19 20 COMMITTEE MEMBERS:

21 JOSE MARCH-LEUBA, Chair 22 RONALD G. BALLINGER, Member 23 DENNIS BLEY, Member 24 CHARLES H. BROWN, JR. Member 25 MICHAEL L. CORRADINI, 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 JOY L. REMPE, Member 3 PETER RICCARDELLA, Member 4 GORDON R. SKILLMAN, Member 5 MATTHEW W. SUNSERI, Member 6

7 ACRS CONSULTANT:

8 STEPHEN SCHULTZ 9

10 DESIGNATED FEDERAL OFFICIAL:

11 MIKE SNODDERLY 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 CONTENTS 2 Opening Remarks . . . . . . . . . . . . . . . . . 4 3 Overview of Topical Report, "Evaluation June 24, 20198 4 Methodology for Stability Analysis of the 5 NuScale Power Module," and Section 15.9 6 "Stability."

7 Topical Report, "Evaluation Methodology for . . . 35 8 Stability Analysis of the NuScale Power 9 Module," Safety Evaluation and Section 10 15.9 "Stability" . . . . . . . . . . . . . . . . 59 11 Opportunity for Public Comment . . . . . . . . . 43 12 Overview of Chapter 15, "Transient Analysis" . 110 13 Adjourn . . . . . . . . . . . . . . . . . . . . 257 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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4 1 P R O C E E D I N G S 2 (8:30 a.m.)

3 MEMBER CORRADINI: Why don't we get 4 started. So this is the second day of our three-day 5 committee meeting on NuScale, I'm not going to run 6 through the pro forma discussion except to mention 7 that we have a closed line for NuScale subject matter 8 experts that should be open.

9 Could the NuScale -- back in Corvallis?

10 MEMBER BLEY: That you're on the line in 11 Corvallis.

12 MEMBER CORRADINI: That you're on the 13 line.

14 PARTICIPANT: NuScale's on the line.

15 MEMBER CORRADINI: Okay, all right. And 16 then we have the public line which is on mute mode.

17 And then I just want to remind everybody since I was 18 the offender last time, turn down your volumes. Turn 19 down your cell phones. Put them on mute, something, 20 so they don't talk back to you.

21 MEMBER BROWN: Are you going to do that?

22 MEMBER CORRADINI: And then other than 23 that I want to just proceed. We have an agenda today 24 where we are going to first go over the topical 25 report, and then transition into Chapter 15 from the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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5 1 applicant. And tomorrow we'll have the whole day to 2 talk with staff about Chapter 15.

3 So I'll turn it over to Dr. March-Leuba 4 who is going to lead us through the two days.

5 CHAIR MARCH-LEUBA: Thank you, Dr.

6 Corradini. So we're going to start with the stability 7 report, topical report, which -- but before you start, 8 let me give a summary for my colleagues of the way I 9 think this is going to go.

10 And so the stability report is an 11 excellent technically speaking. It is -- I found a 12 microphone, yes.

13 It's expanding the methodology for BWR 14 stabilities that we've known for 40 years now, and 40 15 years ago we would not be able to have done this. But 16 now we know all the problems and issues and we know 17 how to arrest them.

18 And better than that we have a particular 19 mass of people that can review your work, understand 20 it and intuitively say yes or no, whereas, it's not a 21 one person saying something, it's a community of 22 people saying.

23 So based on that methodology, we have 24 identified that the limited stability concern for 25 NuScale is natural circulation oscillation inside the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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6 1 core because of buoyancy concerns. So when you 2 generate a lighter mass on the riser you get more 3 flow, which then cools down the riser and you can have 4 oscillations. They become the normative oscillation.

5 And they analyzed it using all this methodology that 6 we know and they have decided there is no safety 7 significance to it.

8 And there could be a possibility of 9 instabilities if we get into boiling in the riser and 10 they prevent that with an exclusion region. It's one 11 that accepted a long time one of the solutions in BWRs 12 in which you don't allow boiling in the riser.

13 So, I think that it's perfectly covered 14 and I would like it to go fast so we can we move to 15 the difficult part of which are the next one on 16 Chapter 15.

17 Now one thing I'm concerned about is on 18 top of the primary system instability we have 19 secondary system instabilities. So you have these 20 long thin tubes in the steam generator where the 21 steam, the bubbles are inside of the tube and what is 22 all the experience we have is with the bubbles outside 23 of the tube.

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7 1 domain that is very large and very destabilizing. So 2 to try to stabilize that NuScale has put, as we always 3 do, an inlet restriction, and we will talk about those 4 details in the closed session. But the possibility 5 exists with non negligible likelihood. We're not 6 talking 10 to the minus 7, we're talking maybe 50/50, 7 that the steam generator will operate with flow 8 oscillations in the secondary.

9 And I'm not sure we covered the solution 10 on the open session or the closed session, but the 11 staff on NuScale have reached a solution and concluded 12 that this would not be of safety significance to the 13 core and would not violate GDC 12. And during the 14 closed session when we have our discussion, I would 15 like to reach an agreement among yourselves and that 16 is we agree with the conclusion.

17 MEMBER BLEY: What's the potential from, 18 I would imagine there would be problems with steam 19 generator supports and vibrations and that.

20 CHAIR MARCH-LEUBA: That's what I want to 21 discuss with them. So I noticed in the presentation, 22 NuScale has maybe one bullet in the whole 23 presentation, so I will ask you to speed up through 24 the next four slides, so we can get to the question 25 and answer area.

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8 1 And we may have to do it in the closed 2 session, so it's up to NuScale to tell us when to stop 3 and move to the closed session, because things get 4 proprietary real fast.

5 NuScale, please go ahead.

6 MR. PRESSON: Thank you for that opening, 7 Dr. March-Leuba. And good morning, I'm Matthew 8 Presson with licensing at NuScale Power, project 9 manager for this topical report. Presenting today 10 will be Dr. Yoursef Farwila, our stability expert, and 11 supported by myself in licensing, and Ben Bristol, the 12 supervisor of system thermal hydraulics.

13 And with that as requested, we'll get on 14 with the presentation.

15 DR. FARWILA: Thank you. Good morning, 16 everyone, Mr. Chairman and all the members. I was 17 introduced already and I already know most of you. So 18 let me speak through the presentation.

19 Our agenda is very simple today. The 20 topic is simple. Just make a quick introduction and 21 tell you about the stability solution type and 22 describe how we investigated it with theoretical, 23 numerical and experimental benchmarks and how we use 24 these tools in a procedure and methodology to 25 demonstrate the stability of the module, summarize, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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9 1 and try to save as much time for your questions.

2 Start with the main message, the NuScale 3 power module design was found to be stable in its 4 entire range of normal operation. This is 5 unconditional stability as long as we are in normal 6 operation.

7 CHAIR MARCH-LEUBA: But you are referring 8 to the primary system, correct?

9 DR. FARWILA: The primary system, yes, 10 anything that affects the core. So it is not going to 11 be unstable or growing power oscillations or flow 12 oscillations that go through the core. Outside of 13 that, other systems are essentially outside of the 14 scope or touching in the scope and we'll be having a 15 chance to discuss how much that connection is.

16 CHAIR MARCH-LEUBA: Can you remind us what 17 GDC, the general design criteria 12 says about 18 oscillations?

19 DR. FARWILA: GDC 12 says you're not 20 allowed to have oscillations unless you can detect and 21 suppress them so that the SAFDLs cannot be violated 22 with whatever thermal limits or other limits, cannot 23 be violated.

24 CHAIR MARCH-LEUBA: The way I read it is 25 you can either detect and suppress them or they cannot NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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10 1 violate SFADLs. So you may, according to GDC 12, you 2 may allow oscillations that do not violate SAFDLs.

3 Specified acceptable fields and limits.

4 DR. FARWILA: Right. That's what I 5 thought I said.

6 CHAIR MARCH-LEUBA: Mm-hmm, yeah. Okay.

7 MEMBER BLEY: I know our role here is 8 reactor safety. On the other hand, if you sell me a 9 plant with steam generators that starts coming apart 10 because of this, I'm not going to be very happy with 11 you; neither is anybody else. Although that's not 12 what we delve into, maybe sometime you can talk about 13 what the effects on steam generators could be.

14 CHAIR MARCH-LEUBA: Yeah, I think we would 15 want to save that discussion for the closed meeting, 16 so I'm going to speed you up so we can have more time 17 for that one.

18 DR. FARWILA: All right. Okay, part of 19 the main message also is to stress that outside of 20 normal operation, for any reason, the reactor can be 21 destabilized when the riser flow is voided, but even 22 then these unstable flow oscillations are limited by 23 nonlinear effect, so the magnitude of those 24 oscillations is going to be limited and there is still 25 not going be any critical heat flux violations.

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11 1 So now we identified where the stability 2 threshold is. That's already protected by scram upon 3 loss of riser inlets of cooling. We put a margin for 4 that eventuality, so, essentially, the hard stability 5 solution is to do nothing.

6 CHAIR MARCH-LEUBA: Going back to bullet 7 number two, you say that even if unstable flow 8 oscillations were to develop SAFDLs wouldn't be 9 violated. Is this because of a specific 10 characteristic of NuScale or is it because NuScale 11 operates with so much CHF margin that even if the CHF 12 oscillation never hit limits?

13 DR. FARWILA: Actually, the critical heat 14 flux ratio is going to increase. You are running a 15 lot of ahead of the presentation, but that's all 16 right. When you void the riser that gives you a much 17 greater feedback.

18 So for a unit of enthalpy added you get a 19 lot of density difference and that density contrast is 20 what drives the flow and the instability. So the flow 21 itself starts to become higher, so the average flow 22 gets higher and the magnitude of the oscillation is 23 bounded by in the upper part of the flow the voids 24 would collapse.

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12 1 increase more than that and the overshoot in the other 2 side would be about the same magnitude.

3 CHAIR MARCH-LEUBA: My question was a 4 little loaded in the sense that if in the future we 5 have a 20, 30, 50 percent power increase, power 6 upgrade that would eat into the CHF margin, you would 7 still have this effect.

8 DR. FARWILA: Yes, of course.

9 CHAIR MARCH-LEUBA: But it would have to 10 be evaluated.

11 DR. FARWILA: Definitely. Any change in 12 the way the reactor is operated is specified that we 13 will have to repeat the stability analysis. But the 14 same tools that we have with the phenomena in it would 15 be covered.

16 CHAIR MARCH-LEUBA: Thank you.

17 DR. FARWILA: So these conclusions that 18 are presented in the main message, they are based on 19 extensive first principles experimental and 20 computational. We just don't believe codes blindly.

21 We have to verify them by other ways, other methods.

22 We have to have first principles understanding.

23 MEMBER BLEY: Is this, the trip on loss of 24 subcooling in the riser, is that based, was that 25 always there or is that something that's been added NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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13 1 and might have been moved into Chapter 7 as part of 2 the instrumentation?

3 DR. FARWILA: It's always there.

4 MEMBER BLEY: Okay, thank you.

5 DR. FARWILA: So just going back, when we 6 first started with this it was a surprise. I mean why 7 do you need to have a stability analysis at all? It's 8 a PWR and PWRs, usually in the FSARs you have a 9 paragraph talking about stability and dismiss it.

10 But in our particular case you can see 11 that we have published reports with the experiments 12 that you can have instabilities in natural circulation 13 flow. And you can see here that there are several 14 configurations with horizontal cooler, horizontal 15 heater and you have vertical cooler and vertical 16 heater and vertical both.

17 And experiments could show that only in 18 the first one that you can have instabilities but none 19 of the others, but you could say the configuration of 20 the experiment is different from what we have. I mean 21 you can, not scale it properly and we don't know much 22 about it, but anyway this last configuration is very 23 similar to the NuScale instability.

24 But instead of taking that for granted, of 25 course an investigation of the modular stability is in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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14 1 order and in order to really verify that and 2 understand it. So the way we started with this, that 3 first box in there in the original if you see it in 4 the printout it was painted gray.

5 So we were in the gray area when we know 6 only that we just have a design. We have expertise.

7 We know the theory, but it's still in the gray area 8 until we go to more colorful boxes with developing a 9 code and have independent models to verify and 10 experimental data to benchmark it and understand it.

11 And once we are confident with the tools 12 we have, we go to the next step and analyze the module 13 in different modes of operation, different powers, 14 just all the range that you have and essentially do 15 perturbations to see if perturbations would grow. And 16 also I'll tell you more about that later, look at 17 stability during transients, not just from a steady 18 state.

19 And so we concluded from this that reactor 20 is stable within normal domain and most important 21 thing, identified what is the threshold for 22 instability which is riser voiding. That gets us to 23 the green area where we have a stability solution 24 which, luckily, does not introduce any additional 25 hardware as the threshold is already protected by the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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15 1 module protection system.

2 Like any new reactor with new phenomena we 3 start with a PIRT. That PIRT committee met for a week 4 and was extra conservative, everything counts, 5 everything, a lot of times were ranked high.

6 MEMBER CORRADINI: When did this occur?

7 DR. FARWILA: That occurred in 2015 or so, 8 so well, well before anything started.

9 MEMBER CORRADINI: Is that on the docket, 10 the PIRT?

11 DR. FARWILA: I don't think so.

12 MEMBER CORRADINI: Okay, right.

13 DR. FARWILA: Right. It's probably not 14 very interesting reading.

15 MEMBER CORRADINI: Because you brought it 16 up so I thought I'd ask.

17 DR. FARWILA: Yes, of course.

18 We have a PIRT, you could say a post-code 19 PIRT. There is a pre-code PIRT and a post-code PIRT.

20 The post-code PIRT is worth reading. It's part of the 21 topical report.

22 MEMBER CORRADINI: Okay.

23 DR. FARWILA: Right.

24 MEMBER BLEY: It's a topical report?

25 MEMBER CORRADINI: That's what we have as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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16 1 a separate.

2 MEMBER BLEY: That is the one we have, 3 okay.

4 DR. FARWILA: Yes. So that's you could 5 say more the relevant one, not the one we started 6 with, with always conservatively as human things.

7 So anyway, we looked at, first, the 8 principles just to look at the riser only and see what 9 happens considering the cold leg as boundary 10 condition. And we look at the stability trends with 11 power because we got surprising results that the 12 higher the power, the more stable the reactor is. We 13 did not see this before in PWRs.

14 So you could say do I believe my code?

15 That's a taboo. We cannot believe a code without 16 understanding, so the understanding comes from first 17 principles and using other models and things like 18 that.

19 So this analysis from first principles is 20 one of the important legs of this project and it 21 essentially, it informs the design of the stability 22 experiments as well because you can go to the facility 23 and you don't know what to measure and how to measure 24 it. Unless you know something from first principles, 25 there is always this organic relationship between an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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17 1 experiment and the theoretical understanding that 2 comes with it. They feed off each other.

3 So anyway, regarding the PIRT, all medium 4 ranked phenomena were treated as highly ranked. There 5 was no complacency in this regard. So important thing 6 in theoretical understanding of instability is that if 7 you have positive feedback it's not allowed, of course 8 we don't allow it, and you cannot operate system with 9 positive feedback.

10 So we have negative feedback that's 11 engineered in, but this negative feedback if it's too 12 strong and delayed it can overshoot in the correction 13 and that can be oscillatory. And so a negative 14 feedback that is delayed and sufficiently strong will 15 be, can become in principle unstable. And when you 16 look at our module, the feedback is negative.

17 I mean if you have a perturbation 18 increasing flow, it cools off the riser and that 19 reduces the density head that drives the flow 20 essentially correcting that perturbation.

21 MEMBER BLEY: On your last slide when you 22 said you treated all ranked phenomena as highly 23 ranked, what did that mean to you? Did you require 24 experiments to verify those things, or what did you 25 do?

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18 1 DR. FARWILA: Mainly put models that I 2 have high fidelity to represent these phenomena.

3 MEMBER CORRADINI: But you did do, you had 4 done -- I want to, here, get to the experiments, so.

5 DR. FARWILA: We did experiments, yes.

6 Actually, when the presentation was shortened, we are 7 not presenting as much. I can talk to them though.

8 MEMBER CORRADINI: Okay.

9 DR. FARWILA: Yeah.

10 CHAIR MARCH-LEUBA: Now it would be nice 11 if you'd give us a two-minute summary of the 12 experiments.

13 DR. FARWILA: Of course, I will.

14 Okay, so we said that this feedback is 15 delayed because it takes time to fill the riser with 16 that different temperature in order to make an effect.

17 And the strength of the feedback is of course is 18 related to the thermal expansion and that's what 19 creates the density difference.

20 So if you have a different fluid that 21 expands more or if you have boiling that will be 22 stronger or if you take the whole thing to a different 23 planet where the gravitational constant is different, 24 it would behave differently. I didn't hear a laugh, 25 but we actually had models where I changed the g in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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19 1 order to find out -- I needed to change it by a factor 2 of 10 in order to make things unstable.

3 So all right, getting to the main tool, 4 it's a code called PIM. And you can see the 5 geometrical simplification from this iconic picture to 6 more of an advanced schematic, more accurate schematic 7 to just the essential loop of really what happened.

8 So it's not looping like this, of course, you know, 9 it's looping like that.

10 But here are the essential element, a 11 small core almost like a point in the bottom of a tall 12 riser, and you have the cold leg filled with a helical 13 steam generator acting as the heat sink.

14 CHAIR MARCH-LEUBA: I don't remember the 15 details. Can you go back? Does PIM consider heat 16 conduction from the riser to the heat exchanger 17 directly through the wall?

18 DR. FARWILA: Yes, it does.

19 CHAIR MARCH-LEUBA: Okay, so it's capable 20 of doing that when the water level drops below the top 21 of the riser, it will still transfer heat?

22 DR. FARWILA: When the water level drops 23 to --

24 CHAIR MARCH-LEUBA: Below the riser.

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20 1 is no longer applicable because you don't have 2 established continuity for natural circulation.

3 CHAIR MARCH-LEUBA: All right, but I think 4 that's -- some of the calculations we've seen you 5 still establish heat transfer from the pink to the 6 blue, even though you don't have natural flow?

7 DR. FARWILA: That must be in NRELAP5.

8 CHAIR MARCH-LEUBA: Okay, thank you.

9 And while we have this here, could you 10 describe how representative with experiments, the 11 experiments that you were talking about and just tell 12 us how representative they were with respect to this 13 reality?

14 DR. FARWILA: The experiments were done in 15 the NIST-1 facility --

16 CHAIR MARCH-LEUBA: Mm-hmm.

17 DR. FARWILA: -- which is a scaled 18 hydraulic loop that looks very much like this except 19 for the height is -- was it one-third?

20 PARTICIPANT: Mm-hmm.

21 DR. FARWILA: And the diameter is much 22 smaller than one-third, so it's more of a long vessel 23 with a riser and --

24 MEMBER CORRADINI: We'll see that in July.

25 DR. FARWILA: Oh, oh. Wonderful.

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21 1 MEMBER BLEY: Out there? I thought you 2 said they did that at NIST.

3 MEMBER CORRADINI: NIST is the name of the 4 facility at Oregon State.

5 MEMBER BLEY: Oh, that's right. I had 6 forgotten that.

7 MEMBER CORRADINI: NIST is not NIST. NIST 8 is the name of the facility.

9 MEMBER BLEY: It's another NIST, got it.

10 MEMBER CORRADINI: NuScale Integral 11 something or other.

12 MEMBER BLEY: Yeah. Yeah, we -- okay.

13 CHAIR MARCH-LEUBA: Yeah. So, basically, 14 it's representative of the NuScale, you have different 15 time constants. You have to scale on the results.

16 MEMBER CORRADINI: Just so the members 17 remember, those that went, we won't ask for hands --

18 that there were prior experiments done in Corvallis 19 with a non-scaled facility. This was then rebuilt as 20 we were visiting in '15 or '16. I can't remember 21 which summer. They were rebuilding the facility to 22 have it based on some scaling discussion which we're 23 going to actually talk about later today in terms of 24 the scale.

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22 1 the scaling in maybe in the closed session, the 2 scaling of the experiments and --

3 CHAIR MARCH-LEUBA: But in summary, the 4 results of the NIST instability experiments did it 5 show it was stable or unstable?

6 DR. FARWILA: It showed it too stable, 7 almost dead stable.

8 CHAIR MARCH-LEUBA: Okay.

9 DR. FARWILA: Okay.

10 CHAIR MARCH-LEUBA: He's disappointed.

11 This is what I will ask you to, is skip 12 through the next four slides.

13 DR. FARWILA: Yes, I'm going to skip 14 these.

15 MEMBER CORRADINI: Before you skip, so I 16 want to understand. So PIM is not a homogeneous 17 equilibrium model; it's non-equilibrium between the 18 phases?

19 DR. FARWILA: Non-equilibrium mechanical 20 and thermal.

21 MEMBER CORRADINI: So you have different 22 temperatures and different velocities.

23 DR. FARWILA: Yes.

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23 1 analysis to show the regions of stability or 2 instability?

3 DR. FARWILA: We did and we published it.

4 MEMBER CORRADINI: Was that the published 5 -- is that the reference that you had referenced a 6 couple slides ago, or is it some other reference?

7 DR. FARWILA: I did not put -- no, no.

8 This reference here is not ours.

9 MEMBER CORRADINI: Okay, so I'd be curious 10 about the reference, but you did do a linear analysis 11 to begin with?

12 DR. FARWILA: Yes, we did.

13 MEMBER CORRADINI: Okay. And then my next 14 question, you can do it when you want. I'm curious 15 about the regimes of stability and instability between 16 the linear analysis and the non-linear analysis.

17 Because in other applications for even just gas 18 instability flows in high-pressure gases, which was 19 done for gas-cooled reactor designs, you find that the 20 regions are hard to predict.

21 So I'm kind of curious how they laid on 22 top of each other, or at least if you looked at these 23 comparisons.

24 DR. FARWILA: Right. In the beginning we 25 thought maybe you have a frequency domain code.

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24 1 MEMBER CORRADINI: Right.

2 DR. FARWILA: But since we know how to 3 write time domain codes that are not hampered by the 4 problems we use to have with numerical dispersion 5 diffusion and all these things, we thought we would 6 just have a time domain code with the small 7 perturbations it's totally equivalent to a frequency 8 domain.

9 MEMBER CORRADINI: Sure.

10 DR. FARWILA: But outside of the normal 11 operation we get limit cycles and non-linear 12 circulation in everything.

13 MEMBER CORRADINI: Okay, fine. Thank you.

14 DR. FARWILA: So, yeah.

15 Okay, so I'm going to skip through the 16 model since you already have the topical report and 17 have read this presentation before. Just one little 18 thing here. We looked at what was not modeled and we 19 do not model the pressurizer. It does not contribute 20 to the momentum.

21 And there are certain things that we 22 anchor to the more detailed code, NRELAP5, so we don't 23 have to model them directly. There are simple things 24 also that we know is conservative not to model, like 25 we take a pneumatic riser because whatever heat NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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25 1 exchange is going to go into them, the transient and 2 things like that, so anything that's not modeled is 3 either conservative or very low ranking.

4 MEMBER BLEY: Conservative always bothers 5 me because it's conservative with respect to something 6 but maybe not some other things. So it's conservative 7 with respect to the generation of oscillations?

8 DR. FARWILA: Yes.

9 MEMBER BLEY: Is that what you mean by 10 conservative?

11 DR. FARWILA: Of course.

12 MEMBER BLEY: Okay. Well, of course is 13 nice.

14 DR. FARWILA: Actually, things that we say 15 conservative we've verified it to be conservative.

16 Like for this methodology we say we do not model heat 17 exchange through the riser, but it's modeled in the 18 code and you can see that there's a little bit of 19 stabilization when you have it, so we say, "Okay, 20 don't worry about it. Let's just continue," and use 21 the historical calculation when we did not have that 22 feature.

23 DR. SCHULTZ: So with all of the features 24 that you've listed in the previous slides then, in 25 order to identify that conservatism you did NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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26 1 sensitivity studies associated with each of the 2 elements that you've described?

3 DR. FARWILA: This is why there's been a 4 lot of sensitivity studies, yes.

5 DR. SCHULTZ: Okay, thank you.

6 DR. FARWILA: Yeah. We had three years.

7 DR. SCHULTZ: Okay, that helps.

8 DR. FARWILA: All right, so how the 9 analysis was done, in the linear regime you have SS, 10 meaning steady state, so we have a steady state at a 11 certain power and for each power we have flow because 12 it's natural circulation. You cannot move power and 13 flow independently without a pump. So when we say in 14 the range of power, we also mean in the range of flow.

15 So we perturb a steady state and look at 16 the result. You will see the flow and power 17 oscillating and the oscillations are decaying, and 18 that by looking at the manner in which these decay we 19 know what the decay ratio is. So that's how we get 20 the stability parameters, the decay ratio and the 21 oscillation period.

22 And we find that as you said before, we 23 have unconditional stability in the entire operation 24 range and a curiosity there that decay ratio decreases 25 when you increase power and also when you increase NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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27 1 exposure. And we find --

2 CHAIR MARCH-LEUBA: And by exposure you 3 really mean the reactivity feedback is larger at the 4 end of cycle?

5 DR. FARWILA: Yes.

6 CHAIR MARCH-LEUBA: So increasing 7 reactivity feedback is good.

8 DR. FARWILA: When you have large negative 9 feedback it's stabilizing. And that also was done 10 from first principles in the paper we published in 11 Nuclear Science and Engineering. It was also in the 12 NURETH-16. It was invited for archiving later on.

13 MEMBER CORRADINI: The Chicago meeting.

14 DR. FARWILA: I don't remember where 15 because I didn't manage to go.

16 MEMBER CORRADINI: Yeah, okay. Fine.

17 DR. FARWILA: All right. These 18 observations all are in agreement with independent 19 reduced older models so we have additional way to 20 trust it.

21 All right, the application methodology, we 22 apply these perturbations to steady state to get the 23 decay ratios. We vary power widely from five to 24 hundred percent. Actually, we do one percent also, to 25 see what could happen. But you cannot get CHF NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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28 1 response or worry about SAFDLs below 20 percent, but 2 we say 5 since we started with that.

3 We do beginning of cycle and end of cycle 4 and if we have any reason to suspect any more limiting 5 with later feedback, maybe closer to the beginning of 6 cycle but not the beginning of cycle itself, we 7 examine that as well.

8 And we have considered assumptions and not 9 only for the moderator temperature coefficient, but 10 also for decay heat fraction because decay heat 11 fraction is just like fission energy except for it has 12 a zero feedback from the moderators. So if the 13 moderator feedback is positive then it is more 14 conservative to have zero decay heat.

15 But if you have negative moderator 16 temperature coefficient then it is more conservative 17 to assume the largest possible in decay heat, like you 18 have been operating at high power for some time and 19 then you drop the power.

20 And I have a question.

21 MEMBER RICCARDELLA: Yeah. On the 22 previous slide, could you go back to that, please?

23 And in the bottom sort of bullets you have decay ratio 24 decreases with power, period decrease. Is that 25 decreases with increasing power?

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29 1 DR. FARWILA: Yes, with increasing power 2 and increasing cycle exposure.

3 MEMBER RICCARDELLA: Okay, and period also 4 decreases with increasing power.

5 DR. FARWILA: Period also decreases with 6 increase in power, yes.

7 MEMBER RICCARDELLA: Okay, thank you.

8 DR. FARWILA: You're welcome.

9 All right, so outside the normal 10 operation, the only transient we -- we did a scoping 11 of all the AOOs in a generic kind of way and the only 12 thing that became unstable was a depressurization 13 transient because only then you can void the riser.

14 And so we could get limit cycles and we could see that 15 the CHFR, the critical heat flux ratio actually 16 increases, not decreases.

17 So these stability conclusions are generic 18 but we intend to, are committed to confirm them if 19 needed, like if a plant upgrades to, or increased rate 20 of power or the plant operation changes, you have 21 anything different that may affect normal rate of 22 temperature, the activity or the fuel design itself 23 that could increase the rated flow or things of that 24 sort, so anything that would affect the nature of 25 natural circulation or feedbacks will be re-examined.

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30 1 MEMBER REMPE: So you said you did some 2 sensitivity studies with respect to power uprates. Do 3 you have any feel for how much of power uprate might 4 affect this conclusion? Like is ten percent enough, 5 or did you do a sensitivity study in that area?

6 DR. FARWILA: Okay, what, fortunately, for 7 this module, when you increase power you become more 8 stable.

9 MEMBER REMPE: Okay.

10 DR. FARWILA: So power upgrades as long as 11 it does not create voiding in the riser we are 12 covered. But they have been analyzed. I did not 13 analyze them myself, but they have been analyzed.

14 MEMBER REMPE: So you don't have other 15 individuals who've done analyses to look at how much 16 of a power uprate would cause voids in the riser?

17 DR. FARWILA: No, I don't have that 18 number.

19 MEMBER CORRADINI: It can be calculated, 20 but I'm more interested in your 50-megawatt machine 21 than other machines.

22 DR. FARWILA: All right.

23 MEMBER BLEY: You've said that several 24 times that it was kind of a surprise that as power 25 goes up, the oscillations go down.

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31 1 DR. FARWILA: Yes.

2 MEMBER BLEY: Because the code said so is 3 one thing, physically how do you explain that 4 phenomena?

5 DR. FARWILA: When the code said so, 6 everything stopped. We had to understand it from 7 first principles. And I probably should get into that 8 in the closed session.

9 MEMBER BLEY: Closed session, that's fine.

10 DR. FARWILA: All right. With the long-11 term stability solution, you already said that before.

12 It's a originally seclusion. It's one-dimensional not 13 two-dimensional like PWRs. And we just protect 14 against void in the riser and that's essentially that.

15 MEMBER CORRADINI: Okay. I think you have 16 to, for the members, explain what you mean by one-17 versus two-dimensional.

18 DR. FARWILA: Okay. For boiling water 19 reactors we have a power flow map, so essentially one 20 acts as power, the other acts as flow. And the 21 operator can take the state of the reactor, you can 22 vary power by putting control rods. You can change 23 flow by changing the pump or pump valves.

24 But in this reactor module we don't have 25 a pump, so if you change power, you change flow with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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32 1 it. So essentially have, it's like a straight line.

2 It's power versus flow is unique curve, so that makes 3 it a one-dimensional seclusion.

4 MEMBER RICCARDELLA: You said a unique 5 curve.

6 DR. FARWILA: Yes. I mean later on when 7 we say we want to examine the transients, so we 8 consider the power and flow as stage variables. So in 9 the phase space you access points other than that 10 unique line only through transients. So you have non-11 zero time derivatives that take you somewhere else and 12 maybe there are islands of instability we haven't seen 13 or anything like that, so that's really the rationale 14 for examining transients.

15 All right, so do I speed through this one 16 since we already know what the conclusions are and go 17 to questions? Because I think I'm almost out of time.

18 MEMBER CORRADINI: Well, I want you to --

19 well, I'm not sure. I don't really care about that.

20 I want to make sure I understand the experiments that 21 were done. So where is the point that we want to talk 22 about the experiments, in closed session or now?

23 DR. FARWILA: Best in the closed session.

24 MEMBER CORRADINI: Okay.

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33 1 proprietary information in the experiments.

2 MEMBER CORRADINI: Okay.

3 DR. FARWILA: Okay.

4 CHAIR MARCH-LEUBA: So anyone have 5 questions by the members?

6 I'm trying to speed up through the open 7 session so we can go into the closed session and talk 8 freely.

9 MEMBER CORRADINI: Thank you all.

10 DR. FARWILA: Thank you.

11 CHAIR MARCH-LEUBA: Okay, Bruce. And 12 rules apply, you guys are authorized to go five miles 13 over the speed limit.

14 MR. BAVOL: Okay, good morning. My name 15 is Bruce Bavol. I'm a project manager for the NRC.

16 This is staff's presentation for the open session for 17 the Stability Topical Report, Safety Evaluation along 18 with 15.9, the Stability section of the Chapter 15.

19 To my right, Dr. Ray Skarda. To his 20 right, Dr. Peter Yarsky will be presenting, Rebecca 21 Karas is the branch chief for the Reactor Systems 22 branch. I'll mention briefly that the staff had 62 23 RAIs, all responses were resolved and closed.

24 Our full committee is scheduled for July 25 10th, and the staff plans to issue the final SER in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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34 1 August, late August, and then the dash A approved 2 version of the SER and topical report in November 3 2019.

4 With that I'll turn it over to the authors 5 to go over the outline and the rest of the 6 presentation.

7 MR. SKARDA: I'm Ray and I'll just start.

8 These are the eight items. The outline comprises 9 these eight items, the regulatory criteria that were 10 used to perform the stability topical, stability 11 methodology topical report review as well as 15.9 DCD 12 review, the long-term stability solution that was 13 proposed by the applicant, the instability modes and 14 the phenomena that are considered important to the 15 NuScale design.

16 The applicant developed as you saw their 17 own computer code PIM to perform the stability 18 analysis. The stability acceptance criteria and 19 uncertainty. Worst rod stuck out analysis which we'll 20 talk about a little bit later was not part of the 21 original submittal. We'll get to that. And then 22 finally we'll summarize the staff's conclusion with 23 respect to the review of the topical report as well as 24 the DCD 15.9.

25 So the five regulatory criteria that are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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35 1 important to stability are shown here, and these are 2 also the ones that are called out in the standard 3 review plan 15.9A which is specific to NuScale. This 4 is talked about a little bit before, so briefly GDC 10 5 says don't exceed the SAFDLs.

6 GDC 12, no uncontrolled out powers, 7 uncontrolled power oscillations. GDC 13, 20 and 29, 8 compressing this would be these relate to 9 instrumentation controls, systems and functions that 10 ensure that the long-term stability solution 11 accomplishes its job of protecting against 12 instabilities. And that long-term stability solution 13 is shown on the next slide. That's the exclusion 14 region.

15 The main thing here as you heard is that 16 the exclusion region criteria is simple and 17 straightforward. And if you look at the figure that 18 the whole main point of that is that you have to 19 maintain a five-degree submargin, a five-degree 20 Fahrenheit riser subcooling margin for operation.

21 If that is not the case, the modular 22 protection system precludes against instabilities and 23 forces a trip, generally either a hot leg temperature 24 trip or pressurizer pressure, low pressurizer pressure 25 trip.

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36 1 MEMBER CORRADINI: So let me ask, I don't 2 remember. But where does NuScale operate normally in 3 terms of riser subcooling compared to the forbidden 4 region and the allowed region?

5 MR. SKARDA: Yeah. So it has to be --

6 their subcooling margin is the riser and it needs to 7 be five degrees Fahrenheit of larger.

8 MEMBER CORRADINI: But what is it in 9 normal operation. That's what I don't --

10 CHAIR MARCH-LEUBA: I seem to remember it 11 was 35 Fahrenheit.

12 MEMBER CORRADINI: That's what I thought.

13 I thought it was in the 10s that you used. Okay fine.

14 MR. SKARDA: Oh, what they're operating 15 at. Okay, I'm sorry.

16 MEMBER CORRADINI: And then my second 17 question is, when I get into the forbidden region and 18 I lose my safety margin, I go to the possible 19 instabilities, is it the riser that is avoiding or the 20 possible instabilities are essentially a SAFDL issue?

21 What I'm trying to understand is am I 22 worried about the void into the core or I'm reading 23 about SAFDLs being violated because I have an unstable 24 flow?

25 MR. SKARDA: The subcooling margin NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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37 1 protects the SAFDLs through minimum, the NCF --

2 MEMBER CORRADINI: Okay, fine. That's 3 what I thought. Okay, thank you. Thank you very 4 much.

5 MR. SKARDA: Next slide.

6 So okay, so there's in terms of excluding 7 instabilities the applicant looked at several types of 8 instability modes, identified those that were relevant 9 to the NuScale design. In the DCD, I think they, as 10 you saw earlier, the one that's more important to them 11 is this natural circulation with density wave 12 oscillation characteristics riser dominant and that 13 will be discussed in more detail in the closed 14 session.

15 The applicant's findings are consistent 16 with the staff's findings based on an independent PIRT 17 that was also performed by the staff. Next slide.

18 So as I mentioned, the applicant developed 19 their own computer model and computer code to perform 20 the stability analysis, PIM. PIM's evaluation model 21 includes simple models for a thermal-hydraulics 22 reactor, kinetics, fuel thermal mechanical response 23 and steam generator to peak conduction and heat 24 transfer, which you saw earlier.

25 And it's been validated against the NIST-1 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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38 1 stability tests which were integral tests. The decay 2 ratio is the principle acceptance criteria with 3 respect to steady state operation at various power 4 levels. The main conclusion's here is that the K 5 ratio is insensitive to variations in most of the 6 important phenomena over, really, the range of 7 analysis related to the NuScale design.

8 The decay ratio acceptance criteria 9 affords adequate margin to account for biases and 10 uncertainty and that includes the numerical effects.

11 Primarily, dissipation is what I would call the main 12 one, dispersion. It's highly dissipative at least 13 with steady operation, so any change in phase is not, 14 it's just not going to be an issue from what we're 15 seeing. Next slide.

16 So I mentioned this worst rod stuck out 17 analysis, and this is something that is in combination 18 with an analysis that was done as part of 1506, the 19 return to power. There was some combination with that 20 instability analysis.

21 From the stability standpoint, it provided 22 an analysis at intermediate pressures from operating 23 pressure down to where ECC would actuate. You'll see 24 both in the SER for 1506 and so forth, just to provide 25 context that event is actually decay, DHRS overcooling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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39 1 event.

2 So you're really driving, you're trying to 3 drive this thing into sort of a recriticality mode.

4 So trying to make sure I say the right things here in 5 terms of the -- we talked about the strong moderator 6 feedback with in terms of recriticality which is 7 important in driving that power or that recriticality.

8 But the other side of that is it really 9 damps oscillations then. So in terms of an event, 10 you'd ideally like something that's going to drive the 11 initial event hard to that point and then you would 12 like to have weak moderator feedback in order to have 13 those oscillations persist longer.

14 And so that will be again discussed --

15 CHAIR MARCH-LEUBA: And we see those 16 details in Chapter 15 this afternoon and tomorrow.

17 MR. SKARDA: Yes. That's correct. That's 18 correct.

19 CHAIR MARCH-LEUBA: But just because we're 20 seeing -- because it's an overcooling event because 21 you cannot reach criticality unless you are, I believe 22 it's under 200, I mean very cold, then your coolant 23 has shrunk and you are below the riser and therefore 24 you cannot have natural circulation anymore.

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40 1 talk about this is the closed session for 15.

2 MR. SKARDA: Yeah. In fact, the next --

3 so basically before that occurs, before he sees 4 actuation, for example, there's actually, actually 5 plenty of subcooling margin in this particular case.

6 Next slide.

7 And so as you were saying, I don't have to 8 say it. Yeah, once he sees this actuates you've 9 broken the natural circulation flow pattern and you're 10 really in a pool boiling kind of mode, but at very, 11 very low power.

12 So the bottom line is this analysis by the 13 applicant demonstrated that flow oscillations are not 14 safety significant. Even if they had large 15 amplitudes, it's really low power. Next slide.

16 So the stability report conclusions with 17 respect to PIM approval for performing stability 18 analysis, PIM's a simple model but it's anchored in 19 upstream, high fidelity models to improve accuracy.

20 The decay ratio is highly insignificant to variations 21 in important phenomena.

22 The PIM predictions, both steady state and 23 transient, have been confirmed by staff through 24 independent confirmatory calculations. And so PIM is 25 -- staff's found PIM to be acceptable for performing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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41 1 the safety related stability analysis for the NuScale 2 module. Next slide.

3 In terms of the long-term stability 4 solution, the primary instability mechanisms were 5 properly identified by the applicant and confirmed by 6 independent staff TRACE analysis.

7 During power operation of the NuScale 8 power module, NuScale power module is very stable.

9 The exclusion region-based, long-term stability 10 solution is effective in preventing the reactor from 11 becoming unstable during normal operation and 12 including effects of AOOs. And then as we mentioned, 13 the potential instability during return to power with 14 respect to worst rod out is not a safety concern and 15 the five GDCs related to safety are met.

16 So briefly summarizing the review of 17 section 15.9 of the DCD, analysis of perturbed steady 18 state conditions demonstrate that decay ratio remains 19 well below the acceptance criteria for power levels 20 greater than five percent of rate of power.

21 The certain transient analyses result in 22 new studies, stable state conditions, and in others 23 that would exceed the hot leg trip or low pressurizer 24 pressure trip, enforce the exclusion region.

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42 1 AOO classes that you find in other sections of 15.9, 2 increase in heat removal, decrease in heat removal 3 from a secondary system and so forth. The first two, 4 really, the bounding events that were performed there 5 were feedwater flow increases and decreases.

6 I won't go through -- some of these were 7 dispositioned, just questions, we can take those. But 8 the bottom line is the long-term stability solution is 9 effective in preventing the occurrence of instability 10 and again the five GDCs important to stability are 11 met. And I believe that's it.

12 We have some backup slides.

13 CHAIR MARCH-LEUBA: Probably the answer to 14 my question is let's wait for the closed session, but 15 anything you can say in open session with respect to 16 the secondary side, the steam tube instabilities?

17 DR. YARSKY: Let's defer that to the 18 closed session. But the staff has a slide 19 specifically on that topic.

20 CHAIR MARCH-LEUBA: I know. And, but in 21 open session can we say that you're satisfied with the 22 proposed solution by NuScale or do you still have 23 reservations?

24 DR. YARSKY: I would say in terms of 25 secondary side instability the staff was able to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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43 1 determine that it does not pose any safety concerns.

2 CHAIR MARCH-LEUBA: Okay. That's what I 3 wanted to put on the record for the open session.

4 Thank you.

5 Any more questions -- okay. At this point 6 we're going to transition to a closed session, but 7 before that we're going to ask for comments from --

8 can we open the phone line?

9 Any members from the public that want to 10 make a comment on the open session?

11 We're waiting to see if the phone line is 12 open.

13 MEMBER CORRADINI: Is anyone on the public 14 line? Please at least acknowledge that you're out 15 there.

16 MEMBER RICCARDELLA: Mike is still trying 17 to verify.

18 MR. LEWIS: That's the worst thing. I've 19 never heard such dancing in my life trying to keep 20 away from the major problems and listening to details, 21 details, detail and then talking them into oblivion 22 instead of answering them strongly.

23 Further, there was a heck of a lot of 24 noise over this line. I presume everybody's shuffling 25 papers and hoping to cover me over.

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44 1 But this is exactly what I'm talking 2 about. You're going into a secret session, a closed 3 session, just to make it easier for the licensee to 4 cover all the problems that should be out there in the 5 public. I object strenuously and I plan to have some 6 more legal objections down the pipeline.

7 Thank you for listening to me. Good bye.

8 CHAIR MARCH-LEUBA: Just a moment. You 9 came on to the open line a little late so we didn't 10 hear your name. Can you state it for the record?

11 MR. LEWIS: Marvin, M-A-R-V-I-N, Lewis, L-12 E-W-I-S. I live in northeast Philadelphia. My email 13 is marvlewis@juno.com. No period between the V and 14 the L and juno is spelled like the goddess, not the 15 city in Alaska.

16 CHAIR MARCH-LEUBA: Thank you very much.

17 Any other comments from the open line?

18 MS. FIELDS: Yes. First of all, I'd like 19 to know when the open session returns. Is that after 20 lunch?

21 CHAIR MARCH-LEUBA: In principle, the open 22 session will start at 11:00 a.m. if we can make up 23 some time, for the open session for Chapter 11 and 24 we'll go one hour before lunch. Chapter 15, my 25 mistake. So in principle 11:00, but we may be late.

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45 1 MS. FIELDS: That's not the time frame on 2 the agenda. So you think you're coming back to open 3 session at 11 o'clock?

4 CHAIR MARCH-LEUBA: That's what we're 5 shooting for.

6 MR. SNODDERLY: This is Mike Snodderly.

7 I'll send you a email when we're going to start back 8 up in open session. But right now we plan to follow 9 the agenda and it'll be sometime around 11:00 a.m.

10 But I'll send you an email and I'll send it to all 11 that -- to that interested members of the public list 12 that I have.

13 MS. FIELDS: Okay, so I do have a comment.

14 I'm glad you brought up the issue of an increased 15 power and the possibility of a different type of fuel 16 because you may be aware there's a parallel process.

17 You have the NRC design certification 18 process which for the most part is based on a lot of 19 documentation, a lot of review. I've listened to 20 quite a few of the NRC's NuScale meetings and it's 21 been very, very valuable. But there's a parallel 22 process, the UAMPS process.

23 But I'd appreciate it if someone would 24 close their -- go on mute so there's not a lot of 25 background on the phone.

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46 1 CHAIR MARCH-LEUBA: Thank you, Sarah.

2 MS. FIELDS: All I hear is shuffling of 3 papers and I can hardly hear.

4 MEMBER CORRADINI: There's no shuffling 5 here. It must be on other lines.

6 MEMBER BLEY: And you're very clear coming 7 in.

8 MS. FIELDS: Oh, okay. So this parallel 9 process, they anticipate a 20 percent of power uprate, 10 but they haven't outlined how they would get to that 11 power uprate and implying that it would be Utah 12 Associated Municipal Power Systems, which is the only 13 entity that currently plans to submit a COL 14 application using the NuScale design.

15 So they're talking about this, they're not 16 talking about how exactly they would get the power 17 uprates whether it would be a design change or whether 18 it would be part of the COL application or whether it 19 would be some future license amendment request.

20 There are other things, if you go online 21 there are articles about new types of fuel and NuScale 22 anticipating using some of these new types of fuel.

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47 1 activities, when they're going to complete the 2 construction, when they're going to start operating 3 their first power module and when they're going to 4 start operating the complete array of 12 power 5 modules.

6 The Department of Energy, one of their 7 staff people at a conference in Salt Lake already 8 announced, "Hey, this is happening." So there's a lot 9 of effort out there and also calling this a carbon-10 free power process.

11 And anyone who knows anything about the 12 nuclear fuel chain knows that carbon is used from the 13 moment that the uranium industry starts going out and 14 exploring for uranium and produces uranium, and all 15 along the fuel chain it uses fossil fuels to make the 16 fuel.

17 And there's -- so calling it a carbon-free 18 power project is disingenuous and disinformation. So 19 on this you should be aware there's other parallel 20 world out there of public relations and trying to get 21 the public to accept this no matter what the cost 22 might be.

23 I live 20 miles from a very low-income 24 community in Utah where the ratepayers there still do 25 not exactly how much this is going to cost. So NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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48 1 whatever's going on in your departments it comes down 2 to a ratepayer getting a bill for a project that over 3 the long term no one has a real perception of how much 4 it's going to cost each and every ratepayer over a 5 period of 50 years. So that's some of the background 6 of community concern.

7 Thank you.

8 MR. SNODDERLY: Thank you, Ms. Fields.

9 So we're going to go into a closed session 10 now. So, Makeeka, could you please close the open 11 phone line.

12 I'd like to ask Bruce and other members of 13 the staff, Rebecca, if there's anybody here from the 14 staff that does not have a need to know, we'd like to 15 ask you to leave the room now. And the same thing --

16 MEMBER CORRADINI: Well, I think we had 17 scheduled a break to allow this to happen easier.

18 MR. SNODDERLY: Okay, if you want, I think 19 we're ready to go if you want to --

20 MEMBER CORRADINI: Can we at least take 21 five minutes to make sure? That's what I would 22 suggest. Let's just take five minutes, make sure the 23 open line is closed.

24 MR. SNODDERLY: Okay.

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49 1 we're going to lose members anyway to use --

2 MR. SNODDERLY: Yes, you are. Okay.

3 MEMBER CORRADINI: So five minutes.

4 (Whereupon, the above-entitled matter went 5 off the record at 9:28 a.m. and resumed at 11:30 a.m.)

6 MEMBER CORRADINI: Okay. Why don't we 7 come back in session. So, our plan is to start off 8 with the Applicant in terms of Chapter 15.

9 And we'll go through an hour. And then 10 break for lunch wherever we get in an hour. Jose?

11 CHAIR MARCH-LEUBA: Go for it.

12 MEMBER CORRADINI: Matthew, you're it.

13 MR. PRESSON: All right. Thank you again.

14 To reintroduce myself, I am Matthew Presson, Licensing 15 Specialist with NuScale Power, and Project Manager for 16 Chapter 15.

17 We are here today to discuss Chapter 15 of 18 the NuScale design certification application. And to 19 cover the unique aspects of how NuScale approaches 20 transient and accident analysis in our design.

21 The presentation today will be provided 22 primarily by the people joining us here at the table.

23 We have myself.

24 We have Megan McCloskey, Thermal Hydraulic 25 Analyst. We have Ben Bristol, Supervisor of System NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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50 1 Thermal Hydraulics. And Paul Infanger, Licensing 2 Specialist with Chapter 15.

3 And potentially joining us on the phone as 4 needed, will be Dr. Pravin Sawant, Supervisor of Code 5 Validation and Methods. Dr. Brian Wolf, Supervisor of 6 Code Development.

7 Dr. Selim Kuran, Thermal Hydraulic 8 Software Validation. Mark Shaver, Supervisor of 9 Radiological Engineering. And Greg Myers, Licensing 10 Specialist for Containment.

11 All right. For the scope of our 12 discussion today, there's a lot of information that is 13 summarized in Chapter 15, where we deal with 14 postulated transients and events.

15 There are also a number of topical reports 16 that define methodologies that we use in Chapter 15, 17 that we have not yet presented to the ACRS.

18 Therefore, our presentation today will be 19 focused on the material provided in the FSAR, citing 20 Chapter 15 itself, and the results of those 21 methodologies which are again, provided in that 22 chapter.

23 We will discuss the high level details of 24 these methods so that we have a basic understanding.

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51 1 here.

2 But any detailed discussion of those 3 methods will need to wait until we present the topical 4 reports for your review later this year.

5 CHAIR MARCH-LEUBA: Are we expecting those 6 in the October time frame?

7 MR. PRESSON: I believe October. Yeah.

8 So this is a big old slide. It is a copy of slide 77.

9 Later on in this presentation, those acronyms and 10 events will be covered in the presentation that 11 follows.

12 But we did want to open this entire 13 discussion on Chapter 15 with the summary of doses, 14 just to cover the health and safety of the public.

15 And how it applies to NuScale design.

16 As you can see, the dose consequences for 17 our postulated events, and even the beyond design 18 basis events listed there, remain very low in the 19 NuScale power module.

20 So for all of our discussions on 21 postulated events that follow, the NuScale power 22 position is that we have a safe design.

23 MEMBER SKILLMAN: What is the purpose for 24 the bolding on the rem TEDE for dose?

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52 1 those details later. But those are bolded to show the 2 highest doses that we had.

3 MEMBER SKILLMAN: Thank you. Okay, I 4 thought that something, yes.

5 MR. PRESSON: This is a quick review of 6 what we'll be going over today. Our design overview, 7 as well as the Chapter 15.

8 The intellectual assumptions for the 9 Chapter 15 analysis. So, some thermal hydraulics 10 analysis, methodologies. Some selected transient 11 results. Our radiological analysis.

12 Some on Chapter 6.2.1 containment response 13 analysis. That's where we follow it up with the --

14 from the loco model it is built off of. And the long 15 term cooling.

16 And with that, again, our overview of 17 safety and non-safety systems with Megan McCloskey.

18 MS. McCLOSKEY: Thank you. All right, so 19 before we dive into Chapter 15, I think for members of 20 -- particularly after the discussion yesterday and 21 this morning, are familiar with the NuScale power 22 module design and the internal reactor design with the 23 core and steam generator and pressurizer in one 24 vessel.

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53 1 this afternoon, we'll see examples of module 2 protection system actuations in response to events.

3 And for Chapter 15 design basis analysis, 4 we're primarily concerned with confirming the 5 effective operation of the safety-related systems for 6 safety related decay heat removal. We have two 7 systems, the emergency core cooling system, and the 8 heat removal system.

9 So to set the stage, on those we have just 10 a couple of slides that I think we can go through 11 quickly on the operation.

12 CHAIR MARCH-LEUBA: I know we say this all 13 the time, but people are on the phone, and they're 14 probably not hearing you.

15 MS. McCLOSKEY: Okay.

16 CHAIR MARCH-LEUBA: Can you just speak or 17 get the microphone, maybe -- people are -- away from 18 the paper.

19 MS. McCLOSKEY: Yeah. I was concerned 20 about the paper shuffling.

21 All right. For the emergency core cooling 22 system design, we have two reactor recirculation 23 valves on the side of the reactor pressure vessel.

24 And three reactor vent valves on the top head of the 25 containment.

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54 1 And these -- when demanded, these valves 2 open to establish a boiling condensing flow path to 3 transfer decay and residual heat from the reactor 4 pressure vessel by venting steam from the reactor 5 pressure vessel into containment where it condenses on 6 the inside surface of the containment wall.

7 And then is transferred through the 8 containment wall to the reactor pool ultimate heat 9 sink. The --

10 CHAIR MARCH-LEUBA: Yesterday we asked 11 this question when we were talking about the ECCS 12 valves. I had read into a figure, what is the water 13 level after you open the ECCS valve, and assuming 14 initial inventory normal now.

15 Where does the water level settle? I read 16 ten feet above active fuel.

17 MS. McCLOSKEY: Yes. That's the --

18 CHAIR MARCH-LEUBA: That's usual?

19 MS. McCLOSKEY: The nominal equilibrium 20 water level.

21 CHAIR MARCH-LEUBA: Okay. Thank you.

22 MS. McCLOSKEY: Yeah. Typically there's 23 more than 20 feet above the top of active fuel when 24 the ECCS valves open under a normal expected 25 progression.

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55 1 MEMBER CORRADINI: And how far -- how far 2 above active fuel are the RRVs?

3 MS. McCLOSKEY: I don't have that number.

4 MEMBER CORRADINI: It's around eight feet, 5 six feet.

6 MR. BRISTOL: It's six feet long.

7 MEMBER CORRADINI: Okay. All right.

8 CHAIR MARCH-LEUBA: Eight feet? You think 9 it's eight feet?

10 MEMBER CORRADINI: No. They said six 11 feet.

12 MR. BRISTOL: Six.

13 CHAIR MARCH-LEUBA: Oh. I thought it was 14 four. But you know better.

15 MEMBER BLEY: And you said that if a coil 16 breaks level at about ten feet?

17 MS. McCLOSKEY: Yes.

18 CHAIR MARCH-LEUBA: And the ECCS valves 19 will open, based on the instrument, that level 20 instrumentation on the containment. Is that correct?

21 MS. McCLOSKEY: Yes. The containment --

22 or sorry, the ECCS valves are actuated by the module 23 protection system either due to a high containment 24 level signal. Or after if there's a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> loss of 25 AC power supply.

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56 1 CHAIR MARCH-LEUBA: But that -- that's not 2 the MPL. I mean, it's --

3 MS. McCLOSKEY: Yes.

4 CHAIR MARCH-LEUBA: The timer -- you lose 5 DC power to the solenoids at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6 MS. McCLOSKEY: If DC power is available, 7 then after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the load for maintaining the ECCS 8 -- to the ECCS valve solenoids is shed.

9 CHAIR MARCH-LEUBA: Yeah.

10 MS. McCLOSKEY: So that's the timer.

11 CHAIR MARCH-LEUBA: So is it -- really the 12 activation is really on the water level.

13 MS. McCLOSKEY: Yes.

14 CHAIR MARCH-LEUBA: And we had a 15 discussion, I wanted to put it on the record again, 16 about the probability of failure of that level 17 instrumentation in the containment.

18 Which is an advanced sensor, which has not 19 been used in nuclear reactors before. And it is a 20 complex sensor. It's a rhythm-based sensor which we 21 haven't seen any details, but it's likely to have, be 22 a digital instrument.

23 Probably microprocessor based. So, it 24 would be subject to common core failures. So all four 25 level sensors may fail if it's a digital system.

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57 1 Or you have to consider a common core 2 failure from the digital point of view. And to 3 complicate things, the containment is always empty.

4 So you're not exercising that 5 instrumentation. So, it can fail and you never know 6 it's blind.

7 So, we are going to have to address that 8 thoroughly in Chapter 7 next time it comes here.

9 MS. McCLOSKEY: Right. In the scope of 10 the Chapter 15 analysis, since the module protection 11 system and the instrumentation are safety related, we 12 do assume that they function.

13 CHAIR MARCH-LEUBA: Right. But it's more 14 of a Chapter 7, Chapter 19 issue.

15 MS. McCLOSKEY: Yes.

16 CHAIR MARCH-LEUBA: But, I just wanted to 17 put the concept on the record again.

18 MS. McCLOSKEY: Yes.

19 MEMBER CORRADINI: Just so I remember, so 20 there are four different level sensors for the 21 containment water level?

22 MS. McCLOSKEY: Four -- there are four 23 channels, I think.

24 MEMBER CORRADINI: Four channels.

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58 1 safety related, and I thought they were, do we have 2 any source of safety-related electric power that 3 supply them?

4 MS. McCLOSKEY: No. The --

5 MEMBER BLEY: You don't. Okay.

6 MS. McCLOSKEY: Because if the -- if the 7 DC power that's supplied from the highly reliable 8 supply is lost, then our -- it is postulated to be 9 lost.

10 The safety systems actuate in the NuScale 11 design. So, the safety --

12 MEMBER BLEY: Well, a level sensor won't 13 actuate. It --

14 MS. McCLOSKEY: No. But the -- the power 15 to the ECCS valve solenoids will wake up.

16 MEMBER BLEY: They will. Yeah. That's 17 right.

18 MS. McCLOSKEY: And so then eventually the 19 valves will open to establish cooling, so.

20 MEMBER BROWN: And just to make sure I 21 understand. The radar-based system that we -- that I 22 think Jose is talking about, is actually during normal 23 operation, measuring the level up in the pressurizer.

24 Isn't that correct?

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59 1 things.

2 MS. McCLOSKEY: Yeah. That's -- I think 3 you're ---

4 MEMBER CORRADINI: Two different things.

5 Inside vessel, outside vessel. You're talking --

6 MEMBER BROWN: I'm inside the reactor 7 vessel.

8 MEMBER CORRADINI: They're -- but that's 9 not what we're asking about.

10 MEMBER BROWN: But it's outside.

11 CHAIR MARCH-LEUBA: Yeah. I'm talking 12 containment.

13 MEMBER BROWN: You're talking about -- oh, 14 you're talking about the containment water.

15 CHAIR MARCH-LEUBA: So we have four 16 sensors in the pressurizer, four sensors in 17 containment.

18 MEMBER BROWN: Okay. I didn't know you 19 were talking -- I just kind of had a disconnect.

20 MEMBER CORRADINI: I'm still -- I'm still 21 awake. That's all it should be.

22 MS. McCLOSKEY: So actually it's on 23 containment level.

24 MEMBER BALLINGER: I mean, we're also 25 making an assumption here, you are anyway. That when NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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60 1 the containment doesn't have any water in it, those 2 sensors are not working.

3 MEMBER BLEY: That's true. We don't know 4 the answer to that.

5 MEMBER BALLINGER: I mean, most 6 electronics have diagnostics that run all the time and 7 things like that.

8 So, we have to be careful when we talk --

9 when we get to Chapter 9 that we don't make the --

10 CHAIR MARCH-LEUBA: Chapter 7.

11 MEMBER BALLINGER: Chapter 7 rather.

12 MEMBER BROWN: Well, I mean, there -- the 13 reason I'm asking, the con -- the radar base one is 14 measuring the pressurizer level are above the reactor 15 vessel.

16 And they're, at least based on the 17 picture, I thought I remembered, they're going down to 18 measure the pressurizer level up in that upper part of 19 the reactor vessel.

20 For the containment ones, are they up on 21 that upper head area outside the pool? Are they --

22 are they located somewhere else?

23 I mean, it's a --

24 MS. McCLOSKEY: I'm not sure of the 25 sites.

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61 1 MEMBER BROWN: Well, you said that the 2 water level in the reactor within the containment is 3 about ten feet above the reactor core.

4 MEMBER BLEY: After everything settles 5 down.

6 MEMBER BROWN: After everything settles.

7 Are you still depending on those? I mean, that's a 8 pretty long shot for the radar detectors to be 9 measuring the water level in the containment.

10 MEMBER BLEY: They're supposed to be 11 measuring it all the way down at the bottom.

12 MEMBER BROWN: That's a -- what I'm 13 saying. That's a long shot.

14 MR. BRISTOL: This is Ben Bristol. So, 15 yeah. A couple of points. I think as the ACRS is --

16 we've described before, there's performance 17 requirements that are established based on the safety 18 analysis for the instrumentation.

19 We've documented what those performance 20 requirements are. And it is yet to be demonstrated by 21 the final sensor selection that, yes, those 22 performance requirements can be met.

23 In addition to the common cause failure, 24 there is consideration given to the sensor design as 25 an input to the diversity and defense in depth NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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62 1 analysis of the MPS system to ensure that common cause 2 sensor failure is either analyzed within the bounds of 3 the Chapter 15 analyses that are otherwise ensured 4 that it's a low probability.

5 CHAIR MARCH-LEUBA: That's a very good 6 answer. Thank you. We just put in a marker that 7 whenever Chapter 7 comes back, or Chapter 19, we'll 8 bring you back again.

9 MR. BRISTOL: Okay.

10 MS. McCLOSKEY: Can we go onto the next 11 slide? The second system for decay heat removal is 12 the Decay Heat Removal System that removes heat after 13 a loss of normal secondary site cooling.

14 When the Decay Heat Removal System is 15 actuated, the containment isolation valves, the main 16 steam and feed water lines close.

17 And the actuation valves on -- for the 18 Decay Heath Removal System open to establish an 19 alternate heat removal path through the steam 20 generators so that energy is transferred from the 21 primary side to fluid inside the steam generator 22 tubes.

23 That steam is transported through the 24 Decay Heat Removal System actuation valves over to the 25 condensers that are located in the reactor pool. Then NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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63 1 the condensate is transferred back to the steam 2 generat -- to the steam generators.

3 MEMBER BLEY: My copy of slide nine shows 4 water in the containment. That is wrong. Right?

5 Your copy doesn't. It's all white.

6 The copy they gave us shows water in the 7 line.

8 MS. McCLOSKEY: No. There's no water in 9 containment. That's the -- that's the pool.

10 MEMBER BLEY: That's slide nine.

11 MS. McCLOSKEY: That's the pool.

12 MEMBER BLEY: Oh. You're right.

13 MS. McCLOSKEY: That's the pool.

14 MEMBER BLEY: Never mind.

15 MEMBER CORRADINI: White is air. It's 16 steam.

17 MEMBER BLEY: Never mind. Never mind.

18 MEMBER BROWN: Oh, talk about a shot.

19 MEMBER CORRADINI: White is like you.

20 MEMBER BALLINGER: But you are right in 21 one sense. There's water -- our slides show water in 22 the vessel.

23 MS. McCLOSKEY: Inside the reactor vessel.

24 MEMBER CORRADINI: It was supposed to be 25 in --

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64 1 MEMBER BROWN: In the reactor vessel, yes.

2 MEMBER BALLINGER: It's light blue here, 3 dark blue in there.

4 MS. McCLOSKEY: Yes.

5 MEMBER BROWN: It's very light blue.

6 CHAIR MARCH-LEUBA: Through all the 7 documentation there are references to the old and the 8 new actuation logic for DHRS. Will you tell us what 9 is the actuation logic for the DHRS?

10 What signals tip it? What signal tells 11 MPS to trip the DHRS valve? To open it?

12 MS. McCLOSKEY: There are a couple of 13 signals. One of them is high secondary side steam 14 pressure.

15 Loss of power also actuates the valve.

16 And --

17 MR. BRISTOL: High RCS temperature.

18 MS. McCLOSKEY: High RCS temperature.

19 MR. BRISTOL: High RCS pressure.

20 MS. McCLOSKEY: And pressure.

21 CHAIR MARCH-LEUBA: Oh, high pressure 22 inside the vessel.

23 MS. McCLOSKEY: Um-hum.

24 CHAIR MARCH-LEUBA: High temperature 25 inside the vessel, or high temperature outside in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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65 1 secondary.

2 MS. McCLOSKEY: Right.

3 CHAIR MARCH-LEUBA: Okay.

4 MS. McCLOSKEY: Indications that secondary 5 site cooling is not effective.

6 CHAIR MARCH-LEUBA: The language indicates 7 that you use to trip on more things. Then I assume 8 that you have identified that they're not necessary?

9 MR. BRISTOL: So, -- this is Ben. The 10 signals used to be combined. There was two functions, 11 and we'll get into this in a little bit more detail 12 later.

13 But there was an isolation function that 14 was required to mitigate some events for the secondary 15 side. And then there was a DHR actuation that was 16 required.

17 The analysis that we did showed that those 18 two didn't necessarily need to come at the same time.

19 And as part of the start up procedure that ended up 20 being a consideration that was fairly limiting, 21 particularly at the lower temperature conditions.

22 CHAIR MARCH-LEUBA: Um-hum.

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66 1 preserve inventory, secondary isolation would occur.

2 But not immediately cause DHR actuation.

3 And not continue to require DHR to be active for the 4 MPS logic.

5 So what it does, is it allows for DHR to 6 actuate and then clear once we have established 7 cooling based on the trip signals.

8 MEMBER BROWN: So you could not actuate 9 DHS, the Decay Heat Removal System totally. You just 10 isolate the main steam isolation valves?

11 MR. BRISTOL: That's correct.

12 MEMBER BROWN: And the feed water valves?

13 MR. BRISTOL: Feed water isolation valves.

14 MEMBER BROWN: And the feed water 15 isolation valves.

16 MR. BRISTOL: Yes.

17 CHAIR MARCH-LEUBA: And only if the 18 secondary steam starts heating up, then you need 19 additional cooling.

20 MR. BRISTOL: That's correct.

21 MEMBER BROWN: So the other path is not 22 established until you met some other preconditions?

23 MR. BRISTOL: That's correct.

24 CHAIR MARCH-LEUBA: That's in the area.

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67 1 you.

2 CHAIR MARCH-LEUBA: And I assume all the 3 analysis in Chapter 15 represent the new logic, not 4 the old logic.

5 MR. BRISTOL: Well, it's a -- that's a 6 great question. It's one of the design changes that's 7 being implemented in the current analysis effort.

8 I think the NRC has asked us to describe 9 that.

10 CHAIR MARCH-LEUBA: Are you redoing 11 everything?

12 MR. BRISTOL: We are in the middle of 13 updating the FSAR analysis. They're very similar from 14 a transient progression to what we've seen before.

15 MEMBER CORRADINI: But this is -- this is 16 what is identified in many cases as Rev 3.

17 MR. BRISTOL: That's correct.

18 MEMBER BROWN: That will also be reflected 19 in Chapter 7 Rev 3 relative to this differentiation of 20 what logic is required?

21 MR. BRISTOL: That's correct.

22 MEMBER BROWN: Okay.

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68 1 you analyze, you're always going to get the same 2 answer, really good.

3 And so, but it is nice to do the paperwork 4 properly. Okay.

5 MEMBER CORRADINI: And you'll continue to 6 cheerlead that.

7 MS. McCLOSKEY: All right. So, in terms 8 of the scope of Chapter 15, the design basis 9 initiating events consider internal events that could 10 affect a module operating at power.

11 The Chapter 15 analyses that we do are 12 performed for a single module response. In terms of 13 shared systems for the NuScale design, the reactor 14 pool ultimate heat sink is the important shared system 15 there.

16 And so our long term cooling analyses 17 consider effects of the reactor pool temperature and 18 the level on the module response that account for up 19 to 12 modules rejecting heat to the reactor pool.

20 We went through a systematic process to 21 assure that we had identified an appropriate scope of 22 design basis events for Chapter 15. Particularly 23 considering the systems and components that are unique 24 to the NuScale design, such as the containment 25 evacuation system for the ECCS valves.

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69 1 MEMBER DIMITRIJEVIC: Excuse me. Could I 2 ask you? Because I just want to make sure I 3 understand what you just said before.

4 MS. McCLOSKEY: Um-hum.

5 MEMBER DIMITRIJEVIC: You said that you 6 considered that changes in pool temperature are only 7 for long term cooling for all units. Right?

8 Just for long term cooling?

9 MS. McCLOSKEY: That was an area of focus 10 for long term cooling, to assure that we had bounded 11 effects of multiple -- of potentially multiple modules 12 rejecting heat to the pool.

13 MEMBER DIMITRIJEVIC: In long term 14 cooling.

15 MS. McCLOSKEY: In long term.

16 MEMBER DIMITRIJEVIC: But you did not I 17 suppose, analyze that one unit, you know, can be long 18 -- or that even by the one unit just goes to transit.

19 MS. McCLOSKEY: We have considered -- the 20 long term cooling analyses also consider potentially 21 only one unit rejecting heat to a cold pool.

22 So that is -- that is covered.

23 MEMBER DIMITRIJEVIC: All right. Well, my 24 question is, can we have 11 units injecting pool --

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70 1 ECCS operations?

2 So, you know, when -- I was wondering how 3 did you model that pool temperature? And what case 4 would you consider for it?

5 As I understood you only considered 6 difference from the beginning only for long term 7 cooling operations.

8 MS. McCLOSKEY: Well, and let me clarify 9 here. We -- okay, in terms of multi-module effects, 10 if one module were to experience a transient or an 11 accident scenario, --

12 MEMBER DIMITRIJEVIC: Right.

13 MS. McCLOSKEY: That potentially affected 14 the reactor pool temperatures for other modules, they 15 would, I believe that they would still be under the 16 control of the applicable technical specifications for 17 those modules, to assure that to consider whether they 18 would remain operating, or what actions operators can 19 take.

20 I'm not sure if that gets at your 21 question.

22 MEMBER DIMITRIJEVIC: Well, I mean, I was 23 thinking about situations like where you have a common 24 thing like loss of offsite power. Or, you know, you 25 can have a loss of all AC power for all units.

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71 1 And they may be entering the transients in 2 the different phases. You know, because they would, 3 you know, I'm not sure, but how we end operation that 4 will go.

5 So that, you know, I mean, they will all 6 three probably. But I'm not sure how they will go to 7 different phases.

8 It's even, you know, will they go in the 9 ECCS operation all in the same time? And how will 10 actuate.

11 So, I was sort of wondering, can we have 12 some normal occurrence like loss of offsite power 13 where they will all be in the accident position of 14 certain taking, you know, depending on the pool?

15 MS. McCLOSKEY: We -- we'll get to some of 16 the example, some of the discussion in long term 17 cooling later this afternoon.

18 We considered different scenarios for long 19 term cooling, both maximum temperature cases as well 20 as minimum temperature cases. And that considers a 21 range of pool conditions.

22 And it's actually the minimum temperature 23 case where you have the most heat removal outside of 24 the containment vessel, where we have our most 25 challenge in terms of level and margin to warm NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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72 1 precipitation.

2 And so we've considered both ranges, both 3 ends of that spectrum for a range of cooling that 4 occur.

5 MEMBER DIMITRIJEVIC: Right. Well we will 6 see as you present them then. And we will get a 7 better idea of how does that fit and all 8 MR. BRISTOL: Yeah. And I'll just add 9 that the pool temperature does a specified range in 10 tech specs. So the analyses are initiated from, 11 within that range.

12 MEMBER DIMITRIJEVIC: Um-hum.

13 MR. BRISTOL: So, we don't necessarily 14 initiate a transient from something outside of that 15 range. In the condition of the loss of offsite power 16 to the, you know, to the entire plant, all the modules 17 immediately go to DHR cooling conditions.

18 MEMBER DIMITRIJEVIC: Okay.

19 MR. BRISTOL: So they don't -- they 20 wouldn't come in at stages kind of like you were 21 talking about.

22 MEMBER DIMITRIJEVIC: Right. Well, I was 23 trying to think, --

24 MR. BRISTOL: Sure.

25 MEMBER DIMITRIJEVIC: Because when we were NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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73 1 have some discussion about operator actions through 2 that. You know, because they cannot, they have enough 3 time to go from one unit to units.

4 So, I will think of a type of accident I 5 was thinking. But, even in these cases, they all come 6 to DHRS when -- I mean, will temperatures stay in the 7 -- okay. Well, we had to go to ECCS, right?

8 MEMBER BROWN: No.

9 MR. BRISTOL: Yes.

10 MEMBER DIMITRIJEVIC: You are having 11 something goes wrong with the DHRS, I'm just trying to 12 think about accidents which can cause that.

13 MR. BRISTOL: So the range of DHRS 14 performance includes analysis of a range that's 15 outside of the tech specs. The transients aren't 16 necessarily initiated from outside of that pool 17 temperature range.

18 But, there is consideration given to that.

19 And I think we'll get into that maybe a little bit 20 later, --

21 MEMBER DIMITRIJEVIC: All right. All 22 right. Well, let's do that.

23 MR. BRISTOL: As we get into the details 24 of the methodology.

25 MEMBER BROWN: During normal operation 12 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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74 1 modules, everything is working just fine. We're 2 inside a vacuum inside the containments.

3 Is it expected that there's -- the pool 4 temperature just stays constant? It doesn't heat up 5 a little bit just due to some type of heat effects 6 coming through the containment and out to the pool.

7 And then stabilizers because the pool stabilizes after 8 some point if you've got them all?

9 Is there a range of operation, 12 modules 10 where if you start cold, start them all up, and the 11 temperature would rise up just because of the reactor 12 operations themselves? Even though they're within a 13 vacuum?

14 MR. BRISTOL: Yes. Yeah, there is some 15 heat loss, known heat loss from the module to the pool 16 that's considered.

17 I think the spent fuel pool is one of the 18 larger heat loads. I'm not extremely familiar with 19 that analysis.

20 MEMBER BROWN: Yeah. Put the spent fuel 21 pool aside. It's a -- I understand that cooling.

22 MR. BRISTOL: Sure. The pool has a 23 cooling system that takes into consideration the 24 normal heat loads and off-normal heat loads of DHR 25 sometimes.

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75 1 MEMBER BROWN: But there will be some heat 2 up even though you've got the reactor vessel and 3 everything within that vacuum area. There's going to 4 be some, has to be, heat conducted through there.

5 MR. BRISTOL: Yeah. The pool definitely 6 has a heat load.

7 MEMBER BROWN: Yeah. Okay. All right.

8 MR. BRISTOL: A normal heat load that's 9 considered part of the systems then.

10 MEMBER BROWN: So, but the rest of your --

11 I'm trying to springboard off of Vesna's comment. And 12 then you've also considered if you lost power, then 13 you -- everybody goes to DHRS.

14 Then you'd have to be able to handle that 15 in terms of the general -- and the pool has no cooling 16 now. It's got the total, the total mass of water.

17 It has to be able to accomplish decay heat 18 removal for all 12 modules. And not go outside of an 19 acceptable band.

20 MR. BRISTOL: That's correct. And that's 21 where we get pretty quickly into the short term 22 transient response, pool temperature of the pool.

23 The pool is a big enough heat sink. It 24 doesn't have a huge temperature transient where the 25 long term cooling analysis is primarily where we have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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76 1 the consideration of the coping period, and what are 2 all the heat loads.

3 And ensuring the --

4 MEMBER BROWN: Yeah. I'm not worried 5 about an accident rate.

6 MR. BRISTOL: Sure.

7 MEMBER BROWN: I was interested in the 12 8 modules now are all in DHRS. Then you could have an 9 accident. And you have to consider that as well?

10 MR. BRISTOL: No. We would consider the 11 loss of power being the initiating event. And after 12 DHR actuation then we're in long term cooling 13 conditions.

14 So we wouldn't postulate a new initiating 15 event during that event progression.

16 MEMBER BROWN: Even long term?

17 CHAIR MARCH-LEUBA: He's already dumping 18 all of the heat from the core into the pool through 19 DHRS. What worse can you make this?

20 MEMBER BROWN: I don't know.

21 CHAIR MARCH-LEUBA: You're dumping 100 22 percent --

23 MEMBER DIMITRIJEVIC: This is what I was 24 trying to actually get to.

25 MEMBER BROWN: Yes. I thought that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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77 1 where you were trying to get to.

2 MEMBER DIMITRIJEVIC: I was trying to go 3 to ECCS where something goes wrong when we -- actually 4 we have additional accidents.

5 But then I, I realized that that's 6 additional accident. And you would not look in that.

7 But, I mean, it could be like some safety valve refuse 8 to close or something, I mean, you know.

9 MEMBER BROWN: No idea. You're right.

10 MEMBER CORRADINI: I guess I'd ask a 11 different question instead of asking all these things.

12 What's the rate of rise if I had to assume the decay 13 heat of all 12 modules into the pool?

14 MS. McCLOSKEY: It --

15 MEMBER CORRADINI: I calculate it to be 16 less than a degree, substantially less than a degree 17 an hour. On the order of a degree an hour.

18 That's what I would ask.

19 MS. McCLOSKEY: I don't know the specific 20 rate of rise. But if you take realistic initial 21 conditions for the pool level and temperature and 22 decay heat loads for the 12 modules, then the --

23 assuming boil off of -- assuming the heat loads to the 24 reactor pool, there's more then 30-days worth of level 25 above the top of the Decay Heat Removal Systems.

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78 1 MEMBER CORRADINI: Okay. Fine. All 2 right. That's another way of doing it. Okay. Thank 3 you.

4 MS. McCLOSKEY: Yeah.

5 MEMBER CORRADINI: I appreciate that.

6 MR. BRISTOL: It's very slow, yeah.

7 MS. McCLOSKEY: It's very slow.

8 MEMBER CORRADINI: I figure that.

9 MR. BRISTOL: Yeah.

10 MS. McCLOSKEY: Yeah.

11 CHAIR MARCH-LEUBA: Thirty days.

12 MEMBER CORRADINI: I mean, I didn't want 13 to interject -- oh, I'm sorry. I didn't want to 14 interject with their questions on the right.

15 But I have a different question. Are you 16 guys done?

17 You were at initiating events, you were 18 giving us a long list. In Table 15.01 and 15.02, you 19 identify a thing called an IE.

20 And I didn't understand why you need an IE 21 versus an AOO and a DBE, because in 15.02 your table 22 for acceptance criteria is essentially, the IE's 23 acceptance criteria is the same as the DBE.

24 So why did you make the distinction to 25 begin with? I'm clueless.

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79 1 MS. McCLOSKEY: Part of the distinction 2 comes in the radiological dose acceptance criteria.

3 MEMBER CORRADINI: Not in the thermal 4 hydraulic. So, --

5 MS. McCLOSKEY: Not in the thermal 6 hydraulic dose acceptance criteria.

7 MEMBER CORRADINI: So, for the IE, the 8 acceptance criteria is more like an AOO?

9 MS. McCLOSKEY: For -- in terms of the 10 radiological dose, it's -- the radiologic -- the 11 acceptance criteria are aligned with a small fraction 12 of the acceptable dose for accidents.

13 MEMBER CORRADINI: But not the full dose?

14 MS. McCLOSKEY: But not the full dose.

15 MEMBER CORRADINI: And it's not the AOO 16 which is essentially mainly just preserving the 17 SAFTLs.

18 MS. McCLOSKEY: Right. So when we get --

19 MEMBER CORRADINI: Where would I find 20 that? I guess I was looking, and I missed that.

21 MS. McCLOSKEY: You -- in the Table of the 22 dose analysis results for the different acceptance 23 criteria.

24 MEMBER CORRADINI: Oh.

25 MS. McCLOSKEY: For the steam generator NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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80 1 tube failure. And for small lines outside of 2 containment, you'll see the acceptance criter -- yeah.

3 Yes, that sample.

4 MEMBER BROWN: The acceptance criteria do 5 change as the --

6 MS. McCLOSKEY: The acceptance criteria.

7 MEMBER CORRADINI: That's the one where 8 it's 6.3 and not 5 or 25?

9 MS. McCLOSKEY: Correct.

10 MEMBER CORRADINI: Oh. Because the only 11 two that you identified in 15.01 was the small line, 12 and another one, which I can't remember.

13 MS. McCLOSKEY: The steam generator tube 14 failure.

15 MEMBER CORRADINI: Thank you. Thank you 16 very much. Okay. Thank you. Appreciate it.

17 MS. McCLOSKEY: So, and --

18 MEMBER CORRADINI: I'm done.

19 MS. McCLOSKEY: Okay. In terms of 20 identifying the scope of initiating events for Chapter 21 15, we started with the PRA initiating events, because 22 this was examined, then summarize the scope of events 23 that could cause a reactor trip or transient.

24 But from there we examined the systems 25 that were identified in the PRA as relevant to causing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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81 1 a reactor trip or transient in -- for additional 2 detail in order to identify the specific impacts on 3 the module, in order to categorize and classify the 4 design basis events for Chapter 15.

5 CHAIR MARCH-LEUBA: Sorry to put you on 6 the spot, but can you give me an example that you 7 identify through PRA that was not on the SRP of 8 Chapter 15?

9 MS. McCLOSKEY: An example might be, would 10 be the containment flooding and loss of containment 11 vacuum events. Because those are --

12 CHAIR MARCH-LEUBA: Okay. That's good 13 enough.

14 MS. McCLOSKEY: One potential cause of 15 loss of containment vacuum is --

16 CHAIR MARCH-LEUBA: Well, that will not be 17 on SRP 15. But it would be in the PRA.

18 MS. McCLOSKEY: Right. When we went 19 through the design basis events that we identified and 20 categorized them, our categories are consistent with 21 those for operating light water reactors.

22 Except that the category of decrease in 23 RCS flow events is not applicable due to the natural 24 circulation design of the plant.

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82 1 and -- or event progressions. PWR stability we've 2 discussed already this morning.

3 And then return to power analysis are part 4 of the Chapter 15 analysis. And we'll discuss that 5 later this afternoon.

6 MEMBER CORRADINI: Where is a good time 7 for me to ask about the evolution of NRELAP from 1.3 8 to 1.4? Since that appears in a lot of the open items 9 as an addendum.

10 You decide where in the discussion today 11 you want to explain that. And then the second one 12 that I want to get explained is the scale distortion 13 in NIST 1 and NIST 2 experiments.

14 MS. McCLOSKEY: Okay.

15 MEMBER CORRADINI: So, you don't have to 16 answer it now. You just decide where you want to 17 inject it. And then I can ask my questions. Okay?

18 MS. McCLOSKEY: Okay. I think -- all 19 right. Well, probably the closed session is 20 appropriate for any discussion of the details there.

21 MEMBER CORRADINI: Okay.

22 MS. McCLOSKEY: But, most of the 23 discussion of scale distortions between the NIST 1 24 facility and the plant, we would defer detailed 25 discussion of that to the topical report --

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83 1 MEMBER CORRADINI: Which is yet to be --

2 MS. McCLOSKEY: Methodologies. Right.

3 With the LOCA topical report methodology when we 4 present that to the ACRS committee.

5 MEMBER CORRADINI: Okay. But that's 6 nowhere in, nowhere in the short term future?

7 MS. McCLOSKEY: I believe it's this fall 8 in the October time frame was the schedule for 9 meetings.

10 MEMBER CORRADINI: At the earliest.

11 CHAIR MARCH-LEUBA: Did you say LOCA or 12 non-LOCA?

13 MEMBER CORRADINI: Both are still out 14 there as open.

15 CHAIR MARCH-LEUBA: I know. But which --

16 which were you talking about? You were talking about 17 LOCA?

18 MEMBER CORRADINI: Both.

19 MS. McCLOSKEY: LOCA.

20 MEMBER CORRADINI: Okay. And then the 21 RELAP 1.4 trans -- the transition from 1.3 to 1.4, 22 because staff has yet to finish its review of that.

23 Or I'm not sure if you've yet to even 24 issue the differences.

25 MS. McCLOSKEY: Yes. The differences are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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84 1 issued. And the staff has also audited that 2 information.

3 So, I think we can say a little bit --

4 MEMBER CORRADINI: In closed session.

5 MS. McCLOSKEY: In the closed session 6 about the details of the code changes.

7 MEMBER CORRADINI: Okay. All right.

8 That's fine. That's good. We'll just wait until 9 closed session. Thank you very much.

10 MS. McCLOSKEY: And the effects. The 11 design basis events, this has already been mentioned 12 here. We are classified as anticipated operational 13 occurrences, infrequent events or accidents.

14 Events that could potentially occur one or 15 more times during the lifetime of the plant, are 16 classified as anti -- as AOOs.

17 Events that are not expected to occur are 18 classified as infrequent events or postulated 19 accidents. Or in some cases they were conservatively 20 classified as AOOs.

21 And one example of that is the inadvertent 22 opening of an ECCS valve. We don't expect that event 23 to occur during the lifetime of the plant.

24 But it was conservatively classified as an 25 AOO. And we demonstrate that those acceptance NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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85 1 criteria are met. Because that bounds a transition of 2 other events to ECCS cooling if a loss of power were 3 to be assumed.

4 And in some cases we also simplified the 5 event classification by applying a deterministic 6 criteria where the event was similar to other PWRs.

7 Particularly where event consequences are small and 8 calculating a NuScale specific event frequency is not 9 warranted.

10 So again, failures such as in the 11 containment evacuation system are classified as an AOO 12 event.

13 MEMBER CORRADINI: So, there was an open 14 item, or an RAI, it maybe an open item. I don't 15 remember if I've got it down right, in terms of how 16 you classified return to power.

17 So, how is that being classified?

18 MS. McCLOSKEY: That is a -- it's not an 19 initiating event. It's an event progression.

20 MEMBER CORRADINI: Right.

21 MS. McCLOSKEY: And so it's not -- the 22 event classification applies to the initiating events.

23 But we demonstrate that the AOO acceptance criteria 24 are met.

25 MEMBER CORRADINI: Ah. Okay. So, the way NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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86 1 you look at it is if it were an AOO.

2 MS. McCLOSKEY: Well, that's what defines 3 the acceptance criteria. Yes.

4 MEMBER CORRADINI: Okay. Fine. Thank 5 you.

6 CHAIR MARCH-LEUBA: Because the initiating 7 event is loss of offsite power for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8 And no non-safety grade power back up coming up.

9 But yet if you -- it's perfectly 10 acceptable to make it an AOO.

11 MS. McCLOSKEY: Right.

12 MEMBER CORRADINI: I'm going to ask the 13 same question of the staff. So I'm just trying to ask 14 it here and get there and see if there's consistency.

15 MS. McCLOSKEY: Um-hum.

16 MEMBER CORRADINI: Because it was left 17 out. Okay, an AOO. Thank you very much.

18 MS. McCLOSKEY: So next slide. This slide 19 and the next several slides summarize the design basis 20 events and their respective categories.

21 So we have the event, the event 22 classification, the evaluation model used to do the 23 system thermal hydraulic analysis if applicable. And 24 the end reliant code used there.

25 Whether the event is analyzed for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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87 1 subchannel analysis with the VIPRE-01 code. And then 2 whether the event is part of the Chapter 15 3 radiological dose analysis.

4 The NuScale specific events are 5 highlighted. And so here loss of containment vacuum 6 or containment flooding is a unique event.

7 The loss of containment vacuum could be 8 postulated due to a malfunction in the containment 9 evacuation system. Containment flooding is postulated 10 due to a break in piping that carries reactor 11 component cooling water to the control rod drive 12 mechanisms on top of the reactor head.

13 So although these events, these postulated 14 events don't directly interface with the primary or 15 the secondary side of the module, they would -- they 16 could result in a slow increase in heat transfer from 17 the reactor vessel compared to the evacuated -- the 18 vacuum conditions normally in containment.

19 And so it's included here as an increase 20 in the removal event.

21 CHAIR MARCH-LEUBA: And anywhere you say 22 in RELAP 5 you mean you plan to do it with the newest 23 version of the model, correct?

24 We have results for 1.3 and you plan to do 25 it for 1.4?

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88 1 MS. McCLOSKEY: Yes. We're in the 2 progress of -- we're in the process of revising those 3 analyses to 1.4.

4 CHAIR MARCH-LEUBA: And remind me again, 5 when you use VIPRE in a transient, you use VIPRE for 6 every entire step? I don't remember the -- I remember 7 the methodology for steady state.

8 But for transient, how do you apply VIPRE 9 during the transient?

10 MS. McCLOSKEY: The boundary conditions 11 from the NRELAP 5 results are provided for power, fl 12 -- total RCS flow, pressure, and core inlet pressure.

13 CHAIR MARCH-LEUBA: At every time slice?

14 Ms. McCLOSKEY: Well, I think --

15 MR. BRISTOL: No. So there's an edit 16 frequency that's generated from RELAP and transmitted 17 to, you know, around a second or half a second.

18 CHAIR MARCH-LEUBA: Okay. But then even 19 time slice.

20 MR. BRISTOL: Yeah. That's right, yeah.

21 Yeah.

22 CHAIR MARCH-LEUBA: And how many roles do 23 you have play with VIPRE? How many roles of the, 24 subchannels does VIPRE simulate? You don't remember?

25 MR. BRISTOL: I don't.

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89 1 CHAIR MARCH-LEUBA: Okay. It's been a 2 year since we reviewed it. But, just I was asking.

3 MR. BRISTOL: That's fair. We can follow 4 up with that if that's of interest.

5 CHAIR MARCH-LEUBA: It would be nice to 6 know. I know we know, but I don't remember.

7 MEMBER CORRADINI: But you have -- you do 8 track the hot channel versus the average channel.

9 That's what I would assume.

10 MR. BRISTOL: Um-hum.

11 MEMBER CORRADINI: At the time when we did 12 this, you had a nodalization approach to as where you 13 determined the hot channel. And then you essentially 14 had less and less subchannel nodalization --

15 MR. BRISTOL: Um-hum.

16 MEMBER CORRADINI: as you went. It went 17 away from the grouping of hot channels. So I assume 18 that's what was done here.

19 MR. BRISTOL: Fifty-two rods. Is the 20 answer.

21 MEMBER CORRADINI: Okay. Thank you.

22 MS. McCLOSKEY: All right.

23 MEMBER BROWN: Before you leave that, the 24 last item, go back.

25 MS. McCLOSKEY: Um-hum.

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90 1 MEMBER BROWN: How much of an effect 2 overall is there if you just lose the vacuum and it's 3 just air? Can -- does that significantly heat up the 4 pool?

5 MS. McCLOSKEY: No. It's a --

6 MEMBER BROWN: So it's -- you could 7 operate that way? Or do you require a shut down if 8 that occurs?

9 MS. McCLOSKEY: We would be outside the --

10 MEMBER BROWN: But just air. It's air, 11 not water. It's not -- it hasn't been flooded.

12 MS. McCLOSKEY: We would be outside the 13 limits established for monitoring containment leakage.

14 MEMBER BROWN: Okay. So it would be a 15 containment leakage issue then.

16 MS. McCLOSKEY: Um-hum.

17 MEMBER CORRADINI: They'd have to shut 18 down.

19 MEMBER BROWN: That's what I guess I would 20 translate. Okay. Thank you.

21 MS. McCLOSKEY: In Section 15.2 are the, 22 you know, it results in a decrease in heat removal 23 from the secondary side.

24 And here the NuScale specific event is 25 inadvertent operation of the Decay Heat Removal NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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91 1 System. Because the Decay Heat Removal System is 2 sized for decay heat removal, so inadvertent operation 3 of -- actually causes a decrease in heat removal.

4 There are several different variations of 5 this event that we analyze. That maybe a single 6 valve, actuation valve opening.

7 It maybe an inadvertent signal that 8 actuates the DHRS train valves and closes the 9 secondary isolation valves on one train or on both 10 trains.

11 And we have some example results on the 12 transient for the single valve opening later this 13 afternoon.

14 In terms of the reactivity and power 15 distribution anomalies, these events are similar to 16 the scope for light water reactors. The flow related 17 events are not applicable to the design.

18 And for the 15.4.6, the inadvertent 19 decrease in warm concentration, it's postulated due to 20 a CVCS, chemical volume and control system malfunction 21 that causes a boron dilution transient to occur.

22 That's analyzed as part of the non-LOCA evaluation 23 model.

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92 1 for the wad ejection analysis where the power response 2 is calculated using the SIMULATE-3K code.

3 And the system response is analyzed with 4 NRELAP-5. And then VIPRE is used to assess crit --

5 margin to critical heat flux ratio.

6 MEMBER CORRADINI: And I think we're going 7 to see that topical in September.

8 MS. McCLOSKEY: I don't know.

9 MEMBER CORRADINI: Well, that's all right.

10 MS. McCLOSKEY: I mean, later -- yeah, 11 we'll see the topical later in the --

12 MEMBER CORRADINI: Later in the fall time 13 frame.

14 MS. McCLOSKEY: Later in the year. Yep.

15 MEMBER CORRADINI: All right. Thank you.

16 MEMBER BLEY: I know you answered Mike on 17 this earlier, but I kind of didn't follow it then.

18 Most of these are AOOs. But, you have an 19 IE in there. And I forget why you said that was.

20 MS. McCLOSKEY: And I forgot about that 21 one when we were talking about IEs.

22 MEMBER BLEY: That's okay.

23 MS. McCLOSKEY: That's the inadvertent 24 loading of a fuel assembly in it.

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93 1 when you picked up the most PRA? Or why is it falling 2 under?

3 MS. McCLOSKEY: That is one where the 4 radiological acceptance criteria are a fraction of 5 those acceptable for the accident dose.

6 MEMBER BLEY: Okay.

7 MS. McCLOSKEY: It was that. It wasn't 8 the steam generator tube failure. I was mistaken 9 earlier.

10 MEMBER CORRADINI: That's okay. I just 11 wanted to know what it was. And you answered it.

12 That's fine.

13 MEMBER DIMITRIJEVIC: Well, you have, I 14 mean usually we have a thorough, infrequently it's an 15 accident. And I thought that your postulate an 16 accident is something where we normally have an 17 accident. And the IE will be what you call infrequent 18 event.

19 But you actually classify everything with 20 AOO, right?

21 MEMBER CORRADINI: No.

22 MS. McCLOSKEY: No.

23 MEMBER DIMITRIJEVIC: No, no. There is 24 the one postulated accident where there is the one IE.

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94 1 decided to just classify everything as AOO and then 2 have a couple of exceptions which don't correspond to 3 this other usual division?

4 MEMBER CORRADINI: We haven't gotten to 5 15.6 yet. There's a lot of postulated events in 15.6.

6 MS. McCLOSKEY: In most cases the 7 initiating events are similar to those for operating 8 PWRs.

9 MEMBER DIMITRIJEVIC: Right.

10 MS. McCLOSKEY: And so we classified them 11 consistently.

12 MEMBER DIMITRIJEVIC: But no. For the 13 loss of offsite power is infrequent event. Which you 14 call AOO here. I mean, and well also --

15 MS. McCLOSKEY: Loss of offsite power is 16 typically analyzed as an AOO.

17 MEMBER DIMITRIJEVIC: Oh, really?

18 MS. McCLOSKEY: Yes.

19 MEMBER DIMITRIJEVIC: That's unusual. I 20 mean, not from my time. But anyway -- well, maybe my 21 time is, my time isn't right.

22 Well, the thing is so what is then like 23 infrequent event, like steam generator tube rupture.

24 What -- how would you classify that?

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95 1 we -- a steam generator tube rupture is actually 2 classified as a postulated accident. And that's on 3 the 15.6 slide.

4 MEMBER DIMITRIJEVIC: And most LOCAs you 5 classify right, as a postulated accident, right?

6 MS. McCLOSKEY: Yes. Yes.

7 MEMBER CORRADINI: They have a long list.

8 We're only half way through their list.

9 MEMBER DIMITRIJEVIC: Okay. I'm going to 10 check with the -- with the one with the specifications 11 in the PLA specimen. I'm going to see.

12 I mean, for me this is like almost 13 everything you've done is AOO. Which is like really 14 spec.

15 MS. McCLOSKEY: And this is in the design 16 basis event space. In terms of events that are unique 17 to NuScale that we conservatively classified as an 18 AOO, in some cases because we have sufficient margins, 19 it wasn't valuable to do a specific analysis of the 20 event frequency in order to justify classifying it as 21 something other than an AOO.

22 And so we treated it as an AOO.

23 MEMBER DIMITRIJEVIC: All right. Because 24 obviously some of those are not going to occur during 25 the, you know, hopefully during the plant life, so.

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96 1 MS. McCLOSKEY: Right.

2 MEMBER DIMITRIJEVIC: All right. Okay.

3 I mean, and you think that this is conservative from 4 your perspective. So, you know, what you need to so 5 express this, that's all right.

6 MS. McCLOSKEY: We go to 15.5 in terms of 7 events that increase reactor coolant system inventory, 8 for the NuScale design as a chemical volume control 9 system is the only system with capability to increase 10 RCS inventory during normal operation.

11 Then in --

12 MEMBER SKILLMAN: Well, what about control 13 rod drive cooling?

14 MS. McCLOSKEY: The control rod drive 15 cooling doesn't interface with the primary system.

16 And so it's -- that's a failure in the lines carrying 17 cooling to the control rod drives, is treated in the 18 containment flooding analysis.

19 MEMBER SKILLMAN: Okay.

20 MS. McCLOSKEY: In 15.1.

21 MEMBER SKILLMAN: Okay. Thank you.

22 MS. McCLOSKEY: In the -- in 15.6 then, we 23 have the events that decrease reactor coolant system 24 inventory.

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97 1 core cooling system valve or also -- we also call that 2 inadvertent opening of a reactor safety valve, is 3 addressed here.

4 And that's analyzed with the valve opening 5 event methodology that's an extension of the LOCA 6 analysis methodology. But we analyze it to 7 demonstrate that AOO acceptance criteria are met.

8 CHAIR MARCH-LEUBA: Is that going to be on 9 the same topical report as LOCA?

10 MS. McCLOSKEY: Yes. The methodology --

11 CHAIR MARCH-LEUBA: So there's like an 12 appendix?

13 MS. McCLOSKEY: Is now described in 14 appendix B of the LOCA topical report.

15 CHAIR MARCH-LEUBA: And has it always been 16 an AOO? Or is this a recent modification?

17 I'm asking in conjunction with an IAB, you 18 know, an inadvertent actuation block valve.

19 MS. McCLOSKEY: Um-hum.

20 CHAIR MARCH-LEUBA: If you consider this 21 an AOO and you survive it, why do we need an IAB?

22 MS. McCLOSKEY: The -- since we -- since 23 the NuScale design doesn't have a safety-related power 24 supply, then in design basis event space, any event 25 where you assume loss of power --

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98 1 CHAIR MARCH-LEUBA: Oh.

2 MS. McCLOSKEY: To occur, could be 3 postulated to transition to ECCS cooling.

4 CHAIR MARCH-LEUBA: But this one -- so 5 this event, 15.6.6 is loss of AC power, which trips --

6 well, I mean you trip the solenoid on the ECCS valve.

7 But IAB still holds it closed?

8 MS. McCLOSKEY: No. 15.6.6 postulates in 9 an inadvertent opening of one ECCS valve.

10 CHAIR MARCH-LEUBA: And release of 11 pressure.

12 MS. McCLOSKEY: While, the RCS is 13 operating at normal operating pressure and 100 -- and 14 2 percent power.

15 CHAIR MARCH-LEUBA: And using AOO 16 acceptance criteria.

17 MS. McCLOSKEY: Yes.

18 CHAIR MARCH-LEUBA: So then the question 19 is, why bother with the IAB?

20 MS. McCLOSKEY: So --

21 MEMBER CORRADINI: I think the answer 22 yesterday was they don't want to have too many invalid 23 actuations. They want to have valid actuations.

24 So that was their protection from an 25 investment standpoint.

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99 1 CHAIR MARCH-LEUBA: I thought the answer 2 is they were not sure. They were not -- that they 3 could survive this.

4 MEMBER CORRADINI: Oh. We can -- I'm not 5 sure, I can't remember who from the audience came up 6 and helped us. But that's what I remembered.

7 MEMBER BLEY: It was Storm. And -- oh, 8 you'll cover it. Go ahead.

9 CHAIR MARCH-LEUBA: Say your name.

10 MR. RAD: So, some of this information is 11 in --

12 MEMBER BLEY: Your name?

13 MR. RAD: I'm sorry, this is Zachary Rad, 14 Director of Reg Affairs, NuScale Power. Some of that 15 information in detail would be in the proprietary 16 session.

17 But, the objective here, so analyzing the 18 inadvertent opening as an initiating event is 19 basically just following the Chapter 15 protocol.

20 What we don't want is to have an 21 inadvertent opening of one, and another one opens 22 because of the basic design. I mean, so that's why 23 the IAB is in there. Multiple openings would be 24 negative.

25 MEMBER CORRADINI: Thank you.

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100 1 MS. McCLOSKEY: The other event I'll 2 highlight here is the small line break outside of 3 containment. That event is traditionally analyzed in 4 PWRs for radiological dose analysis.

5 And we also analyze it for dose analyzes 6 here. The small line break event in the NuScale 7 design is mitigated by closing the containment 8 isolation valves typically on a low pressurizer level 9 signal.

10 And then you have sufficient inventory 11 remaining in the reactor coolant system to support 12 cooling through the Decay Heat Removal System. And so 13 that's why you'll see that's part of the non-LOCA 14 event analysis.

15 MEMBER CORRADINI: A long list.

16 MS. McCLOSKEY: In terms of the thermal 17 hydraulic and fuel acceptance criteria, this table 18 summarizes acceptance criteria for minimum critical 19 heat flux ratio, the primary and secondary site 20 pressures, containment pressure, and event 21 progression.

22 This is generally consistent with the 23 standard review plan guidance, except that the NuScale 24 analysis methodologies are developed to demonstrate 25 that fuel cladding integrity is maintained, and MCHFR NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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101 1 remains above the limit for AOOs, infrequent events, 2 or postulated accident conditions.

3 The secondary side system pressure, design 4 pressure is equal to the primary system design 5 pressure at 2100 psi. And so you'll see in our 6 analysis results we have, we retain significant margin 7 to the secondary pressure limits.

8 The containment design pressure is 1050 9 psi. And we have those results in FSAR 621 to discuss 10 later this afternoon.

11 So, given all of these different 12 acceptance criteria and different event types, the 13 next two slides summarize the NuScale topical and 14 technical reports, describing the analysis 15 methodologies, to just lay out what these connections 16 are.

17 For a typical non-LOCA event, we built the 18 plant model in NRELAP-5. And that's used to calculate 19 the system from a hydraulic response to demonstrate 20 that the primary and secondary pressure criteria are 21 met. And that a safe stabilized condition is 22 achieved.

23 And that's in the non-LOCA topical report.

24 The boundary conditions from that system from a 25 hydraulic analysis are then provided to for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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102 1 subchannel analysis with VIPRE to demonstrate the 2 field cladding integrity are met.

3 And boundary conditions are from the --

4 are provided for the radiological analysis. And the 5 accident source term analysis topical report describes 6 the methodology to establish the source terms for the 7 radiological consequences.

8 MEMBER BLEY: Let me -- and Mike, I know 9 we have the topical on source terms later this year.

10 I think I know that.

11 MEMBER CORRADINI: We -- it has been 12 delivered to staff and they are starting to review it.

13 MEMBER BLEY: Okay. You don't know when 14 we're going to --

15 MEMBER CORRADINI: I'm not sure that we 16 have scheduled it.

17 MEMBER BLEY: Okay. What about the other 18 two topicals there? Have we --

19 MEMBER CORRADINI: We've already looked at 20 the subchannel topical report on how they've analyzed 21 it. We have not -- we -- it's somewhere in the fall 22 that we'll see the non-LOCA and the LOCA.

23 MEMBER BLEY: Okay.

24 MS. McCLOSKEY: In terms of LOCA and valve 25 opening events, we use NRELAP-5 to protect -- to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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103 1 predict the system and the hot channel response.

2 And in the NuScale methodologies the 3 acceptance criteria for these events are to 4 demonstrate margin too minimum critical heat flux 5 ratio. And that water level is maintained above the 6 top of the fuel.

7 The long term cooling analysis with ECCS 8 is analyzed with RELAP to demonstrate that water level 9 remains above the top of the core. And the core inlet 10 temperature remains sufficiently high that boron 11 precipitation is precluded.

12 And then the containment response is an 13 extension of the LOCA EM and also the non-LOCA EM for 14 the secondary side breaks to evaluate peak pressure 15 and temperature levels.

16 MEMBER CORRADINI: So, let me get a 17 clarification. Because I know the topicals are being 18 submitted to us for review -- I'm sorry, submitted to 19 the staff, excuse me, for review.

20 The technical reports are essentially 21 addendums to Chapter 15 analysis?

22 MS. McCLOSKEY: Chapter 15 and Chapter 6, 23 yes.

24 MEMBER CORRADINI: Okay. Which are not 25 being asked for the staff to look at them, only from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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104 1 the standpoint of audit. And they want to do 2 confirmatory calculations.

3 But they are not being submitted to the 4 staff for separate review.

5 MS. McCLOSKEY: They are part of the DCA.

6 MEMBER CORRADINI: Okay. That's what I 7 thought. I just wanted to make sure.

8 MS. McCLOSKEY: So this is part of the 9 Chapter 6 and Chapter 15 analysis.

10 MS. KARAS: Yes, this is Becky Karas from 11 Reactor Systems. We do review the technical reports 12 in concert with the DCA review --

13 MEMBER CORRADINI: But we would find your 14 evaluations buried inside the SEs?

15 MS. KARAS: That's correct.

16 MEMBER CORRADINI: Okay. Fine. Thank 17 you.

18 MEMBER DIMITRIJEVIC: For the Chapters.

19 MEMBER CORRADINI: For the Chapters, yes.

20 MR. SCHMIDT: This is Jeff Schmidt. It's 21 incorporated by reference to the DCA.

22 MEMBER CORRADINI: I'm telling this to the 23 -- I'm saying it out loud so the members, if they get 24 interested, because we felt that we didn't have enough 25 paper to read so far.

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105 1 So, just in case. The 700 pages was a 2 quick read.

3 MEMBER BLEY: Half of it's pictures.

4 MEMBER CORRADINI: That made it quicker.

5 MS. McCLOSKEY: The module protection 6 system actuations in the NuScale design were developed 7 and designed with a focus on this scope of design 8 basis events and the event response and the actuations 9 that we need to support the passive full response to 10 design basis events without crediting operator 11 actions.

12 And so we consider the design basis 13 events, the functions can be broadly classified into 14 four areas. The reactivity control either through 15 reactor trip or through isolation of the source of 16 dilute water that could be causing an inadvertent 17 dilution of event.

18 RCS and secondary site inventory control.

19 The containment isolation assures that sufficient 20 inventory is maintained to support ECCS cooling in the 21 event of a postulated pipe break or valve opening.

22 It also mitigates the loss of inventory 23 outside of the module, limiting dose consequences.

24 And the secondary isolation limits dose consequences 25 for events such as steam generator tube failure.

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106 1 And also assures that appropriate 2 inventory is maintained in at least one train of the 3 Decay Heath Removal System for DHRS cooling.

4 If normal secondary side cooling 5 unavailable, we've talked about the DHRS. And if 6 necessary, the emergency core cooling system 7 actuations.

8 And then finally primary side subcooling 9 and stability are protected by reactor trip.

10 CHAIR MARCH-LEUBA: So would this be a 11 good time to stop for lunch?

12 MS. McCLOSKEY: I think one more slide too 13 just --

14 MEMBER CORRADINI: A good transition 15 point?

16 MS. McCLOSKEY: It would be a good 17 transition point.

18 MEMBER CORRADINI: Okay. Thank you.

19 MS. McCLOSKEY: Because when we take those 20 overall functions of the module protection system and 21 look at the design basis event mitigation for the 22 different types of events that we've gone through, 23 that's what's summarized on this slide.

24 MEMBER CORRADINI: Okay.

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107 1 removal, transients are most often cause by a 2 postulated secondary side malfunction. So secondary 3 isolation is important there.

4 Decrease in heat removal transients 5 generate Decay Heat Removal System actuations.

6 Reactivity in power transients rely on reactor trip, 7 and demineralized water isolation.

8 Increase in RCS inventory events are 9 mitigated with isolation from the increase in 10 inventory source. Decrease in RCS inventory events 11 rely on containment isolation to control the inventory 12 available for ECCS cooling, and to mitigate dose 13 consequences and reactor trip instability.

14 MEMBER DIMITRIJEVIC: Shouldn't you have 15 like a measure control function also complication.

16 Like the, you know --

17 MS. McCLOSKEY: The pressure in the 18 reactor coolant system is limited by the reactor 19 safety valves.

20 MEMBER DIMITRIJEVIC: Right.

21 MS. McCLOSKEY: And so those are passive, 22 passively actuated valves. They aren't controlled.

23 So it's not -- that's not a function of the module 24 protection system.

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108 1 actuations --

2 MEMBER DIMITRIJEVIC: But module systems 3 have like inter-module protection function.

4 MEMBER CORRADINI: Is your green light on?

5 I'm not sure anyone can hear you.

6 MEMBER DIMITRIJEVIC: Oh, well. I will 7 just consider it as a multiple presentation fraction 8 where I was not pressure control part of the function 9 which we're looking at.

10 MS. McCLOSKEY: This was -- sorry, this 11 was focused on the module protection system.

12 MEMBER DIMITRIJEVIC: System, I see. All 13 right.

14 MEMBER CORRADINI: So this is a good time 15 for us?

16 MS. McCLOSKEY: Yes.

17 MEMBER CORRADINI: Okay. So, we're going 18 to try to catch up --

19 MR. BRISTOL: I have a quick follow up on 20 the --

21 MEMBER CORRADINI: Okay. Go ahead Ben.

22 MR. BRISTOL: Common cause failure. For 23 the level sensors, that's actually addressed in DCA 24 Section 7.1.5.1.2. And in Table 7.1-13 kind of has 25 the functions and how diversity and common cause NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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109 1 failure is addressed for the very sensors.

2 CHAIR MARCH-LEUBA: Can you read us the 3 section again?

4 MR. BRISTOL: Yes. So the DCA Section 5 Number 7.1.5.1.2.

6 CHAIR MARCH-LEUBA: Level five, yes?

7 MR. BRISTOL: Yes. And the DCA Table is 8 7.1-13.

9 CHAIR MARCH-LEUBA: Thank you. You know, 10 whenever you say that, we're going to look at it. And 11 then come back and say it doesn't say anything.

12 (Laughter) 13 MR. BRISTOL: Yeah.

14 CHAIR MARCH-LEUBA: I'm only kidding you.

15 MEMBER CORRADINI: That's not a given.

16 He's just being --

17 MR. BRISTOL: I'll be prepared for that 18 follow up. Thank you.

19 MEMBER CORRADINI: Okay. So why don't we 20 take a break until 1:15. So we'll have a very 21 efficient lunch.

22 (Whereupon, the above-entitled matter 23 went off the record at 12:33 p.m. and 24 resumed at 1:17 p.m.)

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110 1 are starting on Slide 24, okay?

2 MS. McCLOSKEY: All right, so as we move 3 into a summary of the analytical assumptions that are 4 applied in the Chapter 15 analyses, this slide shows 5 a map of the analytical pressure and temperature, 6 operation limits, and module perfection system limits.

7 And Dr. Corradini, I think this morning 8 you had a question about the normal T hot operation 9 versus --

10 MEMBER CORRADINI: It was somebody, I 11 don't remember. It doesn't matter.

12 MS. McCLOSKEY: Sorry, so there was a 13 question this morning and the normal T hot is shown 14 there in the green dot in the middle with the 15 operating range that's analyzed.

16 And then the T cold and T hot is about 590 17 degree compared to the hot leg temperature, module 18 protection system analytical limit of 610 degree to 19 protect the margin to subcooler.

20 CHAIR MARCH-LEUBA: My understanding is if 21 we make a mistake completing the flow, T hot will be 22 maintained at that point and T average will oscillate.

23 MS. McCLOSKEY: The plant control is based 24 on T ave and maintaining constant T ave.

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111 1 if -- which we will -- you're assuming a flow, a flow 2 rate. And we will have the flow rate the plant wants 3 to have and if it's only 0.1 percent off there's no 4 difference.

5 But if it's a significant difference, we 6 will keep T hot where it is now or we will move T 7 cool.

8 MEMBER CORRADINI: I think they've got to 9 check on that.

10 MR. BRISTOL: So, for the consideration of 11 operating margin, yes, there is some thought that goes 12 into where the analytical limit, at which point a trip 13 would come in, how much margin needs to account for 14 censor uncertainty, how much additional margin needs 15 to count for normal transience.

16 And so, yes, most likely the 59595 17 condition would be maintained as comfortable operating 18 margin and then T ave would flow from there.

19 MS. McCLOSKEY: And we have the high and 20 low pressure operating limits shown on there, the plus 21 or minus 70 PSI from normal operating condition at 22 1850.

23 And then the margin to high pressure and 24 low low-pressure analytical limits are also shown --

25 (Simultaneous Speaking.)

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112 1 CHAIR MARCH-LEUBA: And the red lines are 2 automatic scrams, correct? If a person goes below 3 1600, the protection system will scram?

4 MS. McCLOSKEY: Yes. I'll go on. In 5 terms of the analytical assumptions for Chapter 15 6 analysis, we'll talk about operator actions, single 7 failures, loss of power and then the scope of event 8 progression.

9 In the NuScale design for the Chapter 15 10 analyses no operator actions are required to achieve 11 the safety functions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after an initiating 12 event occurs.

13 We do consider operator errors in 14 identifying the scope of initiating events, such as 15 inadvertent signals that could occur.

16 But any operator actions that are allowed 17 by procedure will make the consequences of an event 18 less severe, and therefore are bounded by the Chapter 19 15 analyses.

20 The scope of multiple operator errors or 21 errors that result in a common-mode failure are beyond 22 design basis.

23 In the Chapter 15 analyses, we assume the 24 limiting single failure of a safety-related component, 25 and so we went through a systematic evaluation of our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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113 1 systems and the safety-related components to identify 2 failures that could affect the transient progression.

3 And due to both the simplified design in 4 terms of reducing the number of safety-related 5 components and redundancy of the safety-related 6 components, that reduces the overall scope of failures 7 that need to be evaluated in the Chapter 15 events to 8 determine the worst single failure for a particular 9 analyses.

10 And this slide summarizes the results of 11 that evaluation for the safety-related systems, 12 relevant single failures that occur in those systems, 13 and then some discussion.

14 So, in terms of redundancy in components, 15 an example of that are the containment isolation 16 valves on the CVCS piping.

17 Since there are two isolation valves in 18 series, single failure of one isolation valve doesn't 19 change the event progression and so it's not 20 specifically part of a calculation analyzed because it 21 doesn't affect the transient results.

22 CHAIR MARCH-LEUBA: One thing that we've 23 discussed with respect to that in other meetings is 24 what I call analytical redundancy. So, if this 25 particular scram fails, indirectly we will catch it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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114 1 with this other one in scram 2.

2 And there's a lot of flows in NuScale that 3 I don't think we're taking proper credit for. So, if 4 the pressure fails and doesn't scram you, you will get 5 a scram at high temperature.

6 MS. McCLOSKEY: The module protection 7 system functions are modeled in the Chapter 15 8 analyses.

9 And I'm not sure if you're getting at 10 considering common-cause failures of instrumentation, 11 and I think as Ben mentioned earlier this morning, 12 there's a summary of those results in Chapter 7.

13 CHAIR MARCH-LEUBA: Yes, and typically we 14 consider failure of one control rod to insert in all 15 Chapter analyses.

16 Is that in addition to this?

17 MS. McCLOSKEY: That's accounted for in 18 the scram worth.

19 CHAIR MARCH-LEUBA: The one rod out is not 20 a single failure, that's an assumption?

21 MS. McCLOSKEY: No.

22 CHAIR MARCH-LEUBA: On top of that you 23 have a single failure.

24 MS. McCLOSKEY: Yes, and on top of that we 25 consider power availability.

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115 1 CHAIR MARCH-LEUBA: Power is not safety-2 grade. But you have to consider DC power to fail and 3 is that because it was operating before so it should 4 stay there?

5 It would be a failure to make it fail?

6 MS. McCLOSKEY: We consider the 7 availability of DC power because it's not a 8 safety-related Class 1E power supply system.

9 CHAIR MARCH-LEUBA: But in all transients 10 you have the DC power on.

11 MS. McCLOSKEY: For many events that's 12 more limiting for the transient response because the 13 loss of DC power actuates the safety systems.

14 CHAIR MARCH-LEUBA: I didn't see that.

15 MS. McCLOSKEY: In terms of the single 16 failures that are considered in the analyses, in the 17 module protection system single failure of an 18 instrument channel is considered.

19 And that's relevant for asymmetric 20 reactivity events where a censor closer to the power 21 asymmetry could be assumed failed and that would delay 22 the response of the remaining instruments that are 23 operating.

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116 1 valve or a feedwater isolation valve. In those cases, 2 we take credit for the non-safety-related back up 3 valves in both of those lines.

4 Similarly, there are check valves in the 5 feedwater that prevent backflow in the case of a 6 postulated large feedwater line breaking inside of 7 containment.

8 And in the case of failure to close those 9 check valves, the non-safety-related back up check 10 valve in the lines is credited.

11 CHAIR MARCH-LEUBA: So, if we state that 12 in plain English, you're taking credit for non-safety-13 grade back up equipment?

14 MS. McCLOSKEY: And that's consistent with 15 NUREG 0138 and the guidance in Reg Guide 1.206 that 16 non-safety-grade equipment can be credited as back up 17 in the case of a single failure to the safety-related 18 component.

19 CHAIR MARCH-LEUBA: As long as the single 20 failure is not the initiating event. So, let's talk 21 about the two check valves. You have an initiating 22 event, Check Valve 1 fails, you can take credit for 2.

23 Now, if Check Valve 1 fails as your 24 initiating event, well, you need to have another one.

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117 1 fails as the initiating event, that would result in a 2 decrease in the feedwater flow, which is analyzed as 3 an initiating event and doesn't demand response of the 4 check valves.

5 The check valves are used in the case of 6 large feedwater line breaks inside containment where 7 you can get reverse flow from the intact steam 8 generator over to the break if the check valves were 9 not there.

10 CHAIR MARCH-LEUBA: I guess I will ask 11 tomorrow from the Staff about this single failure.

12 The first time I looked at it, it felt to me that 13 we're taking credit for non-safety-grade stuff.

14 MS. McCLOSKEY: And that's consistent with 15 the applicable regulatory guidance and the augmented 16 quality that's applied to these valves. The feedwater 17 reg valve and the backup means isolation valves are 18 seismic Class 1 valves.

19 They're part of the in-service testing 20 program and they're in tech specs with limiting 21 conditions for operability and surveillance 22 requirements.

23 CHAIR MARCH-LEUBA: So, other than not 24 having an N stamp in the outside casing, it's treated 25 as safety-grade?

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118 1 MS. McCLOSKEY: I think effectively.

2 CHAIR MARCH-LEUBA: The most important 3 thing is whether they're in tech specs in Appendix B.

4 So, whenever one of those equipments goes 5 out to service, if it's safety-grade, you have an LCO 6 and you have to start downgrading power and eventually 7 shutting down.

8 If it's not in tech specs, you won't.

9 MS. McCLOSKEY: The back up main steam 10 isolation valves and the feedwater valve are in tech 11 specs.

12 CHAIR MARCH-LEUBA: With an LCO?

13 MS. McCLOSKEY: Mm-hmm. The backup check 14 valve is also seismic Class 1 and part of the in-15 service testing program.

16 CHAIR MARCH-LEUBA: Because check valves 17 in particular have a history of failing a lot and 18 that's why we have --

19 MS. McCLOSKEY: The Chapter 15 analyses, 20 they're only relevant for the postulated accident 21 large feedwater line break.

22 CHAIR MARCH-LEUBA: Okay, we will ask the 23 Staff about their opinion on this.

24 MS. McCLOSKEY: In the emergency core 25 cooling system, we consider failure of one of the ECCS NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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119 1 valves to open either event valve or a recirc valve.

2 We also consider a module protection 3 system failure to actuate a division of ECCS, in which 4 case one vent valve and one recirc valve would remain 5 closed.

6 The IAB failing to close upon demand is 7 not treated as one of the single failures in the 8 accident analyses. That would occur in the case of a 9 postulated loss of DC power that results in a demand 10 to the ECCS valves.

11 CHAIR MARCH-LEUBA: So, you just told us 12 this morning that you were going to analyze that.

13 MS. McCLOSKEY: We analyze the inadvertent 14 opening of an ECCS valve, yes.

15 CHAIR MARCH-LEUBA: By itself?

16 MS. McCLOSKEY: By itself.

17 CHAIR MARCH-LEUBA: So, you don't want to 18 analyze it in conjunction with another initiating 19 event?

20 MS. McCLOSKEY: Correct.

21 CHAIR MARCH-LEUBA: Like loss of DC power?

22 MS. McCLOSKEY: Loss of DC power is part 23 of the loss of power assumptions that are treated.

24 So, loss of DC power is not a specific initiating 25 event.

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120 1 But if we go to the next slide on loss of 2 power, we consider loss of AC power at the time of 3 event initiation or at the time of reactor trip. And 4 in the case of loss of AC power, we consider whether 5 the DC power system is available or unavailable.

6 So, postulating a loss of AC power at the 7 time of reactor trip with loss of DC power at the time 8 of reactor trip, if the IAB were not there, then that 9 could postulate a demand on the ECCS system in 10 conditions far outside normal operation.

11 And this is all in the context of the 12 deterministic design basis assumptions for Chapter 15.

13 CHAIR MARCH-LEUBA: But if I understand 14 the system, AC power feeds the batteries and then the 15 batteries feed the instrumentation and so on and on 16 and everything.

17 The AC doesn't bypass. So, if the battery 18 cable out fails, that would be a failure of DC power 19 even though you have AC power.

20 MS. McCLOSKEY: In which case, that would 21 actuate the reactor trip DHRS containment isolation.

22 CHAIR MARCH-LEUBA: I'm not saying it's a 23 bad thing, it's an event that should have been 24 analyzed.

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121 1 loud.

2 MS. McCLOSKEY: And we consider that range 3 as part of the loss of AC power initiating event, but 4 since the safety systems are actuated, it doesn't 5 progress to more severe condition for the core.

6 CHAIR MARCH-LEUBA: Let me see if I 7 understand. One of the initiating events, 15.6.6, the 8 RRB opens by itself and truly opens.

9 And there are other scenarios that you can 10 imagine where the IAB would prevent that from opening.

11 That's what we have analyzed?

12 MS. McCLOSKEY: The event in 15.6.6 13 postulates some sort of a mechanical failure in the 14 valve. From the perspective of Chapter 15, worked on 15 get into the details of exactly what that postulated 16 failure is.

17 But in other scenarios we assume that the 18 valves operate as designed except in the case of the 19 single failures that we were talking about. Does that 20 answer -- I'm not sure if that gets to your question.

21 CHAIR MARCH-LEUBA: I was trying to figure 22 out the logic of assuming it fails on 15.6.6 and 23 assuming it doesn't fail on everything else.

24 MS. McCLOSKEY: We treat the initiating 25 events separately.

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122 1 MR. RAD: I can provide some insights 2 within the context of the regulatory framework. This 3 is Zachary Rad, Director of Reg Affairs, NuScale 4 Power.

5 Relative to the single failure criteria, 6 a component treated as passive is assumed to fail as 7 an initiating event but is not assumed as an 8 additional failure.

9 So, following that regulatory framework, 10 we've analyzed the inadvertent opening of the ECCS 11 valve.

12 Continuing on, because it's not considered 13 as an additional failure -- so under a single failure 14 criteria in active components you have to consider the 15 worst single failure of an active component in 16 addition to the initiating event.

17 Because this device is treated as passive 18 in our safety analysis, it is not considered to be an 19 additional failure relative to the initiating event.

20 So, if you have a LOCA you don't assume an 21 additional passive failure, an additional LOCA for 22 instance. In this case, it's a valve that we've 23 treated as passive relative to the single failure 24 criteria. We don't consider that as an additional 25 failure either.

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123 1 CHAIR MARCH-LEUBA: So, when do you 2 consider additional failures of check valves?

3 MR. RAD: We consider additional failures 4 on active components as we've treated them relative to 5 single failure criteria.

6 CHAIR MARCH-LEUBA: So, check valves is an 7 active component?

8 MR. RAD: Check valve is an active 9 component that is sometimes treated as a passive 10 component relative to the single failure criteria.

11 MEMBER CORRADINI: Does that make sense?

12 CHAIR MARCH-LEUBA: No.

13 MR. RAD: So, active in the fact that it 14 actually does move so technically it is active.

15 However, its treatment relative to the 16 single failure criteria is as if it were passive as it 17 relates to the information I just covered before, how 18 it's treated relative to subsequent failures.

19 CHAIR MARCH-LEUBA: IAB is also active?

20 MR. RAD: That's correct, it moves.

21 CHAIR MARCH-LEUBA: It moves.

22 MEMBER CORRADINI: But it would be part of 23 the initial failure under -- I'm looking for all of 24 their -- it's 15.6.6, inadvertent ECCS opening so they 25 assume a failure but then they're claiming that any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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124 1 subsequent IAB failures cannot be assumed because 2 they're not an active component.

3 That's what I thought you just said.

4 MR. RAD: That's correct. So, to put it 5 in a different context, how do we assume that was an 6 active component and the initiating event was the 7 inadvertent opening of an ECCS valve for whatever 8 reason?

9 And we have to assume the worst-case 10 single active failure, that evaluation would include 11 the failure of the additional failure of that IAB and 12 the potential opening, additional opening, of a second 13 ECCS valve.

14 Does that make sense? Especially given 15 that our assumptions are loss of AC power and because 16 we don't have safety-related DC power, and correct me 17 if I'm wrong, in our analysis we assume that's lost as 18 well.

19 CHAIR MARCH-LEUBA: It might not be lost.

20 MR. RAD: Right, may or may not be lost.

21 MS. McCLOSKEY: Yes.

22 CHAIR MARCH-LEUBA: Okay, I just wanted to 23 make sure it's on the record that it is an assumption 24 that is passive. And the Commission has not agreed on 25 that yet, correct?

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125 1 There is a request for --

2 (Simultaneous Speaking.)

3 MEMBER CORRADINI: It's still under 4 review. It's an open item also.

5 MEMBER BROWN: But there are a lot of 6 items which appear to be passive, which you might 7 think are active but are referred to as passive.

8 MEMBER CORRADINI: The logic, however 9 unusual, makes sense to me. I can take a passive 10 component and claim a failure but I can't then go on 11 and say other passive components can also be assumed 12 under single failure criteria.

13 That's what he said. I'm sure I said it 14 wrong.

15 (Simultaneous Speaking.)

16 MEMBER BROWN: It's just the rules of 17 doing this sort of thing.

18 MEMBER CORRADINI: Well, let me turn to 19 the Staff, I'll probably muddle it up.

20 MR. NOLAN: Just one point of 21 clarification, this is Ryan Nolan from the Staff.

22 It's agreed that the IAB is an active component, it's 23 whether it's subject to single failure or not.

24 So, I just want to make the clarification 25 --

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126 1 MEMBER CORRADINI: No, no, you said it 2 better than I did. I apologize.

3 MEMBER BLEY: When you say the loss of DC 4 power can either happen or not, I assume that means 5 you look at it and if that's the worst single failure 6 you can have you use it?

7 Or is there some other criteria?

8 MS. McCLOSKEY: It's in addition to the 9 worst single failure.

10 MEMBER BLEY: In addition to?

11 MS. McCLOSKEY: Yes.

12 MEMBER BLEY: And you still might consider 13 it failed. How do you decide which way you consider 14 it?

15 MS. McCLOSKEY: Whichever is more 16 conservative with respect to minimizing margin to the 17 acceptance criteria.

18 MEMBER BLEY: So you go both ways?

19 MS. McCLOSKEY: Yes.

20 MEMBER BLEY: In all of Chapter 15?

21 MS. McCLOSKEY: Yes.

22 MEMBER BLEY: Okay, thank you.

23 MS. McCLOSKEY: So, the next couple of 24 slides go into that loss of power assumption and how 25 it's treated, and what effect it has on the NuScale NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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127 1 design event progression.

2 Because for non-LOCA events, the power 3 availability affects whether the ECCS valves actuate 4 and then at what time they are postulated to open.

5 For LOCA-type events, the power availability affects 6 the time they actuate and when they open.

7 So, next slide.

8 MEMBER CORRADINI: Can I say it a little 9 bit differently just so...So, the only thing in 10 contention in terms of the assumptions, Staff and you 11 are on the same page in terms of assumptions except 12 for the open item relative to the IAB?

13 In terms of the initiating assumptions you 14 just went through?

15 MS. McCLOSKEY: Yes.

16 MEMBER CORRADINI: Is that a correct 17 statement?

18 MS. McCLOSKEY: Yes, I believe so. We had 19 several RAIs related to the backup --

20 MEMBER CORRADINI: I figured that's why 21 you made it green.

22 MS. McCLOSKEY: -- back up valves but I 23 think that's closed.

24 MEMBER CORRADINI: I just wanted to make 25 sure we summarized it in an appropriate manner, that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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128 1 all. Thank you very much.

2 MS. McCLOSKEY: In the case for a non-3 LOCA-type event that the AC power is available and DC 4 power is available, then very early on in the event 5 progression the module protection system will actuate 6 reactor trip and the decay heat removal system if 7 necessary, if normal secondary side cooling is 8 unavailable.

9 And then shortly thereafter, on the order 10 of half an hour to an hour timeframe, stable DHRS 11 cooling is established.

12 If the cooling through the decay heat 13 removal system is very effective and depending on what 14 the initial levels and temperatures in the reactor 15 coolant system are, riser uncovery may also occur in 16 that short timeframe.

17 If the power remains available and there's 18 no credit for any operator actions to restore 19 conditions, that cooling is maintained for the 72-hour 20 duration.

21 If AC power is unavailable, then the 22 beginning of the event progression is the same and the 23 difference comes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the ECCS valves are 24 signaled by the module protection system to open. And 25 then the plant will transition to ECCS cooling.

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129 1 If both AC power and DC power are assumed 2 unavailable, then the ECCS valves are actuated at the 3 time that the DC power is assumed to be lost.

4 But the event progression is the same 5 because the ECCS valves are held closed by the IAB 6 until the DHRS cooling is sufficient to depressurize 7 the reactor coolant system below the release point, at 8 which point the valves open and it transitions to 9 ECCS-cooled.

10 CHAIR MARCH-LEUBA: And there is no issue 11 with timing? And the IAB set point is going to be 12 different on every valve so if an RV opens before an 13 RVV or an RVV opens before an RV it doesn't make any 14 difference?

15 Because when we lose AC power, we trip 16 everything on a 24-hour timeframe at the same time.

17 When you are now relying on the IAB spring to go 18 clean, each valve will trigger at a different time.

19 Is there any possibility of messing up?

20 Because I would prefer to open the RVVs first and then 21 the RRVs.

22 MS. McCLOSKEY: In the case of the non-23 LOCA event progression, by the time that the RCS is 24 depressurized sufficiently, you've been tripped, for 25 a couple of hours you're at decay heat levels.

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130 1 And a significant amount of the initial 2 energy in the RCS has been transported to the cool.

3 So, that's not a transition that --

4 (Simultaneous Speaking.)

5 CHAIR MARCH-LEUBA: The DHRS is cooling 6 slowly, the pressure in the vessel is the saturation 7 pressure at that temperature.

8 So, slowly the pressure is -- and there 9 will be one valve that will open first and the others 10 will still be closed.

11 And then there will be a sudden 12 depressurization and all of them will open within a 13 second or two. But does it make a difference, the 14 order?

15 MS. McCLOSKEY: With respect to margin to 16 the acceptance criteria, I'd say no and that event is 17 bounded by the initiating event of the valve opening 18 while it power-conditions.

19 CHAIR MARCH-LEUBA: The fact that once you 20 open one, within a couple of seconds everything will 21 open and you are in decay heat?

22 MS. McCLOSKEY: Yes, in decay heat you're 23 depressurizing. All right, the next slide has the 24 LOCA event progressions.

25 CHAIR MARCH-LEUBA: I'm sorry, for all of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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131 1 these you have assumed CVCS not to be operational 2 because it's not safety-grade.

3 But if you keep it going and maintaining 4 water level and pressure, which it could because it's 5 non-safety grade, would it be a bad thing? IAB will 6 never click?

7 So, you have this same progression event 8 but now the operator didn't look, CVCS continued to 9 maintain the water levels so you are adding water or 10 how you isolate the -- you probably do not isolate.

11 MS. McCLOSKEY: Go ahead.

12 MEMBER CORRADINI: You would have a 13 containment isolation I would think.

14 MR. BRISTOL: If there's no DC power then 15 yes, the modules have been isolated so the reactor is 16 tripped, DHR is actuated and containment is --

17 CHAIR MARCH-LEUBA: There has to be a loss 18 of power.

19 MR. BRISTOL: -- containment isolated, 20 yes.

21 CHAIR MARCH-LEUBA: But if you have a 22 scram without a loss of power?

23 MR. BRISTOL: And we'll get into this in 24 some of the transient results but yes, there are 25 definitely cases where power available and normal NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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132 1 control systems available is more limiting than not.

2 One specific event is reactivity events.

3 Without pressure control, those trip very quickly on 4 high pressure so it's actually more conservative to 5 assume the spray comes on and maintains pressure 6 control and we eventually progress to a trip condition 7 on overpower over temperature.

8 CHAIR MARCH-LEUBA: Say you have a forced 9 scram for no reason whatsoever, everything else is 10 working.

11 CVCS continues to maintain water level in 12 the pressurizer and DHRS comes on and you isolate the 13 secondary. Will CVCS continue to maintain level and 14 dilute the boron?

15 MR. BRISTOL: With reactor trip, demin 16 water is isolated. So, there's certain conditions in 17 which the demin water supply is available unisolated.

18 CHAIR MARCH-LEUBA: It has to be 19 unisolated only? The reactor trip will isolate clean 20 water?

21 MR. BRISTOL: Mm-hmm.

22 CHAIR MARCH-LEUBA: What I'm coming to 23 hear is there are combinations of things, there are 24 combinations of situations, that work and doesn't 25 work.

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133 1 I'm not sure that you have considered all 2 of them in the recommendation.

3 MEMBER CORRADINI: I got the impression 4 they were trying to consider the limiting. They sure 5 haven't considered all combinations. But I thought 6 you were trying to consider the limiting ones.

7 Am I misunderstanding?

8 MR. BRISTOL: The focus is definitely on 9 the limiting ones, particularly for the purposes of 10 Chapter 15. When we get into the topical reports, we 11 can get into some of the sensitivities that led us to 12 conclusions of the limiting events.

13 A lot of sort of the interesting thought 14 exercises get into the progressions for the extended 15 cooling events.

16 And I think when we get into that 17 discussion later this afternoon, we'll kind of discuss 18 how that progresses and why all the events sort of 19 collapse into a pretty consistent trend.

20 But yes, we've given consideration to 21 postulated CVCS malfunction, extended cold water 22 injections, things of that nature.

23 And that's why we have some of these 24 protection systems that are a little bit unique to 25 NuScale, to ensure that we can have that walk-away NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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134 1 safe story of MPS will isolate things if operators 2 aren't there and paying attention to make sure that we 3 don't just have an unmanaged or unoperated module.

4 And we're trying to make everyone safe.

5 CHAIR MARCH-LEUBA: I'll save it for this 6 afternoon but I'm already thinking of a transient that 7 you haven't done.

8 MR. BRISTOL: Okay.

9 MR. INFANGER: This is Paul Infanger. The 10 inadvertent operation or misoperation of CVCS is an 11 event that's analyzed in 15.5. We looked at all the 12 combinations of what CVCS could --

13 MS. McCLOSKEY: For the increase in 14 inventory event.

15 MR. INFANGER: Increase in inventory and 16 we looked at boron dilution in 15.4.6.

17 MS. McCLOSKEY: The loss of power 18 progressions in a LOCA-type event are typically early 19 in the transient. The module protection system 20 actuates the reactor trip, containment isolation 21 because containment is isolated.

22 The decay heat removal system is also 23 isolated in the design and then ECCS valves are 24 actuated on high containment level in the case of all 25 power available.

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135 1 And for a realistic event progression, 2 they would be expected to open at that time. And then 3 ECCS cooling is established for the 72-hour duration.

4 If AC power is unavailable, that event 5 progression is basically unchanged. If DC power is 6 assumed to be unavailable as well, then the ECCS 7 valves are actuated by the loss of DC power.

8 And we credit the IAB function to hold 9 them closed until the RCS is sufficiently 10 depressurized, and then the valves will open a little 11 bit earlier to establish ECCS cooling.

12 In terms of the design basis event 13 progressions, the short-term analyses are analyzed 14 from the event initiation until a safe, stabilized 15 condition is reached, by which we mean that the 16 initiating event has been mitigated by the module 17 protection system actuations that are expected to 18 occur.

19 We've demonstrated that margin to our 20 acceptance criteria has been met and system parameters 21 such as inventory levels, temperatures, or pressures 22 are trending in a favorable direction.

23 Either the inventory levels are going up 24 or have stabilized, temperatures and pressures are 25 trending down.

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136 1 And then after the safe, stabilized 2 condition is reached, we get into the longer-term 3 analyses of ECCS long-term decay and residual heat 4 removal, the return to power analyses, and extended 5 DHRS operation.

6 CHAIR MARCH-LEUBA: I may need your help 7 on this to put this in proper English. But I'm 8 concerned by a little bit of paperwork and we have 9 mode one, mode two, mode three, mode four.

10 Mode two is hot shutdown that requires you 11 to have a temperature greater than 425 and mode three 12 is safe shutdown which can have a temperature lower 13 than 420.

14 But my suspicion is that to go to mode 15 three to safe shutdown, tech specs will require a 16 boron concentration before the operator can go from 17 mode two to mode three.

18 At least Chapter 4 has a table that 19 calculates how much boron you need to have to go into 20 safe shutdown. The table says unless you have 300 PPM 21 you are not in safe shutdown.

22 It's a function of exposure. So, my 23 suspicion is that's going to propagate into tech 24 specs. Now, by going into the passive cooling on 25 DHRS, the operator is initiating a transfer from mode NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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137 1 two to mode three.

2 You already scrammed, you are in mode two.

3 It initiates a transfer from mode two to mode three 4 without establishing the PPM requirements before he 5 does that. And in Matt's plan that would result in a 6 fine.

7 Is that correct?

8 MEMBER SUNSERI: I understand what you're 9 saying but I'm looking at the table right now and the 10 reactivity conditions are the same for both modes, so 11 less than or equal to 99.

12 So, you don't have to go to the shutdown 13 margin, the cold shutdown margin, if you know what the 14 transition mode looks like.

15 CHAIR MARCH-LEUBA: If you look at the 16 previous table, there is a PPM required to maintain 17 0.99 on Chapter 4. I'm just putting it out there.

18 There's a possibility that by turning 19 passive cooling on, you're actually moving from mode 20 two to mode three. If mode three has some 21 requirements different than mode two, then you are in 22 violation of procedures.

23 And if it does not have different 24 requirements, why have two modes? I'm just putting it 25 out there because if that happens in a control room NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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138 1 today, you will get a fine.

2 MS. McCLOSKEY: I think I would defer that 3 to Operations and our tech spec folks.

4 CHAIR MARCH-LEUBA: It's on the record, I 5 can read it afterwards. But there is a mode two and 6 a mode three. They're different. There has to be 7 some difference between the two.

8 MR. INFANGER: Without operator action, 9 you will transition to mode three and if all rods are 10 in, you will be subcritical under all conditions.

11 If one rod is not fully inserted, there 12 are some scenarios where you could never return to 13 power and that is analyzed in 15.0.6.

14 CHAIR MARCH-LEUBA: Okay, you guys 15 continue, let me look at the table in Chapter 4.

16 MS. McCLOSKEY: Let's go on to the next 17 slide.

18 So, although the discussion of the topical 19 reports are deferred to later in this qual, we did 20 want to provide a high-level picture of the system's 21 thermal hydraulic methodologies and the links between 22 them, starting with NRELAP5, which is the engine of 23 our system thermal hydraulic analysis for Chapter 15.

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139 1 phenomena in systems. And that's what we call the 2 NRELAP5 code.

3 The code description is provided in the 4 LOCA topical report, along with most of the 5 validation.

6 The LOCA EM was developed following Reg 7 Guide 1.203 and was extended for analysis of other 8 events, such as the valve-opening events and to focus 9 on other acceptance criteria including containment and 10 the long-term cooling analysis.

11 The non-LOCA topical report leverages that 12 code description and validation described in the LOCA 13 topical report based on considering differences 14 between the high-ranked phenomena for the different 15 types of events.

16 And then the containment response 17 technical report is an extension of these evaluation 18 models with focus on the containment pressure and 19 temperature acceptance criteria for both the primary 20 release events and the secondary side pipe-break 21 events.

22 And then we have overcooling, return to 23 power, and event progression from either the decay 24 heat removal system operation or ECCS and long-term 25 cooling with ECCS as an extension from the LOCA EM.

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140 1 That's, big picture, how this fits 2 together.

3 MEMBER CORRADINI: The way I read the 4 chart, just to repeat it back to you so I'm 5 understanding, NRELAP is being used in all of the 6 various ways but in a different set of assumptions or 7 procedures?

8 NRELAP is the tool but it's used with 9 different assumptions instead of protocols depending 10 upon the application?

11 MS. McCLOSKEY: Primarily, yes, some 12 different protocols. Primarily different biasing of 13 initial and boundary conditions and different 14 requirements to model secondary side breaks, those 15 sorts of --

16 CHAIR MARCH-LEUBA: So, the different 17 boxes identify different conservatisms that you have 18 to input on the input of NRELAP? Like, for example, 19 you have to use Appendix K?

20 MS. McCLOSKEY: Right, that's part of the 21 LOCA topical report. It's not part of the non-LOCA 22 topical report.

23 CHAIR MARCH-LEUBA: But it tells you what 24 input to provide to RELAP? Instead of providing best 25 testing, you provide Appendix K?

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141 1 MS. McCLOSKEY: Mm-hmm.

2 CHAIR MARCH-LEUBA: And are you going to 3 describe the changes, the new RELAP versus the old 4 RELAP? Are you going to describe those?

5 MS. McCLOSKEY: I think in the closed 6 session this afternoon our subject-matter experts at 7 Corvallis speak to a high-level summary of the 8 changes.

9 CHAIR MARCH-LEUBA: I'm personally more 10 interested in why they were made so late. Did we find 11 an efficiency analysis that triggered the need to do 12 it?

13 MS. McCLOSKEY: No, there were --

14 CHAIR MARCH-LEUBA: It's closed so let's 15 wait.

16 MS. McCLOSKEY: Some of the changes were 17 error corrections that we identified and needed to 18 correct as part of the normal process of development 19 and maintenance.

20 CHAIR MARCH-LEUBA: Typically, when you 21 have an error correction you don't rerun the whole 22 Chapter 15. You have outweighed Chapter 15 against 23 the error which did not affect it.

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142 1 that we are taking.

2 CHAIR MARCH-LEUBA: If we start doing 3 this, you'll never finish. We'll never finish.

4 MS. McCLOSKEY: Just one slide on the LOCA 5 EM development, the topical report was developed 6 following the Reg Guide 1.203 evaluation model 7 development and assessment process.

8 We developed PIRT to identify high-ranked 9 phenomena for the LOCA pipe-break events that is also 10 applicable to the valve-opening events, focused on the 11 short-term response.

12 We developed an assessment basis for 13 NRELAP5, including the separate effects test and 14 integral effects test to address these high-ranked 15 phenomena. Unique phenomena were addressed by 16 NuScale-specific tests in items such as the steam 17 generators.

18 The code development, as I said 19 previously, was developed in RELAP5-3D and we 20 performed an applicability evaluation for the EM 21 including bottom-up evaluation of the models and 22 correlations in the code, and a top-down analysis 23 focused on the performance against the IEDs.

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143 1 Reg Guide 1.203 and applying a graded approach, 2 leveraging that code description and much of the 3 validation that was described in the LOCA topical 4 report.

5 We did a PIRT identifying high-ranked 6 phenomena considering all the different types of non-7 LOCA events.

8 And then we did a GAP analysis for those 9 high-ranked phenomena compared to the high-ranked 10 phenomena addressed in the LOCA analyses to identify 11 what had to be addressed in addition for the non-12 LOCA-specific analyses.

13 We did do some additional NRELAP5 code 14 validations, focused primarily on the decay heat 15 removal system and the integral non-LOCA response.

16 MEMBER CORRADINI: Can I take you back?

17 You don't have to go back in the slide but the way 18 you've applied NRELAP is using Appendix K assumptions, 19 not using best estimate assumptions? Am I remembering 20 correctly?

21 MS. McCLOSKEY: Yes, that's correct.

22 MEMBER CORRADINI: Does that influence 23 some of the changes you made between 13 and 14?

24 MS. McCLOSKEY: No.

25 MEMBER CORRADINI: Okay.

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144 1 MS. McCLOSKEY: All right, and now I'll 2 turn it over to Ben to talk to several of the 3 transient example results that are presented in 4 Chapter 15.

5 MR. BRISTOL: Yes, thanks. So, we have 6 these grouped. I'll try to take this at a little 7 faster pace and just, obviously, you can slow me down 8 when you'd like to.

9 So, analysis results, we have different 10 event types. We'll start with just walking through 11 the sections in 15, how they're categorized.

12 The first type is increase in heat removal 13 by secondary or we call it coolant events. So, 14 there's a variety of them there highlighted. We are 15 going to walk through an example of an increase in 16 feedwater flow event.

17 We've got the limiting analysis results 18 summarized with the acceptance criteria in this table.

19 Are you driving? You're driving.

20 PARTICIPANT: Unless you want to.

21 MR. BRISTOL: I can. There, transition 22 complete. Okay, so for the case that we've presented 23 in the FSAR, this is 100 percent increase in feedwater 24 flow event. So, that's initiated at time zero.

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145 1 by a low-steam super-heat trip so analytical limit is 2 reached. That correlations pretty closely, this is 3 why it's a limiting event, to the high reactor power 4 trip being reached.

5 And we get a reactor trip signal, DHRS 6 actuation, and peak pressure occurs a little bit of a 7 while later.

8 CHAIR MARCH-LEUBA: What initiates DHRS?

9 MR. BRISTOL: Low-steam super heat.

10 CHAIR MARCH-LEUBA: And it's automatic?

11 MR. BRISTOL: That's correct.

12 (Simultaneous Speaking.)

13 CHAIR MARCH-LEUBA: In real life you would 14 expect the operator to take over that reactor and 15 control it? Or would you leave it like that?

16 This is the licensing basis, no hands off, 17 but in real life would you expect the operator to 18 defeat the containment isolation and take it off DHRS?

19 Or just leave it on DHRS?

20 MR. BRISTOL: I think it depends on what 21 the operators are doing but if this is the only module 22 that's being influenced, one of the key things that 23 they try to do is stop the cool-down.

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146 1 to the shrinkage that is caused by the cool-down.

2 So, maintaining inventory in the 3 pressurizer is one of the key parameters and the best 4 way to do that is to maintain temperature.

5 CHAIR MARCH-LEUBA: So, you would expect 6 the pressure to go there and defeat DHRS, open up the 7 secondary and start controlling?

8 MR. BRISTOL: If the conditions are 9 correct, they may consider closing the DHRS actuation 10 valves if they can reestablish normal feed, in which 11 case then the primary temperature can be controlled.

12 CHAIR MARCH-LEUBA: But what I'm talking 13 about is, this is the perfect time to tell you on the 14 record, that the table I was talking about before is 15 Table 4.3-2, nuclear parameters for a cycle.

16 And at the bottom of the table it says 17 boron concentration, for safe shutdown you need 1164 18 PPM at the beginning of the cycle, 240 at the end of 19 cycle.

20 It feels to me that somebody when they 21 made these calculations were planning to define all 22 three requirements to have those PPMs. By doing what 23 you just did now, you transfer the core to mode three 24 without achieving those requirements.

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147 1 perfectly okay to not have those requirements but if 2 you write your tech specs with this PPM for mode 3 three, the transient you just ran now will serve you 4 a fine, will put you in the yellow column of the 5 evaluation of your plant and you'll have to pay 6 $50,000.

7 So, somebody has to see the logic of 8 what's the difference between mode two and mode three 9 and is there a requirement for mode three, and is this 10 mode three -- because you're going into mode three.

11 You're going to go below 420 Fahrenheit.

12 MEMBER SUNSERI: I don't know about that, 13 Jose. Just thinking about how the plants I run work, 14 you have a reactor trip. The tech specs are laid out 15 the same, 0.99 K effective or whatever, right?

16 So, you know you're shut down if you 17 verify all your rods are in or if one's stuck or 18 whatever. Now the question becomes what's my shutdown 19 margin and how is xenon affecting that and all that 20 kind of thing?

21 So, you do a shutdown margin calculation.

22 You plug in the 0.99, you plug in your rod 23 configuration, you plug in your temperature and all 24 the reactivity configurations, and you compute what 25 your boron concentration is.

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148 1 And then you adjust it as necessary over 2 time so you get xenon-free or an equilibrium scenario 3 or whatever it is.

4 CHAIR MARCH-LEUBA: I don't think it's a 5 real problem, I think it's a paperwork problem and 6 somebody that does operations, ask him to read the 7 transcript and see if he understands what I'm saying.

8 Let me repeat it one more, you guys have 9 mode two and three, which are different.

10 They may have different requirements and 11 if you follow what you just described, you jump from 12 mode one to mode three and hands off, didn't touch 13 anything, the possibility exists of you violating your 14 procedures.

15 So it is incumbent on you to write the 16 procedures correctly.

17 MEMBER SUNSERI: But they're going to go 18 from mode one to mode two.

19 CHAIR MARCH-LEUBA: No, they'll go to mode 20 three.

21 MEMBER SUNSERI: No, because mode three 22 has to be less than 420 degrees. You're going to get 23 there in time but you're not going to get there right 24 away.

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149 1 mode one to mode three, you skip over two, and you can 2 stay in mode three or you can go to mode four, which 3 is a colder temperature, right?

4 You have to change your boron 5 concentration if you're going to go colder.

6 CHAIR MARCH-LEUBA: The moment you scram 7 you go to mode two. Within an hour to two hours --

8 MEMBER SUNSERI: On this plant.

9 CHAIR MARCH-LEUBA: On this plant, so the 10 one here described. Within a couple of hours, you're 11 going to be in mode three conditions.

12 Depending on how you write, how you define 13 your modes, you might be in violation of something.

14 Look at it.

15 MEMBER CORRADINI: I think we can move on 16 and I think we've got that point.

17 MR. BRISTOL: Okay. All right, so we 18 walked through the event sequence.

19 This figure here, we've got the increase 20 in feedwater flow that the scales aren't the greatest 21 but we go from 80 up to 160, quickly isolate. Here's 22 reactor power as a function of time.

23 Takeaway events detected in the steam 24 generators are isolated before both steam generators 25 are filled, and in this figure DHRS is actuated. This NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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150 1 is RCS flow, RCS temperature.

2 So we've established stable cooling. This 3 is sort of the truncation criteria for the short-term 4 transient response.

5 CHAIR MARCH-LEUBA: It's important in this 6 transient that you stop -- you turn on DHRS before it 7 overflows.

8 MR. BRISTOL: That's correct.

9 CHAIR MARCH-LEUBA: So, the timing of the 10 scram, not the scram but the transfer to DHRS is 11 important.

12 MR. BRISTOL: That's correct.

13 CHAIR MARCH-LEUBA: Yes.

14 MR. BRISTOL: This figure kind of 15 illustrates that a little bit. We've got secondary 16 pressure which isn't really the point here.

17 The figure is the unimpacted or the steam 18 generator level as the increasing inventory is sort of 19 on this time scale here.

20 This is where we get the low super heat 21 isolation signal and then as DHR drains we have an 22 additional filling of the steam generator, even at a 23 collapsed level of 80 percent, DHR is still effective.

24 CHAIR MARCH-LEUBA: So, the steam 25 generator only fills to about 75 to 80 percent?

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151 1 MR. BRISTOL: That's correct.

2 CHAIR MARCH-LEUBA: So, that does not 3 impede the DHRS operation?

4 MR. BRISTOL: No.

5 CHAIR MARCH-LEUBA: At that level. What 6 level would it fail DHRS? Pretty high up?

7 MR. BRISTOL: Yes, I don't have that 8 number immediately offhand.

9 I know in some of the cases we look at, 10 not necessarily the limiting cool-down cases, but this 11 is an event where the feedwater isolation failure is 12 a limiting event from DHR performance perspective.

13 So, one train, if the feedwater isolation 14 valve were to fail, we're waiting on the feedwater reg 15 valve to close in order to mitigate that overfill.

16 That's a much slower timeframe valve so 17 the one generator will end up with a higher level than 18 the other generator and it degrades the performance 19 accordingly.

20 CHAIR MARCH-LEUBA: Do you guys remember 21 that if your steam generator secondary side is full 22 with water, DHRS doesn't work? If the liquid level 23 hits 100, DHRS will be rendered inoperable. And it 24 hits 80 percent.

25 MEMBER CORRADINI: 100 percent of what?

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152 1 100 percent of the active tube length? Or 100 percent 2 of all the way up to the -- I want to understand what 3 100 percent is.

4 MR. BRISTOL: The tube length.

5 CHAIR MARCH-LEUBA: So there's still 6 plenty of steam volume?

7 MR. BRISTOL: And that's primarily at high 8 RCS temperature conditions.

9 So, if you were looking at this from the 10 sort of minimum or maximum cool-down, as RCS 11 temperature starts to drop, we would quickly see the 12 impact of the increased inventory on performance.

13 CHAIR MARCH-LEUBA: But DHRS isolates 14 everything?

15 MR. BRISTOL: Yes, that's right.

16 CHAIR MARCH-LEUBA: When you close it, 17 whatever inventory you close, that's what you have.

18 MR. BRISTOL: That's right. Okay, so 19 final figures here, here's DHR performance I think.

20 Am I reading that right? And this is the VIPRE figure 21 for the transient.

22 Minimum MCHFR is down here, right at 23 reactor trip. So, the conclusion is DHR is still 24 functionally removing decay heat and MCHFR margin 25 exists.

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153 1 Okay, heat-up events, decrease in heat 2 removal events. Again, here's a summary table, you'll 3 see the RCS pressure in all of these is fairly 4 consistent. That's due to the capacity of the reactor 5 safety valve.

6 These cases all have the reactor safety 7 valve lifting. That mitigates the pressurization 8 response. And secondary pressures, the heat-up events 9 tend to be pressurization events as opposed to actual 10 temperature-driven events in most cases.

11 So, the MCHFR is non-limiting for these 12 event types. An example we're walking through I think 13 is loss of AC power and inadvertent actuation of DHR.

14 So, loss of AC, this is an event we 15 simulate where there's a simultaneous loss of 16 feedwater and turbine trip.

17 So, the loss of flow to this steam 18 generator as well as an increase in steam pressure or 19 steam pressurization on the steam side quickly results 20 in high-pressurized pressure.

21 With that big of a transient to the steam 22 generators, the downcomer side actually heats up and 23 swells pretty quickly. That causes a pressurization 24 response that's detected rapidly.

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154 1 because of those signals.

2 MEMBER CORRADINI: Which of the various 3 AOOs fills the generator the most? Is it the one we 4 just went through?

5 MR. BRISTOL: Increase in feedwater flow, 6 yes.

7 MEMBER CORRADINI: And if one were to --

8 okay, I'll stop there and I'll come back. Thank you.

9 MR. BRISTOL: So, RCS pressure response 10 over here and secondary side pressure response. So 11 the sharp decline here is the pressure increases 12 rapidly until the safety valve lists and then 13 decreases.

14 As DHR cooling is established, we see sort 15 of this stable cooling trend. One of the things about 16 the way the DHR actuates all of our -- for most of the 17 events, tube failure is an exception, the actual 18 actuation of DHR is what's driving the pressure 19 response.

20 So, as the secondary system is isolated, 21 then it reaches equilibrium saturated condition with 22 the RCS temperature. And so that's really what's 23 driving this pressure response.

24 It will keep pressurizing until it's 25 pretty close to the RCS temperature and so for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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155 1 limiting events of that category, RCS temperature is 2 the main driver, actually, of the secondary pressure 3 response.

4 Again, flow, RCS flow, and temperature 5 response with DHR actuation. And I sort of alluded to 6 that in this particular case the slight heat-up 7 pressurization actually creates the limiting MCHFR 8 conditions as the initial condition as opposed to some 9 transient CHF.

10 Inadvertent DHR, so when we're operating 11 at full power conditions the pressure drop across the 12 steam generator is actually a little bit higher than 13 the head, the level head, of the DHR system itself.

14 So, if one of the valves at the top were 15 to open inadvertently, there's actually a little bit 16 of a feedwater bypass that occurs.

17 And so this creates a pretty minor loss of 18 feedwater, a little bit of flow injecting into the 19 steam outlet portion of the steam generator. It 20 reaches an equilibrium condition fairly quickly.

21 In this particular case it takes 400 or 22 500 seconds for the high-temperature trip to be 23 reached, at which point we get, again, reactor trip 24 and DHR actuation. So, here's a little bit slower 25 transient response in terms of the figure.

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156 1 There's a slight increase in secondary 2 pressure as temperature is increasing, and then DHR 3 actuation. This kind of shows the steam level 4 response.

5 The feedwater bypass creates a little bit 6 of a level loss in the steam generators and the 7 difference there in levels is the impact of steam 8 generator versus the non.

9 Again, RCS pressure response, this is a 10 case where we're actually assuming the modular control 11 system is functioning to keep pressure maintained.

12 That creates the limiting temperature response.

13 If this were not assumed, you can see by 14 this progression we would reach the high-pressure trip 15 pretty quickly, much more quickly than the high-16 temperature trip. So, you see the temperature figure 17 over on the spot.

18 So, in summary, the DHR valve opens and 19 diverts some of the feedwater flow around the steam 20 generators but RCS pressures remain within acceptance 21 criteria.

22 Here's our CHF figure in reactor power.

23 So, you see over here there's a slight mismatch 24 between the reactor power and the steam generator 25 power which causes the heat-up event.

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157 1 Reactivity events, again, here's just a 2 quick summary table. Pressure response, these are 3 primarily evaluated for CHF.

4 The increasing power events generate the 5 limiting category of our CHF-limited events, in 6 particular we'll walk through the single rod 7 withdrawal.

8 That reacts a little bit like an 9 inadvertent bank withdrawal but it has a unique 10 peaking that's applied in the analysis and that's what 11 causes the local heat flex to increase.

12 CHAIR MARCH-LEUBA: This is from low 13 power?

14 MR. BRISTOL: These are from...I think the 15 single rod withdrawal is initiated at 75 percent 16 power. The peaking is just a little bit worse than 17 the 100 percent power.

18 But the bank withdrawal is analyzed for 19 sort of the startup transient that has the power --

20 CHAIR MARCH-LEUBA: But the single rod 21 withdrawal is from 75?

22 MR. BRISTOL: That's right. So, in these 23 cases we actually see that the slower reactivity 24 insertions generate more of a thermal hydraulic 25 transient and that's what generates the limiting NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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158 1 conditions.

2 So, it's a little bit of a chase between 3 the thermal hydraulic conditions getting to the trip 4 conditions versus the overpower conditions. This 5 particular case, as you can see, we reach high hot leg 6 temperature at about 150 seconds.

7 RCS pressure limit is reached shortly 8 after that, at which point the reactor is tripped and 9 DHR is actuated.

10 So, here's a figure of the reactivity 11 insertion as a function of time and then the power.

12 So, we start at 75 percent power and increase up to 13 just over 100 percent power.

14 The reactivity insertion overall and 15 temperature feedback kind of create that cap there, at 16 which point it would trip the reactor.

17 Pressure and temperature response for the 18 events, there's a slight pressurization the way that 19 the system is modeled. It's actually a simplification 20 that's applied to the spread.

21 This is another one where the pressure 22 control actually makes the event worse and so there is 23 some pressure control applied until that's overcome by 24 the increasing temperature.

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159 1 MCHFR for this event, or is that 62? I can't read 2 that far.

3 Okay, increase in RCS inventory, this is 4 actually a category of events that NuScale doesn't 5 really have very severe transients. We look at the 6 maximum, sort of an inadvertent actuation of the 7 maximum makeup capacity.

8 That's what analyzed here. The 9 consideration really is for a malfunction event where 10 the pressurizer level is increased above the nominal 11 condition to generate a reactor trip, at which point 12 we would take a loss of AC power and a turbine trip.

13 Could we create a pressure response that 14 looks somewhat different than the normal pressure 15 transients that we look at? So, that's basically what 16 is evaluated.

17 It's a pretty slow response, I don't have 18 any of the specific results in this presentation.

19 They're in the SR.

20 So, decrease in inventory events, I'll 21 spend a little bit more time here. We've got a tube 22 failure scenario we'll walk through and then an 23 inadvertent ECCS actuation.

24 I think this is the RRV and then we'll 25 compare that to the limiting LOCA scenario, kind of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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160 1 set up some of the other topics we'll get into later 2 this afternoon.

3 So, tube failures are detected and 4 mitigated by RCS level. It's a pretty slow event.

5 The first thing that comes in is the low pressurizer 6 level set point. That's actually a protection set 7 point for the pressurizer heaters.

8 So, 35 percent is the pressurizer level at 9 which the pressurizer heaters exist axially. And so 10 once the heater starts to uncover, we want to protect 11 them so there's a pressurizer heater trip and a 12 reactor trip that comes with that.

13 The containment isolation doesn't come 14 further on until 20 percent pressurizer level.

15 CHAIR MARCH-LEUBA: So, the reactor trip 16 gets tripped, it doesn't isolate anything?

17 MR. BRISTOL: That's correct. So, upon 18 reactor trip --

19 MEMBER CORRADINI: How much of it is a 20 change in physical level to go to what you said, 20 21 percent from normal?

22 MR. BRISTOL: Normal is at 60 percent.

23 MEMBER CORRADINI: So what is that?

24 MR. BRISTOL: Pressurizer is ten feet tall 25 I think.

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161 1 MEMBER CORRADINI: So, it's almost a foot 2 per...?

3 MR. BRISTOL: Yes.

4 MEMBER CORRADINI: Okay, thank you.

5 MR. BRISTOL: So, if the reactor trips and 6 the heater's tripped, it's a bit of a race but we get 7 to low low-pressurizer level and low-pressurizer 8 pressure pretty quickly.

9 In this case, low-pressurizer pressure 10 creates a DHR or containment isolation and DHR 11 actuation. And ultimately, that's what mitigates the 12 event.

13 So, we see here secondary pressure, 14 primary pressure, slight drift down in primary 15 pressure until reactor trip and then we see an 16 increase in DHR actuation.

17 So, with the reactor trip we start to 18 increase the secondary pressure. Here you can really 19 tell the difference between the impacted steam 20 generator and the non-impacted steam generator.

21 So, a scaled-out figure just with the 22 pressures, the impacted steam generator quickly 23 reaches equilibrium pressure with the primary system.

24 The other steam generator is on DHR and providing 25 coolant.

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162 1 So, just a figure of pressurizer level, 2 and steam generator level. Again, you can see here 3 upon secondary isolation the level and the impacted 4 steam generator dramatically increasing.

5 So, primary inventory remains well above 6 the top of the core in DHR. The unimpacted train in 7 DHR provides coolant.

8 Here is just a figure of the break flow 9 rate into the integral mass release. We'll get into 10 the radiological results but this is one of the inputs 11 that's provided downstream to the radiological 12 analysis for the dose consequences.

13 Inadvertent RRV opening, this is a very 14 rapid event.

15 MEMBER CORRADINI: Is this the one that 16 maximizes -- maximize is the wrong word. This is the 17 one that is the one that's most severe in terms of 18 depressurization?

19 MR. BRISTOL: No, the inadvertent RVV is 20 a much more rapid pressure transient in terms of the 21 RCS. This is the RRV so this is a liquid space, loss 22 of inventory.

23 It looks a little bit like the discharge 24 line break. This is the event that sets the limiting 25 containment pressure response analysis, and we'll get NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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163 1 into that later as well.

2 MEMBER CORRADINI: So, this maximizes the 3 peak pressure in containment?

4 MR. BRISTOL: That's correct.

5 MEMBER CORRADINI: Whereas the RVV 6 controls or is limiting in terms of the 7 depressurization rate?

8 MR. BRISTOL: That's right, yes. Yes, the 9 steam space release is a much faster factor loss than 10 the liquid space release.

11 CHAIR MARCH-LEUBA: But it does not 12 pressurize the containment as much?

13 MR. BRISTOL: That's right. And we'll get 14 into the containment response, but it's a volume 15 transient for the containment.

16 If we're increasing with liquid without 17 de-energizing the RCS, then there's more energy once 18 the ECCS actually actuates overall.

19 Okay, so in this case we get a flow.

20 there is some pressure transient but there is a flow 21 transient in response to the opening of the valve. We 22 get an immediate high containment pressure signal in 23 this case, and that's what causes the reactor trip.

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164 1 with that, reactor trip comes immediately as well as 2 DHR actuation and containment isolation.

3 And so this is a case we'll kind of see 4 here. This is the case where one RRV is open, the 5 rest go onto IAB immediately with the IAB close 6 function until we reach the IAB release condition, 7 which is the differential pressure.

8 And so that's where we see the rest of 9 ECCS valves opening shortly into the transient. Level 10 figures, I think the nominal level condition in this 11 particular case doesn't look at failure of one of the 12 other valves to open.

13 So, in terms of the top of the active 14 core, the RCS level is right there at about ten feet.

15 The containment level is a couple of feet above that.

16 Here's the flow response I could mention.

17 There's a short flow transient and then 18 with reactor trip we get -- oh, no, sorry, this is the 19 RRV flow rate. So, as the valve opens, containment's 20 pressurizing so the flow drops, and then after ECCS 21 actuation equilibrium is quickly reached.

22 Zoomed in figure on RCS temperature and 23 then this is the temperature response after ECCS 24 actuation.

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165 1 response. So, the limiting CHF condition occurs to 2 this little transient flow response there, it's 3 actually driving it.

4 So, MCHFR occurs in the first couple of 5 seconds and the margins increase after reactor trip.

6 The technique for calculating this we'll get into 7 further in later follow-up discussions, but we have 8 margins to the acceptance criteria.

9 CHAIR MARCH-LEUBA: So, am I reading this 10 correctly? At time zero the MCHFR is 1.5? Or is 11 there a drop that I don't see? Because I thought it 12 was closer to two.

13 MEMBER CORRADINI: I think it's higher.

14 MR. BRISTOL: So, this gets into the 15 differences in the RELAP calculational approach versus 16 the subchannel calculational approach.

17 MEMBER CORRADINI: -- using your approved 18 NSP...

19 MR. BRISTOL: Four.

20 MEMBER CORRADINI: Four, thank you.

21 CHAIR MARCH-LEUBA: So, this RELAP MCHFR 22 is bias low?

23 MR. BRISTOL: Mm-hmm.

24 CHAIR MARCH-LEUBA: Or simply the 25 correlation that you're seeing is bias low?

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166 1 MEMBER CORRADINI: This is Hench-Levy.

2 MR. BRISTOL: That's correct.

3 MEMBER CORRADINI: And you use that to 4 again be conservative? I didn't catch that in the 5 chapter.

6 MR. BRISTOL: I think we'll defer that 7 discussion to -- are we getting into that later today?

8 MS. McCLOSKEY: We don't have slides on 9 that later today because that's part of the LOCA 10 topical report.

11 MR. BRISTOL: Yes, we're giving an 12 overview of the LOCA topical in the closed session and 13 that's bordering on the proprietary so we can discuss 14 it --

15 CHAIR MARCH-LEUBA: At a high level, CHF 16 correlations are fuel-specific because they depend a 17 lot on the spacers? You're using a generic CHF 18 correlation for the real calculations that do not care 19 what the fuel is?

20 MR. BRISTOL: Not completely. The 21 analytical limit or the exceptions criteria is based 22 on --

23 CHAIR MARCH-LEUBA: But that's 24 uncertainties.

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167 1 fuel-specific CHF --

2 CHAIR MARCH-LEUBA: But the actual value 3 of CHF that you predict, with a fuel-specific CHF 4 correlation versus a generic can be off by 50 percent.

5 The generic ones tend to bound everything 6 because they don't account for the spacer turbulence.

7 And the fuel fabricator has spent a lot of money 8 creating that turbulence.

9 MR. BRISTOL: Certainly.

10 CHAIR MARCH-LEUBA: And are charging you 11 for it.

12 MEMBER CORRADINI: But just keep in mind, 13 when we discussed this a year ago we agreed that the 14 NSP4 had a limited range of applicability in terms of 15 pressure and flow.

16 And they're not under these conditions 17 within that range of applicability to pressure and 18 flow. So, my assumption was this was a conservative 19 application.

20 MR. BRISTOL: I think we are with --

21 MEMBER CORRADINI: Well, you fall out of 22 it. I mean you start off there but you fall out of 23 range.

24 MR. BRISTOL: Sure.

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168 1 wait until later?

2 MR. BRISTOL: Yes.

3 MEMBER CORRADINI: Okay.

4 CHAIR MARCH-LEUBA: So, now let's get into 5 the recreational complaining again about the IAB.

6 MEMBER CORRADINI: They still have a lot 7 more slides.

8 CHAIR MARCH-LEUBA: Yes, but let's do some 9 recreational complaining. Filling one IAB gives you 10 a 0.1 delta CHF, nothing compared to your margin where 11 you have two to start with.

12 You could superimpose this transient on 13 every other AOO and be perfectly okay and save a lot 14 of hassle with that IAB that is so complicated.

15 That's my ten cents. Keep going.

16 MR. BRISTOL: Thank you.

17 CHAIR MARCH-LEUBA: It's not worth the 18 fight. 0.1 CHF is not worth the fight.

19 MEMBER CORRADINI: I don't think it's that 20 straightforward but I understand what Member March-21 Leuba is saying.

22 It's the same thing I asked yesterday 23 relative to how all these work in concert.

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169 1 want to have it. But it's not bad, it's actually 2 pretty good. Keep going.

3 MR. BRISTOL: Okay, so to contrast, we'll 4 kind of walk through the LOCA scenario.

5 The limiting event from a level 6 perspective as presented in the FSR's ten percent 7 injection line break, this is a relatively slow 8 transient although the initial detection's quite 9 quick.

10 Again even very small breaks of high-11 energy lines in containment, containment's small, we 12 quickly get a pressure response. And let's see, so 13 this case also assumes loss of AC power at time zero.

14 We get a reactor trip.

15 So, with the loss of AC power we actually 16 get a pressurization response. The inventory loss is 17 not actually driving the event detection in this 18 particular case.

19 So, with the reactor trip, shortly after 20 that we get the high containment response, containment 21 isolation, and eventually low pressurizer level, low 22 pressurizer pressure response. And this is a case on 23 IAB.

24 So, one of the conservativisms in the LOCA 25 EM is DHR heat removal is not credited so there's a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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170 1 slight pressurization that occurs.

2 Eventually, the inventory starts to turn 3 the pressure around as inventory is increasing in the 4 containment vessel.

5 That reaches kind of an equilibrium condition 6 and eventually, we're waiting on IAB release out here 7 in the about 6000 second timeframe.

8 That causes a pretty rapid pressure drop 9 which is driving -- if we look over here, this is our 10 level response. It's at the time of ECCS actuation.

11 CHAIR MARCH-LEUBA: So, you barely, barely 12 hit core uncovery right there at 6000 seconds?

13 MR. BRISTOL: Yes, it doesn't actually get 14 to the top of the core from a collapse-level 15 perspective but it's starting to get close to it.

16 CHAIR MARCH-LEUBA: But you're likely to 17 have some flashing so the actual liquid is higher on 18 that?

19 MR. BRISTOL: Yes, there's a fair amount 20 of flashing. RCS is at saturated conditions, 21 obviously, at this point and so with this 22 depressurization there's some liquid flashing that 23 occurs.

24 CHAIR MARCH-LEUBA: And we are not 25 publishing here the CHFR because for LOCA that's not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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171 1 a concern?

2 MR. BRISTOL: I think we do publish CHFR 3 but it occurs at the initiating event. But CHF is 4 screened as part of the methodology through the entire 5 transient response but not specifically evaluated to 6 --

7 CHAIR MARCH-LEUBA: But it's not a 8 criteria?

9 MR. BRISTOL: That's right.

10 CHAIR MARCH-LEUBA: For LOCA.

11 MR. BRISTOL: It's evaluated. So, one of 12 the mechanisms for the Appendix K evaluation is 13 ensuring that CHF does not occur.

14 CHAIR MARCH-LEUBA: Sure.

15 MR. BRISTOL: As part of the LOCA --

16 (Simultaneous Speaking.)

17 -- so within the code there's a model 18 that's built in and it's screened to --

19 CHAIR MARCH-LEUBA: Screened to criteria.

20 If you didn't go over this, you'd have to do more.

21 MR. BRISTOL: That's right.

22 CHAIR MARCH-LEUBA: Before you go on, more 23 recreational complaining. Can you go back to Slide 24 68? And we see the CHFR starts at 1.5, right? Now, 25 go back to Slide 41?

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172 1 MR. BRISTOL: 31?

2 CHAIR MARCH-LEUBA: 41. Yes, that's the 3 one. CHFR starts at 2.8. What's the difference?

4 MR. BRISTOL: The initial conditions, the 5 models that are applied.

6 CHAIR MARCH-LEUBA: 2.8 versus 1.5?

7 MR. BRISTOL: Well, okay, so let me see if 8 I can do this --

9 CHAIR MARCH-LEUBA: Basically, what I'm 10 saying is make up your mind. Which is it?

11 MR. BRISTOL: 2.8.

12 CHAIR MARCH-LEUBA: Wouldn't this be worse 13 if you just have the 1.5? Indeed, I see an decrease 14 of almost 1.0.

15 MR. BRISTOL: That's right.

16 CHAIR MARCH-LEUBA: If you just have the 17 1.5, it will be bad.

18 MR. BRISTOL: Certainly.

19 CHAIR MARCH-LEUBA: So, why do you have 20 2.8 here and not on the other one?

21 MR. BRISTOL: The conservativisms that go 22 into setting up the channel that's evaluated in the 23 RELAP model are quite conservative and that's really 24 what drives the primary difference in the initial CHF 25 calculation.

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173 1 CHAIR MARCH-LEUBA: So, the DHRS heat 2 removal event has less conservativisms than the ECCS 3 inadvertent opening?

4 MR. BRISTOL: This calculation is 5 performed using the VIPRE subchannel methodology and, 6 therefore, it's going to calculate different local 7 thermal hydraulic conditions than the RELAP approach.

8 CHAIR MARCH-LEUBA: I hope VIPRE will give 9 you a lower ratio. This is rich, right?

10 MR. BRISTOL: Yes.

11 CHAIR MARCH-LEUBA: VIPRE is more 12 conservative, it will give you a lower number. Maybe 13 it's the other way around.

14 MEMBER CORRADINI: It's the other way 15 around. This is using NSP4.

16 MR. BRISTOL: That's correct.

17 MEMBER CORRADINI: So this, in theory, has 18 been done with their experiments at Stern with their 19 essentially reactor-relevant bundle in difference to 20 what is being used as a default within RELAP. Am I 21 close?

22 MR. BRISTOL: That's correct.

23 CHAIR MARCH-LEUBA: So, this is not the 24 RELAP CHFR?

25 MR. BRISTOL: This is not the RELAP CHFR, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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174 1 this is the VIPRE CHFR.

2 CHAIR MARCH-LEUBA: And the ECCS?

3 MR. BRISTOL: Is the RELAP CHFR.

4 CHAIR MARCH-LEUBA: And you used VIPRE 5 here because otherwise you couldn't survive it?

6 MR. BRISTOL: This event we could use the 7 RELAP approach as well and show a large margin.

8 CHAIR MARCH-LEUBA: It doesn't feel like 9 it. If you start at 1.5 but you lose 1 -- you lose 1.

10 You start at 2.8 and you scram at 1.8 or 1.9. You 11 lose one all in all in CHFR.

12 MR. BRISTOL: Understood.

13 CHAIR MARCH-LEUBA: It would have been 14 nice if Chapter 15 had been consistent or at least 15 properly advertise what you're using. And I assume 16 both methods are acceptable.

17 MEMBER CORRADINI: I think we've got to 18 ask the Staff that separately.

19 CHAIR MARCH-LEUBA: It feels like cheating 20 a little bit.

21 MR. LINGENFELTER: Andy Lingenfelter of 22 NuScale. I think, Jose, to answer your question, part 23 of that is wrapped into the topicals.

24 And while we would have loved to have done 25 the topicals first and Chapter 15 second, it didn't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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175 1 work out that way.

2 And so maybe when we get to the 3 proprietary section we can give you a little more 4 color on CHFR --

5 (Simultaneous Speaking.)

6 That might help.

7 CHAIR MARCH-LEUBA: So did you use 8 different evaluation methods --

9 MR. LINGENFELTER: Appendix K for LOCA and 10 we didn't use Appendix K for non-LOCA.

11 CHAIR MARCH-LEUBA: So, the ECCS open 12 valve you used the LOCA methodology?

13 MR. INFANGER: When we did the analysis in 14 RELAP for the non-LOCA, we did calculate an MCHFR and 15 we used that as a scoping for what would be the most 16 limiting events.

17 And then we ran VIPRE on those events to 18 fine-tune it and the VIPRE number is always a lot 19 lower than MCHFR.

20 CHAIR MARCH-LEUBA: So, this is not VIPRE?

21 MR. INFANGER: That is VIPRE. Yes, and 22 just ballpark, you had like two if you use the RELAP 23 calculation. So, the RELAP is much, much higher.

24 CHAIR MARCH-LEUBA: So RELAP would give 25 you 4.8?

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176 1 MR. INFANGER: Yes. No, the minimum would 2 be like 3.8.

3 MS. McCLOSKEY: And just to clarify, that 4 screening technique that's applied as part of the non-5 LOCA EM is different than the CHF evaluation that's 6 performed for the valve opening events using RELAP.

7 So, that's a little bit of a difference 8 there.

9 MEMBER CORRADINI: We've got to go to 10 closed session so we can talk this out.

11 CHAIR MARCH-LEUBA: Right, and I'll make 12 a note to ask the question when we revisit and keep 13 talking, because how come we have factors of two on 14 CHFRs for different methodologies? It would be nice 15 to be semi-consistent.

16 MR. BRISTOL: Okay.

17 MEMBER BLEY: I apologize, I was out for 18 some of the time. I don't think you have any slides 19 on that. 15A, you're not going to talk about that, 20 right? Or are you?

21 MEMBER CORRADINI: It's exempt out. We 22 haven't gotten there yet though.

23 MEMBER BLEY: But I'm going to leave 24 early, and I apologize for that as well and I'd like 25 to --

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177 1 MEMBER CORRADINI: But I think similar to 2 --

3 MEMBER BLEY: There are no slides on it.

4 MEMBER CORRADINI: Right, and similar to 5 APR1400 it essentially is exempt for a Chapter 15 6 analysis.

7 MR. BRISTOL: It is beyond design basis.

8 MEMBER BLEY: There's an argument in 9 Chapter 15 that is part of the submittal.

10 MEMBER CORRADINI: Right, but I think this 11 is consistent with past DCs.

12 MEMBER BLEY: They had an argument in 13 Chapter 15 about it? I don't remember that.

14 MEMBER CORRADINI: I've got the slides.

15 MEMBER BLEY: So, you're telling me to 16 shut up?

17 MEMBER CORRADINI: I didn't say that. I 18 didn't say that at all, I'm just simply saying it's 19 consistent, that's all I said.

20 MEMBER BLEY: I want to make a comment and 21 ask a couple short questions and then I'll be done 22 with it. The argument at the end in here is that ATWS 23 is covered by the PRA and it calculates a very low 24 frequency of ATWS, which I believe for the electronics 25 getting a signal.

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178 1 But most of the likelihood of failure to 2 scram in other systems is due to problems with the 3 rods going in. And I don't know that that was 4 covered. So, I'm suspicious of the 1.7-5 per year.

5 We're not going to argue the PRA here.

6 You begin with the discussion that links back to the 7 anticipatory turbine trip that you don't have, and I'm 8 thinking back 40 years to when that came into play for 9 PWRs with U-tube steam generators.

10 And the original ATWS calculations at 11 least for some of those saw a very -- well, before it 12 failed to scram, it kept running and used up the 13 inventory for certain transients, the inventory in the 14 steam generators.

15 Then all of the sudden the pressure goes 16 up faster than the relief and safety valves can 17 relieve and you were going to break something in the 18 system, so they came up with this anticipatory turbine 19 trip to prevent that.

20 In a design such as this where you don't 21 have any real inventory in the steam generators, I 22 don't think that would apply.

23 But should you have an ATWS -- and are we 24 going to cover that somewhere else? Maybe -- in your 25 design and the pressure starts taking off inside, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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179 1 have a couple nice big vent valves.

2 Does that take the edge off of the ATWS if 3 you actually have one? I'll go back at this in the 4 PRA the next time they come around but if you can say 5 anything about it --

6 MEMBER CORRADINI: So, can I help you a 7 little bit?

8 MEMBER BLEY: -- I'd appreciate it. I 9 don't know if you can or not.

10 MEMBER CORRADINI: There's two things 11 happening.

12 One, we did have a Staff audit calculation 13 that was presented last month by Dr. Yarsky that went 14 over this and I think the answer to the vent valve, 15 the pressure in the vent valve, is it does open, it 16 does have that.

17 (Simultaneous Speaking.)

18 MEMBER BLEY: -- the top off the pressure, 19 okay. I didn't remember that.

20 MEMBER CORRADINI: I also think there's 21 now -- I'm going to look at the Staff -- a completed 22 report by the Staff, a large report, very interesting, 23 on ATWS which we can get.

24 MEMBER BLEY: I wouldn't mind getting 25 that.

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180 1 MEMBER CORRADINI: In fact, I thought 2 Chris Brown sent us an email that that was available.

3 MEMBER BLEY: I don't know.

4 MEMBER CORRADINI: I can find the email 5 and send it out to you. He basically sent us a 6 download site --

7 MEMBER BLEY: For the ATWS?

8 MEMBER CORRADINI: For the proprietary, 9 for what Staff has done in terms of a whole range of 10 --

11 MEMBER BLEY: You have helped me, I 12 assume, as long as I can go through it.

13 (Simultaneous Speaking.)

14 MR. SCHMIDT: This is Jeff Schmidt from 15 reactor systems. So, that was done under Chapter 19 16 and Dr. Yarsky did that, that confirmatory. So, 17 reactor systems may have it or don't have it.

18 MEMBER CORRADINI: But I think we 19 requested it. Some time between last month and this 20 month, Chris sent an email out with a location. I can 21 resend the email.

22 MEMBER BLEY: I missed that if that came 23 out. That probably covers my --

24 MEMBER CORRADINI: It has a complete --

25 (Simultaneous Speaking.)

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181 1 MEMBER BLEY: -- except for the PRA stuff 2 which I'm a little nervous about.

3 MEMBER CORRADINI: Okay.

4 MEMBER BLEY: Thank you.

5 MEMBER CORRADINI: I'll send the email.

6 I'll make sure that Mike or I send the email to the 7 Committee again.

8 MEMBER BLEY: I appreciate it.

9 CHAIR MARCH-LEUBA: With an ML number 10 preferably?

11 MEMBER CORRADINI: Yes, there is an ML 12 number.

13 CHAIR MARCH-LEUBA: And it points to the 14 right place?

15 MEMBER CORRADINI: I was able to download 16 the report.

17 MEMBER BLEY: So, you've seen it?

18 MEMBER CORRADINI: I 've seen it. I 19 haven't read it, I had other homework.

20 CHAIR MARCH-LEUBA: All I know is the last 21 time I saw it, I saw the cover page and they told me 22 they couldn't give it to me.

23 MEMBER CORRADINI: We've got it.

24 MR. BRISTOL: Okay, I think that concludes 25 the transient portion of the presentation. We're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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182 1 going to go into the radiological doses next.

2 MR. INFANGER: This is Paul Infanger.

3 CHAIR MARCH-LEUBA: Before you start, is 4 this going to be similar to what we've done with the 5 LOCA and non-LOCA methodology?

6 Because there is a source term methodology 7 topical report that hasn't been reviewed yet.

8 MR. INFANGER: Right, and we're not going 9 to get into the topical report.

10 CHAIR MARCH-LEUBA: You're using the 11 results of that report assuming it gets approved?

12 MR. INFANGER: That's correct.

13 CHAIR MARCH-LEUBA: Is that what you're 14 doing now?

15 MR. INFANGER: That's correct. Okay, so 16 the radiological analysis, we used the standard 17 radiological dose consequences that are used in the 18 industry. We used Reg Guide 1.183 and we used that 19 for the acceptance criteria.

20 And if you look at the events, this is 21 right out of the guidance, and they talk about lost 22 coolant, accidents, fuel-handling accidents, rod 23 ejection accident. But the acceptance criteria is due 24 to damaged fuel.

25 However, in our events only fuel-handling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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183 1 accidents resulted in damaged fuel. We don't have 2 fuel damage in any of our other accidents.

3 The first three reactor loss of coolants, 4 fuel-handling accidents and rod ejection accident, 5 they talk about the radiation source being damaged 6 fuel. That's for the acceptance criteria.

7 But if it's a NuScale reactor, we don't 8 have any accidents of damaged fuel except the fuel-9 handling accident where it's assumed. And all of our 10 other events we use coolant activity, RCS activity 11 with iodine spiking as the source term.

12 DR. SCHULTZ: Paul, for the pre-incident 13 spike, what do you assume for the coolant activity?

14 What percentage of fuel failure do you have?

15 I've seen a couple of different numbers 16 related to that.

17 MEMBER CORRADINI: You're looking at a 18 fraction of fuel damage?

19 DR. SCHULTZ: It's the pre-incident spike, 20 yes. I presume that it's related to your technical 21 specification?

22 MR. INFANGER: We used the tech spec limit 23 for the initial coolant but then you have an iodine 24 spike on top of that, on top of the RCS allowable 25 load.

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184 1 DR. SCHULTZ: Okay, thank you.

2 MR. INFANGER: Okay, so for the 3 radiological consequences analysis using the 4 alternative source terms, we used essentially the same 5 recommendations in Reg Guide 1.183 to evaluate the 6 consequences of our design basis event, which is the 7 iodine spiking event, and also, a beyond-design-basis 8 core damage event.

9 And that's described in detail in the 10 accident source term topical report. For feedwater 11 line break, we reviewed it and found that the steam 12 line break had more limiting consequences so we used 13 that as a bounding event and didn't do a separate 14 feedwater line break dose analysis.

15 We also looked at the reactor pool 16 boiling, a very large reactor pool, and so Staff had 17 requested us to look at what happens if you had an 18 event where you had long-term boil off of that?

19 And we found that there's a small amount 20 of tritium in that water from refueling and things 21 like that, and we found that the dose was 22 insignificant.

23 We looked at potential shine to the 24 control room operators so if you're having an event in 25 the reactor building, there is potential for some NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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185 1 reactivity shine into the control room.

2 And that was found also to have no impact 3 on total dose, or very small.

4 DR. SCHULTZ: Is that because of the 5 shielding that you have around the control room?

6 (Simultaneous Speaking.)

7 MR. INFANGER: Yes, the wall's very thick, 8 a lot of concrete between the reactor building and the 9 control room.

10 DR. SCHULTZ: And the control room dose is 11 the higher evaluation that you've got in the dose 12 evaluation.

13 What did you assume for in-leakage to the 14 control room, either from egress or from unfiltered 15 in-leaking to the control room?

16 MR. INFANGER: I believe it was 10 cfm.

17 I'll have to check on that. Does anyone back in 18 Corvallis have the in-leakage used in the control room 19 dose?

20 I think that's actually in Chapter 6.4 but 21 Corvallis, are you on the line? I don't know if we 22 have anybody. There's another meeting right now, 23 unfortunately. Our rad protection guys are at another 24 meeting --

25 MEMBER CORRADINI: You can come back to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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186 1 us.

2 MR. INFANGER: We'll get back to you.

3 MEMBER CORRADINI: I would appreciate 4 that.

5 PARTICIPANT: Paul, we can have that in a 6 minute.

7 MR. INFANGER: Thank you. Okay, and of 8 course, the conclusion is that the doses are 9 acceptable for all the events. We used fairly 10 standard industry computer codes to calculate the dose 11 consequences.

12 We used SCALE, TRITON, and ORIGEN for the 13 types and quantities of radioactive isotopes. We used 14 NRELAP to define the thermal hydraulic conditions for 15 the events and the steam line break.

16 For beyond-design-basis events, we used 17 MELCOR to calculate a core damage event. The one 18 thing they did a little different than in other sites 19 is we used ARCON96 to calculate the dose for the EPZ 20 and site boundary.

21 And we did that because the site is so 22 much smaller than other sites, so the industry 23 standard code is put on and that works well for longer 24 distances.

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187 1 distances. And so since our site is so small, ARCON96 2 is used frequently for control room dose and we use it 3 also for our safe boundary dose.

4 We used RADTRAD to calculate the 5 radionuclide transports. STARNAUA was used for 6 aerosol removal in containment.

7 We also used a NuScale-specific code, pHT, 8 to calculate the containment pH, which is important 9 for iodine, the evolution. And then we used MCNP for 10 valuing the shine potential.

11 CHAIR MARCH-LEUBA: You said containment 12 pH?

13 MR. INFANGER: Yes.

14 CHAIR MARCH-LEUBA: What is that a 15 function of?

16 MR. INFANGER: Just the chemistry of the 17 RCS, boron and --

18 CHAIR MARCH-LEUBA: I would assume that 19 you have that -- were controlling the RCS before you 20 start so how does it evolve? Unless you start to make 21 a lot of hydrogen and things like that.

22 Never mind, I'm just curious.

23 MR. INFANGER: This is about NuScale pHT?

24 CHAIR MARCH-LEUBA: Yes, why is the pH 25 varying during the event? I'm just curious.

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188 1 MEMBER CORRADINI: He'll satisfy your 2 curiosity later.

3 MR. INFANGER: Yes, we'll skip that and 4 return to it shortly.

5 CHAIR MARCH-LEUBA: Don't change the 6 slides, I still have another question. Another 7 curiosity, in SCALE 6.1 TRITON, your design cross-8 sections are CASMO-5.

9 So, did you actually run a whole depletion 10 calculation for your fuel with TRITON? I mean, it's 11 a lot of work when you already have it done with 12 CASMO.

13 Is that because ORIGEN is incompatible 14 with CASMO cross-section? Are you going to convert 15 them, or you don't know?

16 MR. INFANGER: No, I'm not aware.

17 CHAIR MARCH-LEUBA: I know people are not 18 using TRITON anymore because it's a big amount of 19 work. You need the whole depletion for so many fuel 20 segments.

21 So, if you truly did that, you did a lot 22 of work that you didn't need to do because you already 23 have the CASMO cross-sections. So, you can find out?

24 MR. INFANGER: Yes.

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189 1 activity.

2 MR. INFANGER: Okay, the accident source 3 term topical report, the number is listed there. It 4 talks about the various sample calculations for each 5 of the events.

6 The FSAR has the actual events that we 7 analyze for DCA, with our sample calculations in the 8 topical report. As far as using Reg Guide 1.183, we 9 deviated from it and in most cases it was for Appendix 10 C and D are related to BWRs and Appendix G is locked 11 rotor.

12 Since we don't have reactor coolant pumps 13 there is no such event. So, we didn't use those 14 sections because they're not applicable to our design.

15 We used Reg Guide 1.183's iodine spiking assumptions 16 and decontamination factors for fuel-handling 17 accidents.

18 We only credited it 23 feet of water even 19 though the pool was much deeper than that. We did 20 iodine removal in the secondary piping or the 21 condenser.

22 Thermal hydraulic response due to the rod 23 ejection accident shows that there was no fuel damage 24 so we did not calculate a dose from the fuel-handling 25 accident. It would just be handled just with the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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190 1 generic evaluation of RCS --

2 DR. SCHULTZ: Bounded by the other events.

3 MR. INFANGER: We used ARCON96 atmospheric 4 dispersion methodology for the short distances to the 5 EAB and LPZ. And we used RADTRAD and modeling 6 techniques consistent with the Reg Guide.

7 So, again, the AST topical accident source 8 term topical report has the methodology. There was an 9 NEI position paper for small modular reactors, 10 investigating some of the uniqueness of the small 11 cores and information on that related to that.

12 I'll use B, spectrum of accidents from 13 MELCOR, surrogate accident scenarios. So, we used 14 MELCOR to simulate the core damage event and that was 15 used a lot also for the PRA.

16 ARCON96 dispersion methodology was used, 17 RADTRAD modeling techniques. And they're now used 18 for the aerosols and pHT.

19 We showed this slide before and this is a 20 summary of the doses. And if you look at the iodine 21 spike design basis source term, the offsite dose is 22 very, very low, less than 0.01.

23 The core damage event, again, is a beyond 24 design basis event since none of our accidents 25 involved core damage. But even those are well within NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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191 1 the limits.

2 CHAIR MARCH-LEUBA: Paul, may I ask if the 3 CR means control room?

4 MR. INFANGER: Yes.

5 CHAIR MARCH-LEUBA: So, this calculation, 6 does it assume the compressed air system doesn't work?

7 MR. INFANGER: No, this assumes the 8 compressed air system functions.

9 CHAIR MARCH-LEUBA: And how does the 10 radioactivity get into the control room?

11 MR. INFANGER: In-leakage.

12 CHAIR MARCH-LEUBA: Even though it's high 13 pressure?

14 MEMBER CORRADINI: That's what I assumed.

15 CHAIR MARCH-LEUBA: How does it physically 16 happen?

17 MEMBER CORRADINI: I don't want to answer.

18 I guessed the answer but I don't know.

19 MR. INFANGER: We'll get the number but I 20 think there's a small amount of in-leakage from 21 ingress and egress.

22 CHAIR MARCH-LEUBA: Every time you open 23 the door?

24 MR. INFANGER: Yes.

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192 1 of how many times you open the window?

2 MEMBER CORRADINI: It's an assumed number.

3 MR. INFANGER: It's an assumed number.

4 CHAIR MARCH-LEUBA: So, it has a large 5 uncertainty, very large uncertainty?

6 DR. SCHULTZ: That's included in that. It 7 also depends on whether you're filtering intake air.

8 MR. INFANGER: Initially, we use air 9 bottles to pressurize the control room and after 72 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> we would go to filtered air.

11 DR. SCHULTZ: And then you have to take 12 consideration of the filters?

13 MR. INFANGER: Right, and the dose 14 analysis is done for 30 days for the control room.

15 CHAIR MARCH-LEUBA: So this is for 30 16 days?

17 MR. INFANGER: 30 days.

18 CHAIR MARCH-LEUBA: Okay, so the --

19 (Simultaneous Speaking.)

20 MR. INFANGER: The bottles run out after 21 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

22 CHAIR MARCH-LEUBA: Okay, so this is the 23 effectiveness of your HEPA filters?

24 MR. INFANGER: Yes.

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193 1 Kenny Anderson from Corvallis to answer a couple of 2 these questions, specifically TRITON.

3 MEMBER CORRADINI: Is he going to be 4 online or are you going to tell us?

5 MR. PRESSON: Online.

6 MR. ANDERSON: This is Kenny.

7 MEMBER CORRADINI: Go ahead.

8 MR. ANDERSON: This is Kenny Anderson with 9 Corvallis.

10 In regards to the question what is TRITON 11 used for, it calculates the microscopic cross-section, 12 that's been said, it calculates macroscopic cross-13 sections similar to the ORIGEN tool and the SCALE 14 package.

15 And then that sets the initial inventory 16 for the core and then that inventory is used with the 17 right processing to get the initial inventory into the 18 RADTRAD models or the other models to eventually 19 calculate the dose or the pH.

20 CHAIR MARCH-LEUBA: My comment was that 21 you already had validated design cross-section sets 22 generated by CASMO-5. And by running TRITON you're 23 reproducing the same numbers that you already have.

24 And my experience, because I used to sit 25 across from the guy that used to do the work, it's a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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194 1 lot of work to do a cycle depletion with TRITON. So, 2 when did you do the work and who did it?

3 MR. ANDERSON: Each time we do an FSAR 4 revision we redo it.

5 CHAIR MARCH-LEUBA: So you do it in house?

6 MR. ANDERSON: Yes, we commercially grade 7 dedicated the components of the SCALE package that we 8 use and I agree, you could use CASMO-5 to simulate the 9 calculation but there are pros and cons associated 10 with that.

11 We thought SCALE is the right tool for the 12 right job.

13 CHAIR MARCH-LEUBA: Well, it gives you a 14 direct plug-in to ORIGEN so I understand it. But my 15 comment was I was surprised because it's a lot of 16 work. But if you know what you're talking about, I'm 17 happy.

18 Thank you very much.

19 MR. INFANGER: And Ken, could you talk a 20 little bit about what we use the pHT program for?

21 MR. ANDERSON: Yes, that takes in the 22 total acids and bases and shows generally a basic 23 solution which as long as you're above a certain 24 threshold, you can prove that there will be no iodine 25 re-evolution.

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195 1 The iodine that's in the water won't come 2 from out of solution into the air space, which would 3 then be eligible to be released. If there was iodine, 4 we have to assume you'd have higher doses.

5 CHAIR MARCH-LEUBA: Yes, I understand that 6 but what would change the pH inside the vessel after 7 an accident? What are the input parameters you put to 8 give to the code?

9 MR. ANDERSON: The initial acid that is 10 released from the fuel and the fuel's damage, that's 11 postulated and then as radiolysis occurs, that's going 12 to be an acid contribution.

13 And then you take credit for the bases 14 that are appropriate and you do the chemistry balance 15 and get the final pH.

16 CHAIR MARCH-LEUBA: Okay, so mostly it 17 releases from the fuel, which are chemical components, 18 and radiolysis. Thank you very much.

19 DR. SCHULTZ: Kenny, we talked about this 20 a little earlier, the shine dose, the way it's 21 presented in a couple of documents is there just isn't 22 any shine dose.

23 And I'm surprised with regards to the LOCA 24 event and the fuel-handling accident that there's no 25 shine dose to the control room. Can you describe why NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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196 1 that would be?

2 Is there just sufficient shielding so that 3 you don't calculate a shine dose?

4 MR. ANDERSON: We do have generic shine 5 doses. We take the worst event and we calculate shine 6 doses to the control room.

7 There's a few different mechanisms where 8 the shine can reach the control room, and then we just 9 apply that to each event. So, that's real 10 conservative.

11 DR. SCHULTZ: Okay, that's more of what I 12 would have expected. So, there is some but it doesn't 13 increase the calculated dose from the other components 14 of release to an extent that it really makes much 15 difference.

16 Is that what you're saying? That's why 17 you use a generic value or a maximum value?

18 MR. ANDERSON: Yes.

19 DR. SCHULTZ: Okay, thank you.

20 MR. PRESSON: And Table 15.0-15 contains 21 some of the information on in-leakage. It looks like 22 around 150 cfm.

23 CHAIR MARCH-LEUBA: I misunderstood. I 24 thought that was not for 30 days but for 3 days.

25 MR. INFANGER: Okay, and with that, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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197 1 next section is on the containment response analysis.

2 Megan's going to address that.

3 MS. McCLOSKEY: Okay, so as was discussed 4 yesterday afternoon in the overview of Chapter 6, the 5 containment is designed to withstand the full spectrum 6 of primary and secondary system, mass and energy 7 releases, accounting for the worst-case single failure 8 and considering loss of power scenarios.

9 And the NuScale methodology for analyzing 10 the containment response is based on NRELAP5 and it's 11 described in our technical report. We are --

12 CHAIR MARCH-LEUBA: For those of us that 13 memorize numbers, is that the LOCA report?

14 MS. McCLOSKEY: No, that's the containment 15 response technical report.

16 CHAIR MARCH-LEUBA: It's a different, 17 third report?

18 MS. McCLOSKEY: Yes.

19 CHAIR MARCH-LEUBA: Thank you. Have we 20 seen that one?

21 MEMBER CORRADINI: It's with Chapter 6.

22 MS. McCLOSKEY: It's with Chapter 6, it's 23 referenced in Chapter 6.

24 MEMBER CORRADINI: The results are in 25 Chapter 6.

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198 1 MS. McCLOSKEY: The results are summarized 2 in a table in Chapter 6.

3 MEMBER CORRADINI: Right.

4 MS. McCLOSKEY: And then they're presented 5 and discussed in more detail along with the 6 methodology for biasing the initial in-boundary 7 conditions in the technical report.

8 And so due to the design of the module 9 with the small high-pressure containment and the ECCS 10 valve opening, we are using NRELAP5 for calculating 11 the mass and energy releases from the RPV and the 12 containment pressurization response as an integrated 13 model.

14 The limiting event scenarios that are 15 addressed, on the primary side we are examining LOCA 16 pipe-breaks in the CVCS discharge and injection lines 17 and the pressurizer high-point vent line. So, two 18 liquid space cases and then the vapor space break.

19 We also are considering the valve opening 20 events for an inadvertent recirc valve opening or vent 21 valve opening.

22 On the secondary side, we look at the main 23 steam line breaks and the feedwater line breaks in 24 site containment.

25 The qualification of the NRELAP5 code for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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199 1 these analyses is based on the qualification and the 2 plant modeling approach that's described in the LOCA 3 and the non-LOCA topical reports for the primary and 4 the secondary-side events respectively.

5 And this report was used because the 6 containment pressure response was considered as a 7 figure of merit from the PIRT development in both of 8 these Ems.

9 And in particular, since the primary-side 10 release events are limiting, the LOCA EM identified 11 high-ranked phenomena that are important for 12 predicting the containment response.

13 So, again, the response is based on the 14 models from those -- developed from the methodologies 15 from those topical reports.

16 But we are biasing the initial and the 17 boundary conditions used in the models in order to 18 bias the mass and energy release and maximize the 19 containment pressure and temperature response as 20 opposed to addressing the critical heat flux or level 21 acceptance criteria that are addressed in the LOCA 22 topical report.

23 For the event analyses, we're applying the 24 maximum break sizes or valve sizes in order to 25 maximize the mass and energy release to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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200 1 containment.

2 In the limiting-condition cases, the peak 3 pressure occurs after ECCS valve opening and so, as 4 was alluded to earlier, the peak pressure is a little 5 bit of a balance between how much volume you have 6 remaining in containment versus the energy still in 7 the RCS at the time the ECCS valves open.

8 This slide summarizes the results of the 9 containment analysis result. In the black or the 10 orange font are the limiting case for each of the 11 break or valve opening cases considered.

12 Grey are the base cases from the 13 containment pressure analysis. So, the difference 14 between the two considers effects of single failures, 15 primarily the effects of single failures and power 16 availability.

17 The initial conditions are biased 18 consistently and then the containment acceptance 19 criteria are shown at the bottom.

20 Our limiting event is the inadvertent RRV 21 opening event with a maximum pressure of 986 psia.

22 Compared to the base case, there's about a 45 PSI 23 delta to the base case in that scenario.

24 That's our largest liquid space discharge 25 event so you get an ECCS valve opening relatively NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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201 1 early in the transient.

2 And that can be distinguished against the 3 RCS discharge line break scenario that's up at the top 4 here, where the base case assumes all power is 5 available so ECCS actuates on high containment level.

6 And therefore, it's more than 1000 seconds 7 into the event before the ECCS valves open. And that 8 allows some cooling through the containment wall 9 before the valves open and the peak pressure occurs.

10 CHAIR MARCH-LEUBA: Remind me, what's the 11 limit for the CMV pressure?

12 MS. McCLOSKEY: It's 1050 PSI here.

13 CHAIR MARCH-LEUBA: That's what I thought.

14 So this is close.

15 MEMBER CORRADINI: There was a change to 16 that. You went from some value to 1050. Has that 17 been reviewed and accepted by Staff?

18 MS. McCLOSKEY: We went from 1000 PSI to 19 1050 PSI and that was submitted to the Staff towards 20 the end of last year.

21 MEMBER CORRADINI: So, that's under review 22 and has been accepted?

23 MS. McCLOSKEY: Sorry?

24 MEMBER CORRADINI: It's been reviewed and 25 accepted?

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202 1 MS. McCLOSKEY: Yes, for the SER.

2 CHAIR MARCH-LEUBA: There are calculations 3 that you satisfy the ASME code at the extra 50 PSI, 4 correct?

5 MS. McCLOSKEY: Mm-hmm.

6 CHAIR MARCH-LEUBA: And you will have to 7 do the hydraulic pressure and everything at this 8 pressure? All the testing?

9 MS. McCLOSKEY: Yes.

10 CHAIR MARCH-LEUBA: You didn't have to 11 make a change on this side? You're convinced that you 12 have sufficient margin?

13 MS. McCLOSKEY: No, it was an analysis 14 change.

15 CHAIR MARCH-LEUBA: I'm always worried 16 when I see a result that is that close to the limit 17 and every single plant has it on other pressure 18 events. So, I will wait until two slides from now to 19 ask you the question.

20 MS. McCLOSKEY: Okay.

21 MEMBER CORRADINI: Can I ask a different 22 question? I know we're now creeping into another time 23 window, but the peak pressure at 986 versus 941, are 24 the conservatisms that were assumed to go to limiting 25 case, is there one particular one or are there a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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203 1 series of --

2 MS. McCLOSKEY: The primary change there 3 is the loss of power assumption, which affects the 4 timing of the ECCS valve opening.

5 MEMBER CORRADINI: Okay.

6 MS. McCLOSKEY: And then the limiting wall 7 temperature case is the injection line break case that 8 is a break from the hot leg condition.

9 The next several slides have results of 10 the limiting pressure case for the inadvertent opening 11 of the RRV. We assume loss of AC and EDSS DC power at 12 the event initiation.

13 We assume a single failure of the 14 remaining RRV to open, which forces all of the -- when 15 the ECCS valves open, it all has to vent through one 16 RRV or the vent valves, which maximizes the energy 17 release to containment with the ECCS valve opening.

18 The low bias IAB opening is assumed. We 19 accounted for a fast release of non-condensable gas 20 into containment to account for non-condensable gas 21 that might be present as dissolved in the RCS fluid or 22 in the pressurizer.

23 Per the methodology, there's no credit for 24 DHRS operation so with the loss of DC power you get a 25 relatively early opening of the ECCS valves about a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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204 1 minute after the event.

2 And then the peak containment pressure 3 occurs shortly thereafter, but it's rapidly decreased 4 to below 50 percent.

5 CHAIR MARCH-LEUBA: Here is where I want 6 to ask the questions.

7 MS. McCLOSKEY: Okay.

8 CHAIR MARCH-LEUBA: If you look at the top 9 left, which is the pressure, and you concentrate on 10 the high times, you see an ISDK which, to me, that's 11 the condensation of the steam on the vessel wall, 12 containment wall.

13 Is that your understanding too?

14 MS. McCLOSKEY: Yes, and what you see in 15 the plot on the lower right are several of the energy 16 balance terms and it's a little hard to read in here.

17 But I want to clarify here that the 18 containment heat removal shown on this plot is the 19 heat removal to the reactor pool, and that's this red 20 line here. That's when the energy gets all the way 21 out to the reactor pool.

22 The total energy transfer from the break 23 from the ECCS valves is this pink line that's coming 24 down here. So, it does take 200 seconds before heat 25 removal to the reactor pool is established, and it's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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205 1 well after the peak occurs.

2 That peak is turned around by the 3 condensation on the inside of the wall and the 4 absorption into the metal.

5 CHAIR MARCH-LEUBA: And there is on the 6 top left a very nice inflection point at 100 seconds.

7 MS. McCLOSKEY: That's the ECCS valve 8 opening.

9 CHAIR MARCH-LEUBA: And there is another 10 inflection point around 400 PSI. My suspicion is the 11 first decreased rate is one ECCS valve dumping steam, 12 then right where you have your mouse, all the ECCS 13 valves start dumping.

14 And at that point you've run out of 15 inventory. You're dump everything you can dump and 16 now you start cooling.

17 MS. McCLOSKEY: You dump everything you 18 can dump and you can see the RCS pressure comes down 19 accordingly.

20 CHAIR MARCH-LEUBA: There's only steam 21 inside the vessel.

22 MS. McCLOSKEY: Yes. So, we don't lose 23 all of the inventory from the RCS, there's a 24 significant amount of inventory that remains --

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206 1 off.

2 MS. McCLOSKEY: It flashes, it becomes 3 saturated, and you establish recirculation through the 4 recirculation valves.

5 CHAIR MARCH-LEUBA: Roughly at 200 6 seconds, if you look at the bottom left, you have a 7 net, not loss. And what comes in equals what comes 8 out because this is the sum of the break plus the --

9 (Simultaneous Speaking.)

10 Now, if you had had higher inventory in 11 the pressurizer, would that peak inflection point on 12 the top one, would it be a little higher? If you had 13 more inventory to lose?

14 MS. McCLOSKEY: We've biased the inventory 15 high to the high-end of the normal operating --

16 CHAIR MARCH-LEUBA: This is already --

17 (Simultaneous Speaking.)

18 MS. McCLOSKEY: Yes.

19 CHAIR MARCH-LEUBA: This is fully closed.

20 Okay, good enough.

21 MS. McCLOSKEY: I think that covers the 22 points on that slide.

23 MEMBER CORRADINI: Maybe I don't remember.

24 The one thing was the biased high pressurizer level.

25 Are you starting with a vacuum?

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207 1 MS. McCLOSKEY: Yes, at three psia --

2 (Simultaneous Speaking.)

3 MEMBER CORRADINI: There was a discussion 4 between you and the Staff about 65 lost pounds of air.

5 I want to understand where that was. There were non-6 condensables that were asked about, do you know what 7 I'm talking about?

8 MS. McCLOSKEY: Yes, we account for the 9 non-condensables both initially present in containment 10 to give you that 3 psia and that are maybe present in 11 the pressurizer vapor space or dissolved in the RCS.

12 MEMBER CORRADINI: So, is this a corrected 13 calculation with the additional 65 pounds?

14 MS. McCLOSKEY: Yes.

15 MEMBER CORRADINI: And the source of the 16 65 pounds was internal to the RCS or external inside 17 the containment? That's what I couldn't understand.

18 MS. McCLOSKEY: Internal to the RCS.

19 MEMBER CORRADINI: And what came out, a 20 solution?

21 MS. McCLOSKEY: Yes.

22 PARTICIPANT: And what's in the 23 pressurizer?

24 MS. McCLOSKEY: Yes.

25 MEMBER CORRADINI: Okay, thank you.

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208 1 CHAIR MARCH-LEUBA: Because if it was in 2 containment it wouldn't be operating. It would be --

3 MEMBER CORRADINI: 65 pounds is not a lot.

4 I'm not sure, one PSI is how many pounds in your 5 containment? I was going to calculate that but --

6 (Simultaneous speaking.)

7 MS. McCLOSKEY: The 3 PSI is on the order 8 of 60 to 65 pounds. So, it happens to be about equal 9 to what we get from the RCS as well.

10 CHAIR MARCH-LEUBA: Well, my brain is 11 telling me is that somebody designed the size of the 12 containment five years ago and we are really lucky 13 because he was right in the know.

14 Or we've been losing margins since then.

15 I'm sure you don't design your containment that close 16 to limits. Probably you'll be losing margin.

17 MEMBER CORRADINI: I don't know, there's 18 an awful lot of dries that are awful close.

19 MS. McCLOSKEY: And I've got two slides 20 where we look at some of the other margins that aren't 21 accounted for in that 986 number.

22 CHAIR MARCH-LEUBA: This is not unusual.

23 You look at all the hours of pressure and every single 24 BWR, they have less margin than you do.

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209 1 sharpen the pencil and do a new calculation which is 2 more accurate until he goes under.

3 MS. McCLOSKEY: The containment is very 4 effective at removing energy after the initial blow-5 down.

6 So, in this case, by about half an hour 7 into the event, we've reduced the pressure to below 50 8 percent of the design limit that's well within the 24 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> expected to support the radiological analyses.

10 And in terms of the long-term cooling, 11 we'd refer to the long-term cooling technical report 12 to demonstrate continued effective decay and residual 13 heat removal.

14 And finally, in terms of the margin 15 assessments, the maximum pressure has less than ten 16 percent margin to the acceptance criteria, which is 17 guidance from the DSRS.

18 Additional factors that are not accounted 19 for in that 986 number that would provide additional 20 margin are both accounting for the external pressures 21 in the assessment and crediting some degree of heat 22 removal from the DHRS.

23 This maximum pressure is taken at the 24 bottom of the containment and it's conservatively 25 evaluated assuming an external pressure of 0 psia that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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210 1 doesn't account for atmospheric pressure or the pool 2 hydrostatic head.

3 It would provide about 22 PSI additional 4 margin. The evaluation model also doesn't credit the 5 single- failure-proof safety-related decay heat 6 removal system that would be actuated.

7 We have some sensitivity calculations 8 indicating about 37 PSI additional margin could be 9 obtained but we've not pursued that due to the 10 additional validations of NRELAP5 that would be 11 required under these scenarios.

12 So, overall, our conclusion is that our 13 analysis provides assurance that we provide sufficient 14 margins to satisfy the requirements of GDC16 and 50.

15 MEMBER CORRADINI: Can I suggest a break?

16 MS. McCLOSKEY: Yes.

17 MEMBER CORRADINI: Please? Can we take a 18 break until about 3:35 p.m.?

19 (Whereupon, the above-entitled matter 20 went off the record at 3:23 p.m. and 21 resumed at 3:35 p.m.)

22 MEMBER CORRADINI: So let's get back into 23 session. So, Ben, we'll turn it back to you.

24 MR. BRISTOL: Okay, thank you. So the 25 next topic we're going to cover here is long-term NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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211 1 cooling. And just to sort of set the expectation, 2 what we're trying to cover here is the scope that was 3 defined in the long-term coolant technical report. So 4 this isn't a full, comprehensive list of all of the 5 issues related to the extended ECCS operation or even 6 the HR operation, but we'll get into those a little 7 bit later. I think we're prepared to address most of 8 them.

9 So as I mentioned, specifically we're 10 looking at demonstrating that post-LOCA ECCS kind of 11 performance acceptance criteria maintaining coolable 12 geometry and demonstrating ample cooling via ECCS. In 13 addition, a couple other DSRS-specific considerations.

14 So in terms of acceptance criteria, 15 there's primarily two that we're looking at in term s 16 of this analysis scope, one being that core cooling is 17 maintained and that is evaluated via acceptance 18 criteria of collapsed liquid level, keeping the core 19 covered.

20 In addition, cladding temperature is 21 evaluated as part of the analysis. And then 22 specifically, something a little different is that 23 coolable geometry is maintained, and that's evaluated 24 through demonstrating boron precipitation limits are 25 not exceeded.

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212 1 So in terms of the technical reports, 2 RELAP is used to evaluate the integral response to 3 decay heat removal paths from ECCS and containment 4 heat transfer to the pool.

5 A point of, again, scope: recriticality 6 is not analyzed as part of this analysis, so design 7 basis decay heat or the design base is shut down 8 condition is assumed, that overcooling analysis we'll 9 actually get into in the next presentation. It's 10 otherwise presented in the SR.

11 CHAIR MARCH-LEUBA: So you mean long 12 cooling under decay heat conditions?

13 MR. BRISTOL: Under decay heat conditions, 14 that's right.

15 So in terms of development of the 16 evaluation approach, PIRT was used, addressing ECCS 17 cooling conditions and within evaluation of long-term 18 cooling PIRT phenomena. The LOCA EM is used as a sort 19 of initial validation source. Much of the phenomena 20 is addressed as part of that analysis.

21 In other places, bounding analytical 22 techniques via inputs or methodology are assumed.

23 We'll address a couple of those.

24 In addition, there were a couple of 25 assessments that were specifically set up using the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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213 1 NIST facility --

2 MEMBER CORRADINI: We'll talk about those 3 in closed session. This kind of goes back to my 4 question about something out of 15.02, and RELAP's 5 ability to model these tests versus what I'd see at 6 full scale.

7 So are we going to discuss these -- we're 8 not going to discuss these now, I would assume.

9 MS. MCCLOSKEY: No, we're not going to 10 discuss them now, and --

11 MEMBER CORRADINI: Otherwise I'll just 12 wait and ask the staff tomorrow. I want to get to --

13 because staff had some comments about RELAP's ability 14 to calculate what they saw in these relative to 15 consistency in calculation for the full scale, and I 16 want to address it. I just don't know where to do it.

17 MS. MCCLOSKEY: I think we'll hear the 18 questions in closed session, and then --

19 MEMBER CORRADINI: Okay. Fine.

20 MS. MCCLOSKEY: -- try to get that one 21 too.

22 MEMBER CORRADINI: All right.

23 MS. MCCLOSKEY: That's the way to respond 24 to this.

25 MEMBER CORRADINI: Okay.

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214 1 MR. BRISTOL: So just finally in terms of 2 qualification, the qualification inclusions from LOCA 3 EM are used for evaluating the LOCA-type initiating 4 events, and then the non-LOCA EM is leverage for the 5 non-LOCA DHRS events that then transition.

6 All of the event types are considered as 7 part of the LTC technical report. All of the events 8 that end up in ECCS mode in terms of what's presented 9 in the technical report.

10 Okay. So just a couple of details on the 11 precipitation analysis. It's a simple volume mixing 12 approach. There's no time dependence; what we mean by 13 that is that the boron redistributes as part of ECCS 14 actuation.

15 That way, that redistribution and any of 16 the boron that goes into containment provides would be 17 non-conservative, so it's assumed that essentially the 18 thermohydraulic transient is evaluated assuming that 19 the core and lower riser region contain all of the 20 initial boron mass from the start of the event.

21 In fact, the actual mass that's used is 22 the bounding sort of startup condition. So for 23 conservatism, the precipitation temperatures are 24 evaluated against the maximum allowable boron mass 25 with no credit for any of the redistribution NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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215 1 mechanisms or potential loss mechanisms.

2 So in terms of the analysis scope, we look 3 at a spectrum of the LOCA Break spectrum, including, 4 in addition to the inadvertent opening of the RRV or 5 RVV. Those are evaluated out through the longer term.

6 As part of the way the analyses are 7 performed, there's a simplified model; I'm going to 8 get into that in the next slide, but the calculations 9 are set up from the start of the event, and then to 10 ensure that the boundary conditions are effectively 11 transferred.

12 So instead of transferring them, we 13 actually do comparisons of the full EM models versus 14 the simplified model to demonstrate that we're 15 capturing the transient correctly such that the point 16 that we distinguish the onset of long-term cooling is 17 starting from the correct initial conditions.

18 As part of the transient scope, we look 19 for -- it's about a 12-and-a-half-hour period that the 20 transient calculations are performed. Beyond that, a 21 state point method is used, and we'll discuss the 22 basis for that.

23 Simply put, the transient phase or 24 transient progression has recovered within the 12-hour 25 window such that what we're looking for beyond the 12-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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216 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time frame is simply just limiting temperature 2 conditions. That's where the state point analysis can 3 give us a conservative result.

4 MEMBER CORRADINI: So maybe this is 5 getting ahead, but do these include the results in 6 terms of your response in RA8930, or is this prior to 7 this?

8 MR. BRISTOL: No. We'll get into the 8930 9 topic in the closed session.

10 MEMBER CORRADINI: Okay. Fine.

11 MR. BRISTOL: So we essentially have three 12 cases that we're kind of looking for in terms of the 13 way the models are biased. The biases for minimum 14 level are a little about different than minimum 15 temperature. And then a more traditional sort of ECCS 16 performance analysis, looking at maximum temperature 17 is also evaluated.

18 It turns out maximum temperature cases are 19 non-limiting or minimum cool-down cases are non-20 limiting for either the precipitation analysis or the 21 minimum-level analysis, and that has to do with the 22 vapor pressures that are reached and their effect on 23 ECCS performance.

24 So just a summary of the results here:

25 minimum level analysis confirms that the core remains NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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217 1 covered, and the minimum temperature analysis confirms 2 that precipitation limits are not reached.

3 The maximum temperature analysis just 4 shows that we have a trend that's successful from a 5 cladding perspective.

6 So what we've got in the figure here is 7 100 percent injection line break as compared to the 8 more limiting LOCA SAL break. The five percent and 9 sort of slower break. We see here that the definition 10 of the onset of long-term cooling for the larger break 11 is after the equilibrium conditions are reached, which 12 is right about here for the larger break.

13 For the smaller break, it's actually this 14 scope all the way up to the -- this dip is due to the 15 ECCS actuation itself. In three of the recovery 16 phase, this is what we consider in the LOCA 17 methodology, and so the long-term cooling then picks 18 up from there.

19 And so from kind of a phenomena 20 perspective, we have two different things going on.

21 There's depressurization-driven ECCS kind of capacity 22 minimum level that we get through here. You see a 23 little bit of it in this response here as well.

24 That's driven by the maximizing 25 containment cooling, which minimizes the vapor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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218 1 pressure and containment and maximizing the stored 2 energy and decay heat in the RCS, which maximizes the 3 pressure in the RCS to maximize the differential 4 pressure across the vent valve. And that's what 5 really drives our long-term cooling minimum levels 6 phenomenon.

7 MEMBER CORRADINI: Say that again slow.

8 MR. BRISTOL: So because it's a manometer 9 problem --

10 MEMBER CORRADINI: Sure.

11 MR. BRISTOL: -- we want to maximize the 12 deep heat in the vent space, and that's going to 13 establish the actual equilibrium condition in the 14 liquid space. Does that make sense?

15 So the deep heat across the vent valve, 16 that's the flow-driven DP. The flow through the reset 17 valve is very, very, very small, very low DP. So head 18 is what's driving flow backward.

19 If my DP across my vent valve, my steam 20 flow rate is too small, I'll continue to accumulate 21 level in the containment, but now allow it to 22 recirculate back into the RPB.

23 So what we want to maximize is the 24 containment heat removal, which draws the vapor 25 pressure down on the containment side, but the vapor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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219 1 generation rate on the RCS side. And that maximizes 2 the flow capacity that's required for the vent valves.

3 And these are all evaluated with one train 4 of ECCS assumed failed, which is really the only way 5 we ever see this thermohydraulic transient response.

6 So in conclusion, I think I covered that 7 in these couple of bullets here. The summary is that 8 once decay heat begins to drop, and it's a combination 9 of decay heat and the stored energy in the system, the 10 level starts to recover the DP, the demand across the 11 vent valves continues to drop, and we see a level 12 recovery. That gives us the confidence that we can 13 kind of switch analytical techniques; we don't need to 14 continue running RELAP through these really kind of 15 slow transient progressions or calculations.

16 So here's just a couple more figures 17 describing the difference in the two events. Unless 18 there's specific questions, I don't need to cover them 19 in much more detail.

20 Okay. Minimum temperature case: so this 21 would be a case where all of ECCS is assumed to 22 function. I don't have the level figure in this 23 particular case, but it's one the quickly reaches an 24 equilibrium condition. So minimizing the temperature 25 just by itself, including 100 percent actuation if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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220 1 ECCS doesn't drive that same level of response, which 2 is why we distinguished those two cases.

3 So in this case, we're primarily looking 4 at ensuring that as temperature continues to drop, we 5 don't reach the precipitation limits for boron.

6 MEMBER CORRADINI: Remind -- you're going 7 to go to the next slide?

8 MR. BRISTOL: Yes.

9 MEMBER CORRADINI: So again, is the red 10 line five percent break?

11 MR. BRISTOL: Yes.

12 MEMBER CORRADINI: So what is going on 13 that I almost come up with the same behavior? I guess 14 I'm a little confused. I'm sorry. I'm looking at the 15 water level again.

16 MR. BRISTOL: Okay. So in this case, I 17 believe it's an IAB blocked-out case, and so what we 18 have is, without DHR cooling available, so we have a 19 pressurized RCS with a very small --

20 MEMBER CORRADINI: So both trains of DHRS 21 are blocked?

22 MR. BRISTOL: Yes, they're not credited.

23 That's correct. So it's a very slow system 24 depressurization, because they're mostly losing liquid 25 inventory out the break, so that's a pretty low-energy NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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221 1 release.

2 Eventually this level begins to accumulate 3 in containment. I get more and more energy released 4 through the RPV wall itself. Eventually, I 5 depressurize to the IAB release pressure window, at 6 which point ECCS actuates and then I continue the 7 transient progression.

8 So this slide we're just looking at the 9 maximum temperature case for both the two break 10 scenarios. Again, this case would be where the pool 11 temperature is biased to sort of the high condition, 12 assuming we're practically at pool boiling conditions 13 to ensure that even if the pool boiled and we 14 initiated transient from there, we still would provide 15 adequate ECCS performance.

16 So that's one of the cases where we sort 17 of take an analytical conservatism that's outside the 18 initial conditions of tech specs per se, in order to 19 just sort of deterministically addressing pool heat up 20 effects.

21 None of the long-term cooling specifically 22 address transient pool temperature, but we do address 23 it from a minimum condition in terms of the constant 24 minimum or a constant maximum.

25 So long-term cooling conclusions: maximum NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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222 1 temperature cases show adequate core cooling. Minimum 2 temperature cases show margin to precipitation, and 3 the minimum level cases show the core stays covered.

4 Of note, the limiting 100 percent 5 injection line case had the minimum level of about 2.8 6 feet above the core. That minimum occurred about 7 three and a half hours.

8 MEMBER SKILLMAN: Ben, are you assuming 9 your minimum temperature for precipitation as 65 10 degrees Fahrenheit? Was that your marker for 11 precipitation?

12 MR. BRISTOL: No. We're using, we're 13 evaluating precipitation, so there's some calculations 14 that are performed within the analysis that actually 15 look at the volume, the mixing volume that's assumed, 16 relative to the temperature in the core as a function 17 of time.

18 MEMBER SKILLMAN: What was the maximum 19 boron concentration you had on that track?

20 MR. BRISTOL: So I don't have --

21 MEMBER SKILLMAN: Yes, it isn't in any of 22 your images here.

23 MR. BRISTOL: I believe it's in -- there's 24 a table of results in the technical report that 25 actually describe the precipitation conditions and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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223 1 way that calculation's performed. I don't have a 2 slide on that, no.

3 MS. MCCLOSKEY: And we look at it in terms 4 of margin to the solubility temperature, since we 5 calculate the solubility temperature as a function of 6 the mixing volume, which is the liquid in the core and 7 riser region.

8 And my recollection from the technical 9 report results is that there's at least 30 degrees 10 Fahrenheit margin to the solubility temperature in all 11 cases.

12 MEMBER SKILLMAN: Thank you.

13 MEMBER CORRADINI: Is this the -- I guess 14 maybe I should have asked this earlier. Is this the 15 two-volume calculation where you've got the riser and 16 the core in one volume, and all what's outside of in 17 containment as the second volume?

18 MR. BRISTOL: No. This is an even more 19 simplified conservative analysis. We'll get into 20 that; I have a presentation --

21 MEMBER CORRADINI: I don't want to ask 22 much more, but I thought it was -- I wanted to ask 23 when it was -- we'll wait to closed session.

24 MR. BRISTOL: Yes. And we can circle back 25 to how it relates to these results --

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224 1 MEMBER CORRADINI: Okay.

2 MR. BRISTOL: -- when we get to that 3 session.

4 MEMBER CORRADINI: Okay. Thank you.

5 MR. BRISTOL: Okay. Are we ready to move 6 on?

7 MEMBER CORRADINI: Can't wait.

8 MR. BRISTOL: Loss of shutdown margin.

9 This is the analysis of our postulated return to power 10 event. A couple of elements of licensing vices, I 11 suppose. The NuSCALE DCA includes an exemption 12 request from GDC 27. I think the ACRS has been 13 briefed on that before.

14 The reason for that: ECCS design does not 15 include boron addition as part of the design, and 16 therefore there's almost no way to meet the as-written 17 definition of GDC 27.

18 As an alternative, NuSCALE has proposed --

19 and this is in review with the staff -- the principal 20 design criteria 27. What we're trying to capture is 21 the intent of the GDC as applied to past designs that 22 wouldn't have necessarily active poison addition 23 capabilities.

24 And so the demonstration of the 25 acceptability of that variant from GDC is highly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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225 1 dependent on showing acceptable results from both an 2 overcooling event with a stuck rod, and the one key, 3 I think, for us in terms of the design criteria is the 4 safety-related reactivity control system is designed 5 and is required to always ensure the core stays sub-6 critical through cold-check conditions without boron 7 addition.

8 So it's all rods in on a reload basis, we 9 will demonstrate that the core can go Re critical and 10 stay Re critical out cold conditions.

11 So compliance immediate shutdown margin is 12 sufficient to turn the events around that we analyzed 13 in Chapter 15 and ensure that the SAFDLs are met.

14 Again, cold shutdown is achieved with all rods 15 inserted, and the loss of shutdown margin consequences 16 are sufficiently benign.

17 We've defined that to be a SAFDLs 18 evaluation of a limiting return to power condition and 19 that the overall heat removal capacities of ECCS and 20 DHRS are not challenged by the event.

21 It turns out that heat removal 22 capabilities above those systems are actually what 23 drive the event.

24 And then on top of that, the overall 25 probability of the combination of events that lead to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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226 1 a loss of shutdown margin is sufficiently small.

2 So just the reactivity kind of physics 3 overview: there's two main drivers, or three, I 4 guess, that one could postulate. Primarily one is 5 moderator cooling that drives an insertion of 6 reactivity as the systems decrease RCS temperature.

7 Under the limiting kind of cold pool conditions with 8 either DHRC or ECCS, the temperature decrease can 9 happen fairly rapidly on the order of a few hours. We 10 just looked at a couple of those figures already.

11 Another primary driver is the immediate or 12 the time-dependent fission product decay. Core 13 poisons, and so xenon being the primary one there. It 14 actually inserts negative reactivity for the first few 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> post-event, and then gradually decays over the 16 course of 12 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, about that time frame.

17 So in that sort of after six to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 18 is really when it starts inserting a fair amount of 19 reactivity.

20 MEMBER CORRADINI: In your analysis, you 21 neglect this, so you conservatively come potentially 22 under some conditions to return to power soon. Am I 23 misunderstanding?

24 MR. BRISTOL: No, that's correct. That's 25 correct.

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227 1 MEMBER CORRADINI: And that was something 2 that was required by staff, or is that something you 3 chose to --

4 MR. BRISTOL: No, that was a conservatism 5 that we discretionarily applied.

6 MEMBER CORRADINI: Okay. Have you done a 7 calculation to see how much later you actually then 8 delay return to power?

9 MR. BRISTOL: Yes. The next couple of 10 slides, I'm going to get into that characterization 11 and circle it back to what we've actually presented in 12 15.6 to try to give that a little more context.

13 One of the other open items, and again, 14 we've got a presentation on this, but I just wanted to 15 address from a return to power perspective is, boron 16 redistribution.

17 Overall in ECCS mode, the core is boiling, 18 so that tends to be a concentrator of boron. The 19 containment is the condenser. The vapor leaves with 20 a very low fraction of the boron, as opposed to the 21 inlet conditions.

22 So we sort of know if the system starts 23 with boron in it, then it's going to tend to 24 accumulate in the core, leading to a lower boron 25 condition probably being more limiting for the return NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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228 1 to power analysis. That's one of the things we're 2 working with the staff in reviewing of.

3 I think for us, the conclusion is, as long 4 as the main loss mechanism of concern would be, could 5 there be some solidification mechanism that then just 6 leads to a loss of soluble boron from the system even 7 after it starts to concentrate, because of the 8 extended period of time that we're talking about.

9 And so that's a big part of the 10 presentation that we'll get into. Our conclusion is 11 that there's not.

12 CHAIR MARCH-LEUBA: The other concern is 13 that you're accumulating lower rate in the 14 containment, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later when you open the 15 ECCS, the low ratio goes into the lower plane and gets 16 into the core.

17 MR. BRISTOL: Sure.

18 CHAIR MARCH-LEUBA: You can get the slag 19 into the water. I just have difficult problem to 20 monitor.

21 MR. BRISTOL: Right, and that's something 22 I think we have addressed with the staff. Just the 23 way that the ECCS actuation occurs and the way the 24 circulation rates are, there's no mechanism to get a 25 large slug actually through the system. So we've got NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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229 1 some analysis of that.

2 Okay. So I think I covered this. Oh, so 3 this is what I was alluding to. We're working through 4 a better characterization, I think, of the transient 5 nature of the event, using the physics a little more.

6 Initially, the OCRP event that we presented as part of 7 the SR was set up to be analyzed a little bit more 8 like kind of a stylistic AOO analysis where we're 9 using core design limits in trying to approximate 10 these defects.

11 So the overall reactivity balance was 12 mischaracterized a fair amount from a tiny 13 perspective. The point really was to show that even 14 through a pretty severe return to power kind of 15 analysis, we were a long way from CHF limits and were 16 well bounded by other event analyses.

17 So the purpose of this work is really a 18 little bit more detailed characterization of the three 19 things I talked about before: specifically that the 20 cooldown rates of both ECCS and DHRS, relative to the 21 xenon-driven reactivity insertion. So I kind of 22 alluded to that before.

23 Over the 12- to 72-hour range, the 24 reactivity insertion rate is really, really slow. So 25 in sort of looking at this, worst rod stuck out Re NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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230 1 criticality is achieved in RCS temperatures below 200 2 degrees, around 200 degrees, such that we used the 3 state point analysis that we developed for the long-4 term cooling approach to then do a state point 5 reactivity calculation --

6 CHAIR MARCH-LEUBA: Using what tool?

7 MR. BRISTOL: SIMULATE-5 for the --

8 CHAIR MARCH-LEUBA: Does the SIMULATE 9 effectively calculate -- calculation?

10 MR. BRISTOL: That's right. So we 11 actually iterated to critical power levels based on 12 thermohydraulic conditions from equilibrium either 13 ECCS or DHR cooling conditions.

14 CHAIR MARCH-LEUBA: That's good, because 15 you're reaching a calculation with a constant MTC as 16 you're going through the whole transient, which is 17 crazy. SIMULATE is when the regulate operations, 18 correct? You get a 200 degrees Fahrenheit if you have 19 depressed rise, which you open to see if you are 20 boiling.

21 MR. BRISTOL: That's right.

22 CHAIR MARCH-LEUBA: So with decay heat, 23 does this calculation come from boiling, or is it not 24 boiling?

25 MR. BRISTOL: It does not. We NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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231 1 conservatively assumed sub-cooled conditions for the 2 boundary conditions to simulate. So that gives us a 3 much more conservative critical power level.

4 The power levels are low enough; I'll get 5 to them my next slide. But I think we've got overall 6 confidence in the characterization of the event, even 7 with the conservativisms that are applied.

8 We have looked at the influence of density 9 with SIMULATE and demonstrated that very little 10 voiding in the core would suppress the power, 11 obviously. And so that's an important conclusion, but 12 there is some concern if the low void fraction rates, 13 where we would justify one, two, three percent boiling 14 rates.

15 And as any ECCS mode, as we get to lower 16 and lower vapor pressure conditions, the level head in 17 the riser starts to drive a saturated kind of 18 temperature curve, and we could postulate that the 19 boiling could, or some of the two-phase exchange heat 20 removal could be a current closer to the level of 21 interface, and with convective mixing driving the 22 cooling of the core region.

23 CHAIR MARCH-LEUBA: Considerable.

24 MR. BRISTOL: Yes. So like I said, we did 25 a steady-state characterization of RCS as a function NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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232 1 of the decay heat for both DHR and ECCS heat removal.

2 There's a big spectrum calculation that was done; that 3 was using a read-out model.

4 And then like I said, the worst rod stuck-5 out critical power level is iterated to using SIMULATE 6 as a function of RCS temperature.

7 Conclusions that a loss of shutdown margin 8 can be achieved on kind of the time frame of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 9 or so with zero boron in the system after xenon's 10 decay. Temperatures are below 200 degrees.

11 ECCS cooling is a little bit more 12 effective, so it drives a little higher temperature 13 conditions. Again, that's with the sub-cooling 14 assumption is part of that.

15 Pool temperatures: so one of the things 16 that I thought was important to characterize is that 17 any pool heat of effect that could occur would 18 actually mitigate the event as the pool gets above 140 19 degrees. It's going to shut -- the RCS temperature 20 just climbs with it, and it's going to shut everything 21 back off.

22 And then simple CHF analysis; I don't have 23 any of the results, but the pool in the CHF analysis 24 demonstrates large margin.

25 MEMBER CORRADINI: What do you use for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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233 1 CHF analysis?

2 MR. BRISTOL: That particular case is 3 Zuber.

4 MEMBER CORRADINI: Foster-Zuber?

5 MR. BRISTOL: Zuber-Griffith.

6 MEMBER CORRADINI: Oh, okay.

7 MR. BRISTOL: So up here we've got the 8 riser uncovered, so one of the things we'll also get 9 to in the closed session, but we've addressed somewhat 10 is riser uncovery nominally adds resistance to the 11 cooling capability. So we looked at a comparison of, 12 if inventory is maintained in the RCS, DHR is going to 13 be more effective than if the riser uncovers, so 14 that's kind of a quantification of the difference 15 there.

16 Down here this is ECCS cooling or kind of 17 the equilibrium temperature as a function of continual 18 decreasing decay heat.

19 MEMBER CORRADINI: I don't understand this 20 graph. I looked at it a few times, and I'm lost. Can 21 you help us, please?

22 MR. BRISTOL: Okay. So each of these 23 points represents a RELAP calculation that we did that 24 set a constant heat input; constant decay heat, 25 constant core power level, using our system model.

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234 1 MEMBER CORRADINI: Oh, so you basically 2 did a series of parametrics.

3 MR. BRISTOL: That's right.

4 MEMBER CORRADINI: Where you set some 5 decay heat level.

6 MR. BRISTOL: Yes. And the equilibrium 7 temperature condition in the core then is a function 8 of the actual decay heat level.

9 MEMBER CORRADINI: Right.

10 MR. BRISTOL: So that's what we're doing 11 there. If we were to assume that there's no residual 12 heat removing, we're at a pure steady-state condition.

13 This would be the temperature solution for that power 14 level. This upper graph, like I said, is with the 15 riser uncovered.

16 CHAIR MARCH-LEUBA: Only DHRS is working, 17 right?

18 MR. BRISTOL: And only DHRS working.

19 CHAIR MARCH-LEUBA: Both of them or only 20 one?

21 MR. BRISTOL: Two.

22 MEMBER CORRADINI: Okay. Which maximizes 23 cooling.

24 MR. BRISTOL: That maximizes the heat 25 removal. This is with the riser covered, so at any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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235 1 given heat level we're going to get quite a bit lower 2 temperature condition, and this is the same 3 calculation performed in ECCS mode.

4 MEMBER CORRADINI: And the black line?

5 MR. BRISTOL: This is the SIMULATE 6 results. So --

7 MEMBER CORRADINI: So you're looking for 8 the intersection of the little dots and the black 9 line?

10 MR. BRISTOL: That's right. So the 11 critical power level in ECCS mode would be evaluated 12 to be about one percent reactor power --

13 CHAIR MARCH-LEUBA: Again, how do you do 14 the SIMULATE?

15 MR. BRISTOL: So the SIMULATE calculation 16 is performed taking the thermohydraulic boundary 17 conditions --

18 CHAIR MARCH-LEUBA: From?

19 MR. BRISTOL: From -- well, it's a 20 spectral analysis as well, so we just define --

21 iterate to a critical power level assuming a core 22 inlet temperature of 100 degrees, 120 degrees, 130 23 degrees, et cetera. And so --

24 CHAIR MARCH-LEUBA: In SIMULATE you have 25 to specify the power and the core inlet temperature NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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236 1 and the flow?

2 MR. BRISTOL: In this case we're iterating 3 to critical power level.

4 CHAIR MARCH-LEUBA: Critical power level?

5 MR. BRISTOL: That's right.

6 CHAIR MARCH-LEUBA: But even -- you have 7 to specify a temperature and a flow?

8 MR. BRISTOL: A temperature and a flow, 9 that's correct. So what we ended up doing is mapping 10 the flow, the DHR flow, correlating to the riser cover 11 condition as the flow condition. What the flow does 12 is essentially just sets the delta T across the core 13 to a given power level.

14 CHAIR MARCH-LEUBA: But what I'm going to 15 is the black line is a function of core power and 16 flow, and you've brought in the function of decay 17 heat, which is power.

18 MEMBER CORRADINI: Yes, I don't think 19 that's what he -- that isn't how I understood the 20 curve. The black line is basically, for a set of 21 thermohydraulic conditions of density and temperature, 22 what's the power I achieve? And he's just simply 23 plotting out the excess as if it were an equivalent 24 decay heat.

25 MR. BRISTOL: That's right.

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237 1 MEMBER CORRADINI: And so he's basically 2 saying that at the worst set of conditions was the --

3 now I think I get it. The ECCS little dots, I have 4 one percent plus 0.35 percent as the total power that 5 I'd be producing. It's the red line plus the black 6 line. Am I understanding --

7 MR. BRISTOL: No, the intersection, we're 8 saying is that --

9 MEMBER CORRADINI: Oh, you're adding the 10 two together?

11 MR. BRISTOL: Yes. That would be the 12 equivalent critical power level for the --

13 MEMBER CORRADINI: But the conservatism 14 that I'm seeing here, unless I misunderstood, is the 15 black line is assuming zero void.

16 MR. BRISTOL: That's correct.

17 MEMBER CORRADINI: So I'm way to the 18 right.

19 MR. BRISTOL: So it's pretty 20 characteristic of the DHR conditions where we're sub-21 cooled. So that's about the real answer for DHR, and 22 again, in ECCS conditions we do see under the extreme 23 cooldown conditions that we can actually move the 24 boiling up out of the core region itself and into the 25 lower riser as a flashing type phenomena with an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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238 1 amount of convective --

2 MEMBER CORRADINI: Okay. Fine. But the 3 black line right now is zero void?

4 MR. BRISTOL: Is zero void.

5 MEMBER CORRADINI: So I think it's like a 6 reactor static calculation. The black line is a 7 reactor static calculation overlaying and intersecting 8 essentially a thermohydraulic parametric on power.

9 MR. BRISTOL: That's correct.

10 MEMBER CORRADINI: I got it, finally.

11 CHAIR MARCH-LEUBA: I ain't got it, but 12 that's okay.

13 MR. BRISTOL: Okay.

14 MEMBER CORRADINI: Just so you guys have 15 seen this before, the shine application for the 16 construction permit when they were doing their annular 17 thing, showed exactly the same sets of calculations 18 about how they would essentially achieve criticality 19 or sub-criticality under the shine medical technology.

20 Same sort of parametric calculation; 21 that's why --

22 CHAIR MARCH-LEUBA: I just don't know 23 where the decay heat power comes into the SIMULATE 24 calculation if you specify what the temperature is.

25 MR. BRISTOL: So decay heat is -- maybe NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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239 1 decay heat is the confusing parameter here. RELAP is 2 just given a point, basically, of a constant power 3 level, and that generates four ECCS conditions at 4 given temperature condition for DHR conditions at a 5 different temperature condition.

6 And then we're recorrelating the SIMULATE 7 results, which we specify flow and temperature, so we 8 run a spectrum of temperatures. We assume just a 9 constant flow for all of these analyses that's on the 10 conservative high end. It correlates to the DHR 11 covered conditions.

12 So for ECCS mode it's really pretty high.

13 The internal flow isn't something we would necessarily 14 --

15 CHAIR MARCH-LEUBA: In a realistic 16 condition the flow would be essentially zero, but you 17 will be uncovered.

18 MR. BRISTOL: Right.

19 CHAIR MARCH-LEUBA: And -- anyway --

20 MR. BRISTOL: Sure.

21 CHAIR MARCH-LEUBA: I think two reasonable 22 people can come up with five different answers.

23 MR. BRISTOL: I suspect we'll see some 24 more tomorrow. So on the topic of different answers, 25 going back to what we've actually got in 15.06; so NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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240 1 again, the purpose is, when we submitted the DCA was 2 to do a pretty bounding analysis.

3 So the way that's performed is, a DHR 4 cooldown is more productively performed. There's a 5 couple of key conservatisms; on is that there's no 6 xenon applied as a poison in terms of the overall 7 reactivity balance. The minimum kind of core design 8 limit shutdown margin is assumed. There's a constant 9 MTC applied, and really all that's doing is setting 10 the overall defect that gives us essentially the point 11 at which we go Re critical.

12 When we do the CHS evaluation, we actually 13 transition and use a density-based curve, so that's 14 sort of a state point difference that's applied. And 15 again, the calculation was actually updated to just 16 apply the density curve through the entire transient.

17 So with DHR we actually use the 18 temperature transient; generate a power overshoot, and 19 that's used in core physics codes to calculate LOCA 20 peaking factors. They are then applied in the CHF 21 analysis, and then CHF techniques similar to what's 22 used in the LOCA models is applied in terms of 23 evaluating the CHF.

24 And so these are, I think, the updated 25 figures in the SR. Again, we have the DHR-driven NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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241 1 power overshoot. This particular case doesn't actuate 2 ECCS at the peak; that's only performed through the 3 CHF case, so this is the equilibrium DHR condition, 4 and this is was the type of analysis that was used 5 with the stability that Dr. Yarsky was talking about 6 earlier today.

7 So we take these, apply them in a way in 8 PIM and evaluate the decay ratio that results. Here's 9 a figure of the temperature conditions.

10 So this is where we take the same scenario 11 and actually open the valves at the point of the power 12 peak. That drives the density-driven feedback. We 13 reach sub-criticality. I don't have the extended 14 figures in this analysis.

15 There's an RER response that shows if you 16 were to extend this model, which really wasn't the 17 purpose; the purpose was really to make the argument 18 that actuating ECCS up here is more limiting than 19 anything that would be kind of self-limited 20 oscillations down in this regime.

21 But because the reactivity balance is so 22 far from what's more characteristic of the core, it 23 does show extended oscillations out this way.

24 And the CHS results: the conclusion in 25 the SR, it's the conclusion still that this is --

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242 1 either way you model it, it's non-limiting from a CHF 2 or SAFDLs perspective in terms of transience. We're 3 still under review on some of the finer points that 4 I'm sure Jeff will cover more. And that's all I have 5 on that topic.

6 MEMBER CORRADINI: Just, can we go back?

7 MR. BRISTOL: Yes.

8 MEMBER CORRADINI: But these, what you 9 view as bounding analyses with a point kinetics in 10 dynamic, in difference to the previous plot that we 11 were asking about, which essentially is a what-if in 12 terms of power, right?

13 MR. BRISTOL: That's right.

14 MEMBER CORRADINI: Okay.

15 MR. BRISTOL: That's right.

16 MEMBER CORRADINI: So these are steady-17 state calculations of a series of what-ifs, whereas 18 the other two are considered to be your bounding 19 calculations for dynamic?

20 MR. BRISTOL: Yes. So we don't actually 21 believe there would be the dynamic response that we've 22 got presented. And just to kind of contextualize that 23 a little bit, again the purpose of that calculation 24 was to do it kind of generically enough that it would 25 be obvious that it was conservative.

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243 1 The downside of that is that it paints it 2 maybe a slightly variant picture as to the actual 3 conditions that we would reach in long-term cooling 4 phase.

5 MEMBER BROWN: Go back to 107. How far 6 out does the line go?

7 MEMBER CORRADINI: That's 266 minutes.

8 MR. BRISTOL: That's line -- yes, that's 9 where we truncated -- what we see extending this line 10 is --

11 MEMBER BROWN: How long do you stay at 12 power?

13 MR. BRISTOL: Pretty flat response.

14 MEMBER BROWN: What is it, three percent 15 or two and a half? Something like that?

16 CHAIR MARCH-LEUBA: No, that's megawatts.

17 MEMBER BROWN: Oh, megawatts; that's fine.

18 Whatever the power is, how long does it stay there?

19 MR. BRISTOL: So if we had no temperature 20 increasing in the pool it would stay at that solution 21 until an operator came and did something about it.

22 Again, this is the --

23 MEMBER BROWN: What would he do?

24 MR. BRISTOL: Insert the rod, add boron:

25 any number of things.

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244 1 MEMBER BROWN: But the rod's stuck.

2 MR. BRISTOL: Shut off the DHR and stop 3 the cooldown.

4 (Simultaneous speaking.)

5 MEMBER DIMITRIJEVIC: -- on-site power --

6 MR. BRISTOL: Actuate ECCS initiates the 7 same --

8 CHAIR MARCH-LEUBA: I thought the obvious 9 action in the ECCS would be to flood containment with 10 borated water from the pool.

11 MR. BRISTOL: Yes, that would be --

12 CHAIR MARCH-LEUBA: And open ECCS.

13 MR. BRISTOL: Right, right. Again, this 14 isn't really characteristic of the timing of the 15 event. The purpose of this analysis is to show that 16 we can conservatively evaluate.

17 CHAIR MARCH-LEUBA: Yes, I just feel that 18 a 100 percent realistic analysis will show that this 19 doesn't happen. I mean, it would have been nice that 20 this didn't happen, and we didn't need GDC 27 21 exception, because look at Charlie. Every time he 22 talks about it, he gets upset, and he is pro-nuclear.

23 (Laughter.)

24 CHAIR MARCH-LEUBA: All members of the 25 public will hear this, they say, Oh, what are you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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245 1 building in my neighborhood? I think it's just an 2 issue of the conservatisms that we had it. I don't 3 think this happens.

4 MEMBER BROWN: It's not a matter of 5 principle. Reactors should shut down under some 6 particular design condition. The ones I'm familiar 7 with that I've worked with over 35 years, when the rod 8 stuck out, we shut down.

9 MR. BRISTOL: Sure.

10 MEMBER BROWN: We had to work at it. It 11 wasn't easy, okay? We had to work at it, but that's 12 what you did.

13 MR. BRISTOL: And I think, to that point, 14 because of the unique aspects of our reactor design, 15 that one stuck-out rod, which is traditionally applied 16 to a big core, is a ton of margin, and --

17 CHAIR MARCH-LEUBA: His reactors are this 18 size.

19 MR. BRISTOL: Fair enough. It's a lot of 20 margin for our design, and so in demonstrating passive 21 cooling for a coping period that outlives your poison, 22 that's a lot of temperature defect to try to 23 accommodate in deterministic analysis phase.

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246 1 that would show, yes, we can squeak by without a 2 critical configuration.

3 CHAIR MARCH-LEUBA: Can you go back to the 4 dotted line? The one with -- that one. So if I read 5 this correctly, all the red points have a K effective 6 less than one.

7 MR. BRISTOL: Yes.

8 CHAIR MARCH-LEUBA: All the blue points 9 that are above the line are K effective less than one, 10 and same with the this. Anything that is under the 11 line has a K effective greater than one and it boils 12 to maintain criticality.

13 MR. BRISTOL: Or would respond back up to 14 that temperature equilibrium condition.

15 MEMBER CORRADINI: They'd move up. It 16 would move up.

17 MR. BRISTOL: So if the coolant were to 18 drive you to here, you'd bump back up to here and --

19 CHAIR MARCH-LEUBA: You're grossly 20 overcooling. Yes, okay. So if -- you could do it 21 with sub-cooling or with voids.

22 MR. BRISTOL: Right.

23 CHAIR MARCH-LEUBA: Correct. So the 24 nominal case for this is the ECCS, which is purple.

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247 1 yes? Anything less than one percent decay heat.

2 MEMBER BROWN: You stay critical.

3 CHAIR MARCH-LEUBA: When do you get one 4 percent decay heat for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />?

5 MEMBER BROWN: No, it's below.

6 MR. BRISTOL: That's a good question. I 7 would have to follow up on --

8 MEMBER CORRADINI: What do you mean?

9 CHAIR MARCH-LEUBA: You get one percent 10 heat is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />?

11 MEMBER CORRADINI: Three hours.

12 CHAIR MARCH-LEUBA: Three hours?

13 MR. BRISTOL: Three hours.

14 MEMBER CORRADINI: Based on the ANS 15 standard. Rule of thumb is 10,000 seconds-ish.

16 DR. SCHULTZ: Did you put a team on this 17 to do what Jose suggested? Look at the best estimate?

18 Forget about the uncertainties for a moment and do the 19 evaluation approach with the best code you've got and 20 the best assumptions you can make and worry about 21 applying uncertainty and evaluation later?

22 MEMBER BROWN: Steve's got a reasonable 23 suggestion. The point is, this is a conservative 24 analysis, and you do a best-estimate analysis under 25 the particular circumstances. If you've done it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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248 1 right, it doesn't go critical.

2 (Simultaneous speaking.)

3 MEMBER CORRADINI: I think there's -- I 4 guess maybe I'm different than the rest --

5 (Simultaneous speaking.)

6 MEMBER BROWN: -- get up critical and stay 7 critical.

8 MEMBER CORRADINI: Yes, but I think 9 there's a bounding calculation. That's the way I look 10 at it. There's a set of bounding calculations under 11 two initiators. If I start doing best estimates, I 12 don't have two initiators; I've got a plethora of 13 these, so I've got a plethora of best estimates. And 14 as long as they stay under this, then --

15 CHAIR MARCH-LEUBA: No, because they all 16 end up in the same place. They'll end up losing DC 17 power after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or pull a timer and open the ECCS 18 valve and going into the ECCS cooling through the 19 vessel. It continues through; any initiator in 20 transient will end up there, in that position.

21 MEMBER DIMITRIJEVIC: Well, you have loss 22 of power and stack roll.

23 CHAIR MARCH-LEUBA: You have stack roll --

24 MEMBER DIMITRIJEVIC: And loss of power.

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249 1 power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, plus.

2 MEMBER DIMITRIJEVIC: Well, I mean, I 3 think -- it's 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> because, I mean, if you can get 4 that, you can get boron after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, you will not 5 get --

6 CHAIR MARCH-LEUBA: Yes. And you don't 7 take credit for any of the on-site AC power in these 8 times.

9 MR. BRISTOL: So I think to answer that 10 question, yes, there was a whole lot of consideration 11 that went into any ways that we could, in Chapter 15 12 space with Chapter 15 rules, and it starts -- you 13 know, you start pulling the thread, and it starts 14 breaking down because again, when deterministically 15 taking that stuck rod is driving the problem, and so 16 if I start trying to pull in, Well, my pull is really 17 going to be 100 degrees. Yes, that totally changes 18 the dynamic response; it improves it quite a bit.

19 Am I going to operate a plant with a tech 20 spec of 100 degrees? No. So that's not a thread that 21 I can pull.

22 CHAIR MARCH-LEUBA: Yes, but that's still 23 -- presume these results: say it is considerable that 24 you can have these, but under nominal conditions, when 25 my pool is going to be at 90 degrees or maybe 78 --

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250 1 MR. BRISTOL: Certainly.

2 CHAIR MARCH-LEUBA: -- under nominal 3 conditions, it will have it. Because if you could put 4 75 percent decay heat, you have enough heat up, it 5 boils or heat up that pool will move up.

6 MR. BRISTOL: Yes.

7 CHAIR MARCH-LEUBA: So you're just killing 8 yourself and ourselves with you by being too 9 conservative. I would rather not have to give you an 10 exception for anything, because you're satisfied.

11 It's a marketing decision.

12 MEMBER BROWN: It's a principles decision.

13 MEMBER CORRADINI: We're shooting for 14 about an hour late. Do we want to continue, or do you 15 want to go into closed session?

16 MEMBER BROWN: I'm finished here.

17 MEMBER CORRADINI: I think we have to hear 18 from the public.

19 MS. MCCLOSKEY: I have one correction to 20 make. When we were discussing the margin to the boron 21 precipitation and the solubility limits, in most of 22 the long-term cooling cases that we have, there's 30 23 degrees margin of solubility limit or more. The 24 limiting case for an injection line break at low power 25 conditions, and therefore low decay heat has about 16 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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251 1 degrees margins of solubility temperature. I needed 2 to get that correction in.

3 MEMBER CORRADINI: Thank you.

4 MEMBER DIMITRIJEVIC: I have a question.

5 Because of the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> loss, when you did this 6 analysis, did you just assume the loss of power is not 7 favorable? It doesn't matter how long, right?

8 MR. BRISTOL: That's correct.

9 MEMBER DIMITRIJEVIC: Because if you said 10 that this occurred in three hours, and you analyzed 11 this, if you get loss of power only before three 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is irrelevant. It's just that you 13 lost on-site power?

14 CHAIR MARCH-LEUBA: You need to open the 15 ECCS valve so you would have additional cooling to the 16 vessel to get here.

17 MEMBER DIMITRIJEVIC: I know, but that's 18 okay. I mean, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; why do you -- you don't have 19 on-site power. That's what I'm --

20 CHAIR MARCH-LEUBA: You have batteries for 21 24 hours. That's what the assumption is. The 22 batteries last for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and then the ECCS open, 23 and --

24 MEMBER DIMITRIJEVIC: So there is no on-25 site power, and you will have your batteries deplete.

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252 1 CHAIR MARCH-LEUBA: You have a timer that 2 you should log.

3 MS. MCCLOSKEY: There's a timer.

4 MEMBER DIMITRIJEVIC: Yes. How do you 5 open the ECCS in this case? I mean, what happens?

6 MEMBER CORRADINI: It just fail --

7 MR. BRISTOL: So the transient -- if we go 8 to this case, the event progression would be starting 9 at hot zero power conditions in terms of RCS 10 temperature and inventory, and then a simultaneous 11 loss of AC and DC such that we go to IAB.

12 And right around the time that we're 13 starting to deterministically applying an IAB release 14 condition at this particular point.

15 MEMBER DIMITRIJEVIC: So you lost AC and 16 DC simultaneously?

17 MR. BRISTOL: In this particular case, 18 yes, because there would be no -- Dr. Yoo was saying 19 that there wouldn't be an actuation of ECCS until the 20 24-hour point, except that ECCS was already actuated.

21 So we're just riding the pressure down to the IAB 22 release point and then sort of deterministically 23 applying it as a release conservatism at the time of 24 the power peak. So that's how we get the ECCS 25 actuation.

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(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

253 1 CHAIR MARCH-LEUBA: This one you assume a 2 DC failure of type zero?

3 MR. BRISTOL: Mm-hmm.

4 CHAIR MARCH-LEUBA: So as they cool down 5 is the one that triggers the --

6 MEMBER CORRADINI: Right.

7 MR. BRISTOL: Right.

8 CHAIR MARCH-LEUBA: Yes. That's even less 9 probable.

10 MR. BRISTOL: Right.

11 MEMBER CORRADINI: But that's why -- I 12 think we're circling here. That's why I guess my 13 personal view is, these are bounding. We have to make 14 sure the range of potentials are bounded by them, but 15 to try to hit a best estimate, I see that to be a very 16 difficult assignment.

17 CHAIR MARCH-LEUBA: Oh, this is bounding, 18 but that's two hours after shutdown, and the xenon is 19 sufficient to keep you way down.

20 (Simultaneous speaking.)

21 CHAIR MARCH-LEUBA: In the 24-hour --

22 (Simultaneous speaking.)

23 CHAIR MARCH-LEUBA: -- because xenon is 24 getting out. Unless you started from a xenon-free 25 startup and you have this crammed right in -- well, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

254 1 anyway, we have to live with it.

2 MEMBER CORRADINI: Should we go to the 3 public line? It's open? So does anybody in the room 4 want to make a statement?

5 And on the public line, do we have anybody 6 on the public line? Can you please acknowledge that 7 you're out there?

8 Okay. Why don't we close the public line, 9 please? And then can I ask NuSCALE and staff to make 10 sure that -- we're going to close the line. We'll 11 keep the NuSCALE staff line open, and we'll go into 12 closed session. We'll just verify the public line is 13 closed. Thank you, Mike. You guys going to stay 14 where you are, and we'll get another set of slides?

15 I can't wait.

16 CHAIR MARCH-LEUBA: He can misspell your 17 name.

18 MEMBER CORRADINI: Since we're kind of 19 quasi-open, quasi-closed, if I were to do a best 20 estimate, I would rather just take your steady-state 21 set of calculations and look and see the range of 22 uncertainty on those. I wouldn't go through a 23 dynamic. The dynamic to me just strikes me like a 24 dead end, but at least your steady-state ones were 25 looking for a cross point between a reactor static NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

255 1 calculation and a thermohydraulic parametrics.

2 That, at least, would show you the range 3 of what these things are as I make some assumptions of 4 what's realistic and what's not realistic.

5 DR. SCHULTZ: You have to set that base 6 case in some fashion and then do parametrics.

7 MEMBER CORRADINI: Because if you look at 8 their steady state one on slide whatever, it's 9 slightly more than the dynamic, because we're talking 10 about 15 megawatts, where you're predicting in the 11 dynamic of about 10 megawatts under those two bounded 12 conditions.

13 So I'd rather, if I were an analyst, I'd 14 rather try to noodle with this, which is a whole lot 15 easier to do to try to see what the uncertainties 16 might be, or what the range of uncertainties are, or 17 the assumptions. Because you neglected a whole bunch 18 of things to get to this already.

19 MR. BRISTOL: Is actually close to two.

20 MEMBER CORRADINI: Oh, I thought it was 21 one percent of decay heat. Oh, one percent of decay 22 heat.

23 MR. BRISTOL: One percent of reactor 24 power.

25 (Simultaneous speaking.)

NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

256 1 MEMBER CORRADINI: That's 15 megawatts.

2 CHAIR MARCH-LEUBA: No, 1.6.

3 MEMBER CORRADINI: Oh, one percent; 1.6?

4 I'm sorry.

5 MR. BRISTOL: Yes, so the dynamic response 6 is supposed to be at 10 percent.

7 MEMBER CORRADINI: Okay. Sorry. But this 8 would be --- well, no, I guess I'm -- this would be --

9 I wouldn't look at that as the dynamic. I'd look at 10 that at the long term -- the one that Charlie was 11 asking about, where I set a limit.

12 MEMBER BROWN: The 104?

13 MEMBER CORRADINI: It's the one over long 14 periods of time, yes? It's not the peak; it's --

15 MEMBER BROWN: It's 104, not 108 --

16 MEMBER CORRADINI: Yes. That's what I 17 would --

18 (Simultaneous speaking.)

19 CHAIR MARCH-LEUBA: Yes. But that's a 20 calculation based on a constant MTC using point 21 kinetics. That's crazy. I mean, that has nothing to 22 do with reality.

23 Since we're offering advice, what I would 24 do is, I would take and pull up five. I would input 25 the power one percent decay heat, 1.6 megawatts, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

257 1 let it settle to whatever it wants to settle, on the 2 flow, on the void, or the temperatures. Then take 3 that void and temperature and put in SIMULATE.

4 It's going to factor into the one or not.

5 That's a DC calculation. I think you complicated it 6 too much by doing the critical search. And each of 7 those purple points, you're -- it's -- but then on a 8 closed-loop calculation, I recalculate what the flow 9 is, I recalculate what the voids are, and put that 10 into SIMULATE. That's what I would do.

11 MEMBER CORRADINI: Okay. So you had two 12 pieces of advice, and they're worth about as much as 13 the people that gave them to you, since there's no 14 money involved. So we're closed.

15 (Whereupon, the above-entitled matter went 16 off the record at 4:31 p.m.)

17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

(202) 234-4433 WASHINGTON, D.C. 20005-3701 (202) 234-4433

LO-0619-65988 June 17, 2017 Docket No. PROJ0769 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Topical Report - Evaluation Methodology for Stability Analysis of the NuScale Power Module, PM-0619-65962, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on June 19, 2019. The materials support NuScales presentation of the Evaluation Methodology for Stability Analysis of the NuScale Power Module topical report.

The enclosure to this letter contains the nonproprietary version of the presentation entitled ACRS Subcommittee Presentation: NuScale Topical Report - Evaluation Methodology for Stability Analysis of the NuScale Power Module.

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-7H4 Michael Snodderly, NRC, TWFN-2E26 Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Bruce Bavol, NRC, OWFN-8H12

Enclosure:

ACRS Subcommittee Presentation: NuScale Topical Report - Evaluation Methodology for Stability Analysis of the NuScale Power Module, PM-0619-65962, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0619-65988

Enclosure:

ACRS Subcommittee Presentation: NuScale Topical Report - Evaluation Methodology for Stability Analysis of the NuScale Power Module, PM-0619-65962, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

ACRS Subcommittee Presentation NuScale Topical Report Evaluation Methodology for Stability Analysis of the NuScale Power Module June 19, 2019 1

PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Presenters Dr. Yousef Farwila System Thermal Hydraulics Ben Bristol Supervisor, System Thermal Hydraulics Matthew Presson Licensing Specialist 2

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Agenda

• Introduction

• Stability Solution Type

• Stability Investigation Description

- Theoretical

- Numerical Using New Code PIM

- Experimental Benchmark

• Procedure and Methodology

• Summary

• Questions and Discussions 3

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The Main Message

• The NuScale Power Module (NPM) design was found to be stable in the entire range of normal operation

• Outside of normal operation, the reactor is destabilized when the riser flow is voided, however

- Unstable flow oscillation amplitude is limited by nonlinear effects and the critical heat flux ratio actually improves

• The stability threshold is protected by scram upon loss of riser inlet subcooling

- Conceptually equivalent to a region exclusion not a detect and suppress solution type

- No action required to implement a stability solution hardware

• These conclusions are based on extensive first principles, experimental, and computational studies.

4 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Stability Evaluation

• Natural circulation instabilities were reported

- See for example D.S. Pilkhwal et al., "Analysis of the unstable behaviour of a single-phase natural circulation loop with one-dimensional and computational fluid-dynamic models," Annals of Nuclear Energy 34 (2007) 339-355.

a) HHHC: horizontal heater and horizontal cooler (the only unstable configuration);

b) HHVC: horizontal heater and vertical cooler; c) VHHC: vertical heater and horizontal cooler; d) VHVC: vertical heater and vertical cooler (qualitatively like NuScale module)

(a) (b) (c) (d)

HHHC HHVC VHHC VHVC

• Investigation of the NuScale module stability commenced to demonstrate stability, identify threshold conditions, and license stability protection methodology 5

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Stability Investigation Elements Module Design and Operational Domain First Principles Theory -- Experience -- PIRT Construct Models and Main Code Independent Models (PIM) Experimental Data and Reduced Order Model Benchmarking (RADYA) (NIST-1)

Comprehensive Analysis & Results

• Steady State Perturbations

• Stability during Transients

Conclusions:

• Stable within Operating Domain

• Threshold is Riser Voiding Stability Solution:

• Protect riser subcooling with margin

• Hardware already part of MPS 6

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Theoretical Investigation

• Kick off with an expert committee to generate a first PIRT

• Scoping review of thermalhydraulic instability modes and contrasting with the NPM design features

• Identification of the possible instability mechanism

• Analysis from first principles

- Riser-only mode (separate from cold leg)

- Stability trend with power using a simple SG model

- Inform design of stability experiments

• All medium ranked phenomena treated as highly ranked 7

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Theoretical and First Principles

• A system with feedback processes may undergo oscillatory instability if the feedback is:

- Negative (positive feedback is unconditionally unstable)

- Delayed

- Sufficiently strong

• NuScale natural circulation mode is examined

- Feedback is negative. A perturbation increasing core flow decreases exit temperature thus decreases riser density head

- Feedback is delayed. Transport delay for core exit condition to fill the riser and reach maximum density head effect.

- Feedback strength is related to liquid thermal expansion and possibility of phase change, riser length, SG characteristics, reactivity feedback Requires detailed modeling 8

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Main Stability Analysis Tool: PIM

• Transient 1-D 2-phase non-equilibrium primary loop flow Pressurizer Superheated Steam Cooling Heat Exchanger Feedwater Core 9

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Model Equations of the PIM code

• Thermalhydraulic conservation equations t time

- Liquid and vapor mass balance Ml liquid mass dM l , n Mg vapor mass

= m l , n 1 m l , n n m l liquid mass flow rate dt m g vapor mass flow rate dM g , n rate of evaporation

= m g , n 1 m g , n + n dt I integrated momentum Pgrav gravitational press. drop

- Mixture momentum conservation with drift flux Pfriction friction pressure drop (integrated momentum) Plocal local pressure drop Presid residual pressure drop dI

= Pgrav Pfriction Plocal + Presid hl liquid enthalpy dt h fg latent heat Q power

- Energy conservation (assume saturated vapor) n control volume index

( M l , n hl , n ) m l , n 1hl , n 1 m l , n hl , n n h fg + Q n d n 1

= upstream index dt 10 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Model Equations of the PIM code

• Point Nuclear Kinetics C Concentration of the delayed neutron precursors d Decay constant of the delayed neutron precursors

= ( 1) + C Neutron flux amplitude dt Delayed neutron fraction dC

= C Prompt neutron lifetime dt Reactivity

- Thermalhydraulic model provides reactivity input

• Moderator density reactivity feedback model (equivalent to moderator temperature reactivity under single-phase flow)

• Doppler fuel temperature reactivity feedback

- Heat source from neutron kinetics feeds back to thermalhydraulics

• Energy deposited in fuel pellets (proportional to neutron flux)

• Fraction of fission energy deposited directly in coolant

• Decay heat: input by the user as fraction of initial power 11 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Model Equations of the PIM code

• Heat conduction in fuel rods

- Pellet conductivity is function of temperature and burnup

- Driven by energy deposited in fuel pellets

- Heat flux at outer rod surface as power source to coolant

- Pellet temperature needed for Doppler reactivity

• Heat transfer models for heat sink (steam generator)

- Secondary side flow is driven by user-provided inlet forcing function

- Secondary flow is subcooled, 2-phase equilibrium, and superheated

- Primary flow parameters calculated from transient conservation equations

- Heat transfer between primary and secondary flow

• Heat transfer correlations

• Transient heat conduction in tube walls 12 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Model Equations of the PIM code

• Closing Relations and Correlations

- Frictional pressure drop (single- and two-phase friction and local losses)

- Drift flux parameters

- Non-equilibrium evaporation and condensation model

- Thermodynamic properties for water

- Physical material properties (pellets, cladding, SG tubes)

- Pellet-clad gap conductance

- Reactivity coefficients as functions of exposure and moderator density

• What is not modeled

- Pressurizer; pressure is input provided constant or forcing function

- Heat transfer through riser wall, adiabatic riser is default option

- Heat capacity of structures; only ambient heat losses through vessel 13 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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PIM Results of Perturbing SS

• Purpose is to calculate stability parameters of decay ratio and period at different conditions of power and exposure

- Following a user-applied small perturbation flow will oscillate

- Oscillations will grow with time if system is unstable

- Oscillations will decay eventually returning to the pre-perturbation state if the system is stable

• Stability parameters, decay ratio and period, are extracted from the transient output. Observations:

- Unconditional stability in the entire operational range

- DR decreases with power and exposure

- Period also decreases with power

- Observations agree with the independent Reduced Order Model 14 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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PIM Application Methodology

• For perturbations of steady state to get DR

- Vary power within 5-100% of rated

- BOC and EOC, and any point in between if warranted

- Conservative assumptions for MTC and decay heat fraction

• For a depressurization transient (scram not credited)

- Verify that unstable oscillations limit cycle without CHFR decrease

• Stability conclusion is generic, but confirmation is needed

- For plant upgrades such as power uprates

- Plant operation changes such as operating temperatures and maximum boron concentration

- Changes in fuel design that would change natural circulation flow 15 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Stability Solution

• Region Exclusion for NuScale

- Unstable region defined by a single parameter (core exit subcooling)

- Monitor and protect margin to riser exit subcooling (with temperature margin below saturation point at pressurizer pressure)

- Operator alarm when subcooling margin is approached

- Riser exit subcooling will be controlled by the reactor protection system as part of normal operating limits - not only for preventing instabilities

- Generic solution: there are no fuel or cycle design elements 16 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Summary and Conclusions

• Stability of the NuScale module was evaluated using a dedicated code (PIM) and supported by first principles analysis and experimental data benchmarking

• The module was found unconditionally stable within normal operation domain using conservative criterion

• Stability boundary identified as associated with riser voiding (loss of riser inlet subcooling)

• Stability protection methodology protects riser inlet subcooling with a margin to define the exclusion region enforced by the module protection system with scram 17 PM-0619-65962 Revision: 0 Copyright 2019 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 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 18 PM-0619-65962 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Presentation to the ACRS Subcommittee Staff Review of NuScale Topical Report TR-0516-49417-P, REVISION 0 EVALUATION METHODOLOGY FOR STABILITY ANALYSIS OF THE NUSCALE POWER MODULE Presenters:

Ray Skarda, Ph.D.- Reactor Systems Engineer, RES Peter Yarsky, Ph.D.- Senior Reactor Systems Engineer, RES Bruce Bavol - Project Manager, Office of New Reactors June 19, 2019 (Open Session)

Non-Proprietary 1

NRC Technical Review Areas/Contributors

• Ray Skarda - RES/Division of Systems Analysis (DSA)/Code and Reactor Analysis Branch (CRAB)

• Peter Yarsky - RES/DSA/CRAB

• Rebecca Karas (BC) - NRO/Division of Engineering, Safety Systems and Risk Assessment (DESR)/Reactor Systems, Nuclear Performance, and Code Review Branch (SRSB) 2 Non-Proprietary

Staff Review Timeline

• NuScale submitted the Topical Report (TR) TR-0516-49417-P, Evaluation Methodology for Stability Analysis of the NuScale Power Module, on July 31, 2016, (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML16250A851). Applicant provided supplemental information by letter dated December 3, 2016 (ADAMS Accession No.

ML16340A756).

• Staff issued 62 requests for additional information (RAIs) with NuScale providing responses - all responses were resolved/closed.

• Staff plans to brief advisory committee on reactor safeguards (ACRS) full committee on July 10, 2019

• Staff plans to issue its final SER in late August 2019.

• Staff plans to publish the -A (approved) version of the TR in late November 2019.

3 Non-Proprietary

Outline Primary Review Areas

• Regulatory Criteria

• Long Term Stability Solution

• Instability Modes and Phenomenology

• PIM Evaluation Model

• Uncertainty and Acceptance Criteria

• Stability with Worst-Rod-Stuck-Out (WRSO)

• Stability Topical Report Conclusions

• Design Certification Document (DCD) 15.9 Stability 4 Non-Proprietary

Regulatory Criteria General Design Criteria (GDCs) from Design-specific Review Standard 15.9.A

• GDC 10, Reactor Design, requires that specified acceptable fuel design limits (SAFDL) not be exceeded during any condition of normal operation, including conditions that result in unstable power oscillations with the reactor trip system available.

• GDC 12, Suppression of Reactor Power Oscillations, requires that oscillations be either not possible or reliably and readily detected and suppressed.

• GDC 13, Instrumentation and Control, includes requirements for the hardware implementation of long term stability (LTS) solution.

• GDC 20, Protection System Functions, requires the reactor protection system to initiate automatic action so unstable power oscillations are avoided.

• GDC 29, Protection against Anticipated Operational Occurrences, requires stability LTS solution design for an extremely high probability of accomplishing safety functions.

5 Non-Proprietary

Long Term Stability Solution (LTSS)

Exclusion Region Based Solution

• The NuScale LTSS is based on an exclusion region principle. GDC 12 and GDC 10 are met by preventing instabilities that could challenge specified acceptable fuel design limits (SAFDLs).

• The Module Protection System (MPS) precludes instability by enforcing riser subcooling margin and tripping the reactor. GDC 13, GDC 20, and GDC 29 are met by operation of the MPS to sense adverse conditions and trip the reactor.

6 Non-Proprietary

Instability Modes

• Dynamic and static instability modes were considered.

• Applicant identified and evaluated many modes.

• The applicants findings in terms of modes are consistent with staff findings from an independent Phenomena Identification and Ranking Table development process.

7 Non-Proprietary

PIM Evaluation Model The PIM Evaluation Model is Simple but Acceptable

• The PIM evaluation model includes simple models for thermal-hydraulics, reactor kinetics, fuel thermal-mechanical response, and steam generator tube heat conduction and heat transfer.

• Integral validation provided against NIST-1 stability tests.

8 Non-Proprietary

Decay Ratio (DR) Acceptance Criterion

• DR is insensitive to variations in most of the important phenomena over the PIM application range.

• DR Acceptance Criterion affords sufficient margin to account for bias and uncertainty.

• Numerical effects were considered as part of the DR bias.

9 Non-Proprietary

Stability with WRSO NuScale is stable at intermediate pressures

• Applicant conservatively analyzed stability for intermediate pressures (i.e, before emergency-core-cooling-system (ECCS) actuation).

• Strong moderator feedback increases likelihood of recriticality with WRSO, but strong moderator feedback is stabilizing.

• Applicants analysis demonstrates stability margin at intermediate pressures.

10 Non-Proprietary

Stability with WRSO NuScale will experience mild flow oscillation at low pressure

• After ECCS actuation, level drops below the riser, natural circulation flow pattern is broken and flow oscillations occur where core flow is driven by density head differences provided by void formation in the core region.

• Analyses performed by the applicant demonstrate flow oscillations that are not safety significant.

11 Non-Proprietary

Stability Topical Report Conclusions PIM-based Stability Analysis Method is Acceptable

• PIM is a simple model, but its models are anchored to upstream, high-fidelity models to improve accuracy.

• The DR is highly insensitive to variations in important phenomena and their models, leading to relatively small uncertainty in the DR.

• PIM predictions in steady-state and transients have been confirmed by the staff with independent TRACE confirmatory calculations.

• PIM is acceptable for performing stability analysis for the NuScale power module.

12 Non-Proprietary

Stability Topical Report Conclusions LTS Solution is Acceptable

• Primary instability mechanism properly identified by the applicant and confirmed by independent staff TRACE analysis.

• During normal, at power operation, the NuScale power module is very stable.

• The exclusion region based LTS solution is effective in preventing the reactor from becoming unstable during normal operation including the effects of AOOs.

• Potential instability during return to power with WRSO is not a safety concern.

• GDCs 10, 12, 13, 20, and 29 are met.

13 Non-Proprietary

DCD 15.9 Stability Review Stability Performance during Steady State Conditions is Acceptable

• Stable under steady-state conditions

- Analyses demonstrate at all power levels > 5 percent of rated that the DR remains well below the acceptance criterion.

• Transients analyzed

- Certain events result in new stable, steady state conditions.

- Certain events result in reactor trip due to MPS enforcement of the exclusion region prior to the onset of instability.

14 Non-Proprietary

DCD 15.9 Stability Review Stability Performance during Transients is Acceptable

• All AOO classes considered in the applicants analysis.

- Increase in heat removal by the secondary system

- Decrease in heat removal by the secondary system

- Decrease in reactor coolant system flow rate

- Increase in reactor coolant inventory

- Reactivity and power distribution anomalies

- Decrease in reactor coolant inventory

• LTS Solution is effective in preventing the occurrence of instability.

• Therefore GDCs 10, 12, 13, 20, and 29 are met.

15 Non-Proprietary

Questions/comments before the closed session starts?

16 Non-Proprietary

Backup Slides 17 Non-Proprietary

DCD 15.9 Stability Review Increase in Heat Removal by the Secondary System

• Analysis consistent with the stability analysis methodology topical report.

• Applicant analyzed maximum feed flow increase that does not produce an automatic, prompt MPS trip.

• PIM calculations confirm that the reactor remains stable 18 Non-Proprietary

DCD 15.9 Stability Review Decrease in Heat Removal by the Secondary System

• The staff reviewed the feedwater (FW) flow reduction event analyzed in the DCA.

• The DCA demonstrates that even mild FW flow events will progress in similar manners, eventually leading to a MPS trip based on hot-leg temperature (i.e., the LTSS).

• Therefore, the staff finds that the LTSS is effective in preventing the reactor from reaching an unstable condition.

• The staff review of the DCA analysis does not impact the staff review of the Stability TR.

19 Non-Proprietary

DCD 15.9 Stability Review Decrease in Reactor Coolant System Flow Rate

• CVCS pump over-speed would be considered an AOO.

• The staff found that a CVCS flow reduction or a reduction in secondary side heat removal event would bound this class of events.

• Analysis of CVCS pump over-speed is not required to demonstrate compliance with GDC 12.

20 Non-Proprietary

DCD 15.9 Stability Review Increase in Reactor Coolant Inventory

• Events that increase inventory but also increase reactor coolant system pressure are non-limiting because higher pressure increases stability margin.

• Staff considered events such as CVCS or spray malfunction that could increase inventory but maintain pressure, and these would be bounded by events that increase or decrease secondary side heat removal.

• Analysis of this class of event is not required to demonstrate compliance with GDC 12.

21 Non-Proprietary

DCD 15.9 Stability Review Reactivity and Power Distribution Anomalies

• A bounding event is analyzed based on a maximum control rod withdrawal that does not produce an automatic, prompt MPS trip based on high flux or high flux rate.

• PIM calculations confirm that the reactor remains stable.

22 Non-Proprietary

DCD 15.9 Stability Review Decrease in Reactor Coolant Inventory

• The limiting event is a slow depressurization where the low pressure trip is not credited.

• PIM calculations show that, eventually, the reduction in pressure leads to low riser subcooling, which initiated the LTS solution MPS trip to enforce the exclusion region.

• PIM calculations demonstrate that the onset of instability would occur well after the reactor is shutdown by control rods.

23 Non-Proprietary

DCD 15.9 Stability Review LTS Solution MPS Trip Timing 24 Non-Proprietary

LO-0619-65982 June 17, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation, NuScale FSAR Chapter 15, Transient and Accident Analyses, PM-0619-65981, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee meeting on June 19, 2019. The materials support NuScales presentation of Chapter 15, Transient and Accident Analyses, of the NuScale Design Certification Application.

The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Presentation, NuScale FSAR Chapter 15, Transient and Accident Analyses, PM-0619-65981, 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-7H4 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12

Enclosure:

ACRS Subcommittee Presentation, NuScale FSAR Chapter 15, Transient and Accident Analyses, PM-0619-65981, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0619-65982

Enclosure:

ACRS Subcommittee Presentation, NuScale FSAR Chapter 15, Transient and Accident Analyses, PM-0619-65981, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Nonproprietary ACRS Subcommittee Presentation NuScale FSAR Chapter 15 Transient and Accident Analyses June 19, 2019 1

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Presentation Team Ben Bristol - Supervisor, System Thermal Hydraulics Dr. Pravin Sawant (*) - Supervisor, Code Validation and Methods Dr. Brian Wolf (*) - Supervisor, Code Development Dr. Selim Kuran (*) - Thermal Hydraulic Software Validation Meghan McCloskey - Thermal Hydraulic Analyst Mark Shaver (*) - Supervisor, Radiological Engineering Matthew Presson - Licensing Project Manager Paul Infanger - Licensing Specialist Greg Myers (*) - Licensing Specialist

(*) On the phone 2

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Scope of Chapter 15 Evaluation of the safety of a nuclear power plant requires analyses of the plants responses to postulated equipment failures or malfunctions

• FSAR Ch 15 addresses deterministic design basis safety analyses

• FSAR Ch 15 summarizes results of the NuScale design basis events identified, event classification, methodology for analysis, event results and margin to acceptance criteria

• FSAR Ch 15 provides results demonstrating radiological dose consequences remain below acceptance criteria 3

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Chapter 15 Radiological Dose Consequences Acceptance Criteria Dose Event Location (rem TEDE) (rem TEDE)

EAB 25.0 <0.01 Iodine Spike Design Basis Source Term(1)

LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 <0.01 EAB 25.0 <0.01 Iodine Spike Design-Basis Source Term(1)

LPZ 25.0 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 0.63 Core Damage Event(2) LPZ 25.0 1.37 CR 5.0 2.14 EAB 25.0 <0.01 Main Steam Line Break LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 0.01 EAB 2.5 <0.01 Main Steam Line Break LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 0.08 Steam generator tube failure LPZ 25.0 0.08 (pre-incident iodine spike)

CR 5.0 0.20 EAB 2.5 <0.01 Steam generator tube failure LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 6.3 0.02 Primary coolant line break LPZ 6.3 0.04 CR 5.0 0.08 EAB 6.3 0.55 Fuel handling accident LPZ 6.3 0.55 CR 5.0 0.89 (1) The iodine spike DBST is not an event, rather it serves as a bounding surrogate for design-basis loss of primary coolant into containment events.

(2) The CDE is a beyond-design-basis special event.

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Chapter 15 Overview Agenda - June 19th

• NuScale design overview

• Ch 15 overview

• Analytical assumptions for Ch 15 analysis

• System T/H analysis methodologies

• Selected transient results

• Radiological analysis

• Ch 6.2.1 Containment response analysis

• Long term cooling 5

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 15 NuScale Design Overview Safety/Non Safety Systems June 19, 2019 6

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Power Module Overview Integral Pressurized Water Reactor steam line

• Core, steam generator and pressurize in one vessel feedwater line

• Integrated reactor design, no large-break containment vessel loss-of-coolant accidents

• Reactor coolant system operated in single reactor vessel phase (liquid) density driven flow pressurizer

• Safety decay heat removal systems are passive and fail safe steam generator

• Module protection system designed to automate event mitigation actuations (no operator actions) feedwater header reactor core module support skirt 7

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ECCS Emergency Core Cooling System

• ECCS valves open to a boiling/condensing circulation flow path to transfer decay and residual heat to reactor pool

- Liquid from containment vessel enters reactor vent valves RCS through reactor recirculation reactor vent valves valves

- Vapor vented from RCS to containment vessel through reactor vent valves

- Steam condenses on inside surface of containment vessel

- Heat transfer through vessel walls to the reactor pool reactor recirculation reactor recirculation

• Actuation Signals: High CNV valves valves level, 24hr loss of AC power

• Fail safe: ECCS valves open on loss of DC power 8

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Decay Heat Removal System (DHRS)

FWIVs Removes heat after loss of MSIVs normal cooling Boiling/condensing loop DHR actuation valves Two redundant trains Redundant actuation and reactor pool isolation valves on each train Initiates on:

DHR passive condenser Loss of power Loss of cooling indication (ESFAS Signal) 9 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 15.0 NuScale Design Overview Event Classification, Acceptance Criteria, Methodology June 19, 2019 10 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NPM Design Basis Event Identification

• Ch 15 initiating events considered for internal events in a single NPM while at power Ch 15 long term cooling analysis considered effects from up to 12 modules rejecting heat into the shared reactor pool UHS Ch 21 discusses the suitability of shared components and design measures taken to ensure these components do not introduce multi-module risk

• Use of NuScale PRA (DCA Section 19.1) as starting point to identify design basis initiating events PRA initiating events summarize scope of internal events that cause a reactor trip/plant transient response, considering NPM systems and as well as industry references and advanced reactor PRAs Examined systems identified as relevant to PRA initiating events for additional detail to identify specific impacts on module for design basis event identification 11 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NPM Design Basis Event Classification

• Event categories consistent with operating LWRs:

• Increase in heat removal by secondary system

• Decrease in heat removal by secondary system

• Decrease in RCS flow rate (n/a to NPM design)

• Reactivity and power distribution anomalies

• Increase in RCS inventory

• Decrease in RCS inventory

• Radioactive release from a subsystem or component

• NuScale-specific phenomena/event progressions:

• PWR stability

• Over-cooling return to power 12 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NPM Design Basis Event Classification

• Events categorized as AOOs, infrequent events, accidents

• Events that are expected to occur one or more times during an NPM lifetime (1E-2 per year or more) are classified as AOOs.

• Events that are not expected to occur are classified as IEs or postulated accidents, or may be conservatively classified as AOOs.

• Event classification is simplified by substituting a deterministic classification considering similarity to operating PWR event classification, where event consequences are small and calculation of a NuScale-specific event frequency is not warranted.

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Design Basis Events 15.1 Increase in heat removal by secondary system Sec. Event Classification System T/H Subchannel Radiological (code) (code) 15.1.1 Decrease in feedwater AOO Non-LOCA Yes n/a temperature (NRELAP5) (VIPRE-01) 15.1.2 Increase in feedwater flow AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.1.3 Increase in steam flow AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.1.4 Inadvertent opening of steam AOO Non-LOCA Yes n/a generator relief or safety valve (NRELAP5) (VIPRE-01) 15.1.5 Steam piping failures Postulated Non-LOCA Yes Yes accident (NRELAP5) (VIPRE-01) 15.1.6 Loss of containment AOO Non-LOCA Yes n/a vacuum/containment flooding (1) (NRELAP5) (VIPRE-01)

(1) NuScale unique event 14 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Design Basis Events 15.2 Decrease in heat removal by secondary system Sec. Event Classification System T/H Subchannel Radiological (code) (code) 15.2.1 Loss of external load AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.2.2 Turbine trip AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.2.3 Loss of condenser vacuum AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.2.4 Closure of main steam AOO Non-LOCA Yes n/a isolation valve (NRELAP5) (VIPRE-01) 15.2.5 Steam pressure regulator n/a n/a n/a n/a failure (closed) 15.2.6 Loss of non-emergency AC to AOO Non-LOCA Yes n/a station auxiliaries (NRELAP5) (VIPRE-01) 15.2.7 Loss of normal feedwater flow AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.2.8 Feedwater system pipe breaks Postulated Non-LOCA Yes Yes accident (NRELAP5) (VIPRE-01) 15.2.9 Inadvertent operation of the AOO Non-LOCA Yes n/a decay heat removal system (1) (NRELAP5) (VIPRE-01)

(1) NuScale unique event 15 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Design Basis Events 15.4 Reactivity and Power Distribution Anomalies Sec. Event Classification System T/H Subchannel Radiological (code) (code) 15.4.1 Uncontrolled control rod AOO Non-LOCA Yes n/a assembly withdrawal from (NRELAP5) (VIPRE-01) subcritical or low power 15.4.2 Uncontrollled control rod AOO Non-LOCA Yes n/a assembly withdrawal at power (NRELAP5) (VIPRE-01) 15.4.3 Control rod misoperation AOO Non-LOCA Yes n/a (NRELAP5) (VIPRE-01) 15.4.4 Startup of an inactive loop or n/a n/a n/a n/a recirculation loop at incorrect temperature 15.4.5 Flow controller malfunction n/a n/a n/a n/a causing an increase in core flow rate (BWR) 15.4.6 Inadvertent decrease in boron AOO Non-LOCA Yes n/a concentration in RCS (n/a) (VIPRE-01) 15.4.7 Inadvertent loading and IE n/a Yes n/a operation of a fuel assembly in (VIPRE-01) improper position 15.4.8 Spectrum of rod ejection Postulated Rod ejection accidents Yes Yes accidents accident (SIMULATE-3K, NRELAP5) (VIPRE-01) 16 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Design Basis Events 15.5 Increase in reactor coolant inventory Sec. Event Classification System T/H Subchannel Radiological (code) (code) 15.5.1 Chemical and volume control AOO Non-LOCA Yes n/a system malfunction (NRELAP5) (VIPRE-01) 17 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Design Basis Events 15.6 Decrease in reactor coolant inventory Sec. Event Classification System T/H Subchannel Radiological (code) (code) 15.6.1 Inadvertent opening of reactor AOO Valve opening event n/a n/a safety valve (NRELAP5) 15.6.2 Failure of small lines carrying IE Non-LOCA n/a Yes primary coolant outside (NRELAP5) containment 15.6.3 Steam generator tube failure Postulated Non-LOCA n/a Yes accident (NRELAP5) 15.6.4 Main steam line failure outside n/a n/a n/a n/a containment (BWR) 15.6.5 Loss of coolant accidents Postulated LOCA n/a n/a resulting from a spectrum of accident (NRELAP5) postulated piping breaks within the reactor coolant pressure boundary 15.6.6 Inadvertent operation of AOO Valve opening event n/a Yes (2) emergency core cooling system (NRELAP5)

(1)

(1) NuScale unique event (2) See FSAR Ch 12 18 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Event Acceptance Criteria -

Thermal/Hydraulic and Fuel Event Fuel RCS Pressure/ Containment Event Progression Classification (4) Clad (1) Main Steam Pressure (3)

System Pressure (2)

AOO MCHFR 110% Pdesign Pdesign Does not develop into a more limit serious condition without other faults occurring independently IE MCHFR 120% Pdesign Pdesign Does not cause a limit consequential loss of function of systems needed to cope with the fault Postulated MCHFR 120% Pdesign Pdesign Does not cause a Accident limit consequential loss of function of systems needed to cope with the fault

1. NuScale safety analysis methodologies developed to demonstrate fuel cladding integrity maintained.

Fuel centerline temperature is examined but not challenged due to low linear heat rates.

2. RCS and secondary side design pressure are equal, 2100 psia

3. Containment design pressure 1050 psia; for peak pressure and temperature analysis see FSAR 6.2.1

4. Event-specific acceptance criteria for LOCA and rod ejection accidents are applied 19 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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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 Subchannel topical report Radiological Non-LOCA topical report TR-0516-49416-P TR-0915-17564-P dose acceptance criteria Accident source term General approach for most non-LOCA analyses; topical report different codes/methods apply for some events such as for control rod ejection, fuel misload TR-0915-17565-P 20 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Evaluation Models -LOCA, Valve Opening, Containment Analysis Approach Plant design, Core design, Fuel rod design, Plant initial conditions, SSC performance NRELAP5 system T/H Long term Containment and hot cooling response channel response Water level MCHFR, above top of fuel, Containment Water level Cladding temp, peak pressure, above top of Boron temperature fuel precipitation LOCA topical report LTC technical report CNV technical report TR-0516-49422-P TR-0916-51299-P TR-0516-49084-P General approach for LOCA or valve opening event analyses 21 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Module Protection Functions

• Reactivity Control

- Reactor trip

- CVCS/Demineralized Water Isolation

• RCS and Secondary Inventory Control

- Containment Isolation

- Secondary Isolation

• Heat Removal

- DHRS Actuation

- ECCS Actuation

• Subcooling

- Reactor trip 22 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Event Mitigation Increase in heat removal Increase in RCS inventory transients transients

• Reactor trip

• Reactor trip

• Secondary Isolation

• CVCS Isolation Decrease in heat removal Decrease in RCS inventory transients transients

• Reactor trip

• Reactor trip

• DHRS Actuation

• CNV Isolation Reactivity and power

• ECCS actuation transients Stability

• Reactor trip

• Reactor trip

• Demineralized water isolation 23 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Pressure vs. Temperature Operation Map High Temperature Analytical Limit (610°F) 2100 High Pressure Analytical Limit (2000 psia) 2000 High Pressure Operating Limit (1920 psia) 1900 Minimum Critical Temperature (420°F)

Low Initial TAvg (535F) High Initial TAvg (555F)

Pressurizer Pressure (psia)

Normal Operating Pressure (1850 psia)

TCold TAvg THot 1800 Low Normal Pressure (1780 psia)

Low Pressure Analytical 1700 Limit (1720 psia)

Saturation Curve 1600 Low Low Pressure Analytical Limit (1600 psia)

Subcooled Margin 5°F 1500 400 450 500 550 600 650 Reactor Coolant System Temperature (°F)

Module protection system (Ch. 7, red) Technical specification LCOs (Ch. 16, blue) 24 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analytical Assumptions for Ch 15 Analysis

• Operator action

• Single failure

• Loss of power

• Scope of event progression 25 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Operator Actions

• No operator actions required to achieve safety functions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after an initiating event occurs

• Operator errors considered in identifying initiating events

• Operator actions allowed by procedures will make consequences less severe and therefore are bounded by Ch 15 analysis

• Multiple operator errors or errors that result in common mode failure are beyond design basis 26 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Single Failures Safety-related system Relevant single failure(s) (3) Comment considered (1) (2)

Module protection system Single failure of instrument Relevant for asymmetric channel reactivity events Containment isolation valves: Failure to close Nonsafety-related backup MSIV Main steam credited in safety analysis Feedwater Nonsafety-related feedwater regulator valve credited in safety analysis Feedwater line check valve Failure to close Nonsafety-related backup check valve credited in safety analysis Emergency core cooling system Failure of one RVV to open IAB failing to close upon Failure of one RRV to open demand (due to assumed loss of DC power supply) is not MPS failure to actuate one RVV treated as a single active failure and one RRV (1) The following systems were considered and no impact from a single failure was identified due to design redundancy: Containment isolation valves on CVCS, RCCWS, CFDS, CES piping; DHRS; reactor safety valves; demineralized water system isolation valves.

(2) The highest-worth control rod failing to insert is assumed in calculating scram worth 27 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Loss of Power Chapter 15 event analyses consider availability of AC power and DC power

• Loss of AC power at time of event initiation or time of reactor trip

• If AC power is lost, highly reliable DC power system may be assumed available or unavailable

• For non-LOCA type events, power availability affects whether ECCS valves actuate and what time they open

• For LOCA-type events, power availability affects the time ECCS valves actuate and when they open 28 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Loss of Power - Non-LOCA Event

• Availability of AC, DC power affects whether ECCS valves actuate, and what time they open 29 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Loss of Power -LOCA Event

• Availability of DC power affects whether ECCS valves open on level actuation or on IAB release

• Break size and credit for DHRS operation also affect whether ECCS valves open on level or on IAB release 30 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Design Basis Event Progression 15.0 Event Progressions

• Safety analyses of design basis events are performed from event initiation until a safe, stabilized condition is reached:

- Initiating event is mitigated

- Acceptance criteria are met

- System parameters such as inventory levels, temperatures, and pressures are trending in the favorable direction

• After safe, stabilized condition is reached:

- ECCS long term decay and residual heat removal

- Overcooling return to power

- Extended DHRS operation 31 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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System T/H Analysis Methodologies

• NRELAP5 code developed from RELAP5-3D

• Modified to address NuScale-specific phenomena/systems

• LOCA EM developed following RG 1.203 EMDAP

• LOCA EM extended to derived EMs for other events by addressing unique aspects 32 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LOCA EM Development: TR-0516-49422-P

• Followed RG 1.203 EMDAP process

• PIRT developed to identify high ranking phenomena for LOCA and valve-opening event short-term response

• Assessment basis developed that includes SETs and IETs to address high-ranked phenomena Unique phenomena addressed by NuScale-specific tests

• NRELAP5 code developed from RELAP5-3D to address NuScale-specific unique phenomena/systems

• Applicability evaluation performed including bottom-up and top-down analysis 33 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Non-LOCA EM Development: TR-0716-49416-P

• Non-LOCA evaluation model developed to perform conservative analyses, following intent of the RG 1.203 EMDAP and applying a graded approach

• PIRT developed to identify high ranking phenomena considering different types of non-LOCA events

• Gap analysis performed to evaluate how high ranked phenomena are addressed

• Additional NRELAP5 code validation performed focused on DHRS and integral non-LOCA response 34 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 15.0 Transient Examples June 19, 2019 35 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analysis Results 15.1 Increase in heat removal by secondary system Sec. Event (1) Peak RCS Pressure Peak SG Pressure MCHFR (Acceptance criteria) (< 110% Pdesign: 2310 psia) (< 110% Pdesign: 2310 psia) (> limit: 1.284)

(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.1.1 Decrease in feedwater temperature 1959 1432 1.921 15.1.2 Increase in feedwater 1936 1424 1.944 flow 15.1.3 Increase in steam flow 2018 1208 1.957 15.1.4 Inadvertent opening of steam generator relief 2156 1346 1.861 or safety valve 15.1.5 Steam piping failures 1992 1342 2.761 15.1.6 Loss of containment vacuum/containment 1936 1424 1.944 flooding (1)

(1) NuScale unique event Significant margin to acceptance criteria for all events 36 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: Increase in Feedwater Flow (IFF)

• Sequence of events 37 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IFF: Feedwater Flow/Reactor Power

• 100% increase in feedwater flow

• Event detected and steam generators isolated before both steam generators overfill, to preserve adequate DHRS capacity 38 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IFF: RCS Flowrate and Ave Temp

• RCS flow and average temperature response characteristic of NPM after reactor trip with DHRS actuation 39 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IFF: SG Pressure and Level

• Peak secondary side pressure occurs after DHRS actuation

• Event detected and steam generators isolated before both steam generators overfill, to preserve adequate DHRS 40 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IFF: DHRS Heat Removal and MCHFR

• DHRS removes decay and residual energy early in transient response

• MCHFR margin increases after reactor trip 41 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analysis Results 15.2 Decrease in heat removal by secondary system Sec. Event (1) Peak RCS Pressure Peak SG Pressure MCHFR (Acceptance criteria) (< 110% Pdesign: 2310 psia) (< 110% Pdesign: 2310 psia) (> limit: 1.284)

(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.2.1 Loss of external load 2158 1474 2.579 15.2.2 Turbine trip 2158 1474 2.579 15.2.3 Loss of condenser vacuum 2158 1474 2.579 15.2.4 Closure of main steam 2160 1481 2.567 isolation valve 15.2.6 Loss of non-emergency AC 2162 1361 2.569 to station auxiliaries 15.2.7 Loss of normal feedwater 2165 1434 2.569 flow 15.2.8 Feedwater system pipe 2164 1328 2.607 breaks 15.2.9 Inadvertent operation of the decay heat removal system 2163 1582 2.489 (1)

(1) NuScale unique event Significant margin to acceptance criteria for all events 42 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: Loss of AC Power 43 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LOAC: RCS Pressure - SG Pressure

• RCS pressure decreases quickly after 1st reactor safety valve lift; Peak secondary side pressure occurs after DHRS actuation

• Maximum pressures remain well below acceptance criteria 2310 psia 44 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LOAC: RCS Flow - RCS Temperature

• RCS flow and average temperature response characteristic of NPM after reactor trip with DHRS actuation 45 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LOAC: Net Reactivity - MCHFR Results for MCHFR case

• Heatup event does not challenge MCHFR acceptance criteria 46 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: Inadvertent Operation DHRS (IODHRS)

• Sequence of Events for spurious opening of 1 valve

- Peak SG pressure case 47 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IODHRS: SG Pressure - SG Level

• DHRS valve opening at power diverts some feedwater flow from steam generator, resulting in a slow overheating event

• Maximum steam generator pressure well below acceptance criteria 2310 psia 48 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IODHRS: RCS Pressure - Temperatures

• DHRS valve opening at power diverts some feedwater flow from steam generator, resulting in a slow overheating event

• Maximum RCS pressure well below acceptance criteria 2310 psia 49 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IODHRS: Power - MCHFR

• Heatup event does not challenge MCHFR acceptance criteria 50 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analysis Results 15.4 Reactivity and Power Distribution Anomalies Sec. Event (1) Peak RCS Press. Peak SG Press. MCHFR Fuel LHR (Acceptance criteria) (< 110% Pdesign: (< 110% Pdesign: (> limit: centerline (< 21.22 2310 psia) 2310 psia) 1.284) (< Tmelt) kW/ft)

(< 120% Pdesign: (< 120% Pdesign:

2520 psia) 2520 psia) 15.4.1 Uncontrolled control rod assembly withdrawal from NA 2038 685 >10 890.8 F subcritical or low power 15.4.2 Uncontrollled control rod 2160 1326 1.624 NA 8.97 kW/ft assembly withdrawal at power 15.4.3 Control rod misalignment NA NA 2.509 NA 7.10 kW/ft 15.4.3 Control rod withdrawal NA NA 1.624 NA 7.84 kW/ft 15.4.3 Control rod drop NA NA 1.641 NA 8.42 kW/ft 15.4.6 Inadvertent decrease in boron NA NA NA NA NA concentration in RCS 15.4.7 Inadvertent loading and operation of a fuel assembly in improper NA NA 1.916 NA 7.87 kW/ft position 15.4.8 Spectrum of rod ejection NA 2076 NA 2.477 2162 F accidents Control rod withdrawal has limiting MCHFR for reactivity events 51 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: Single Control Rod withdrawal (CRW)

• Sequence of Events 52 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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CRW: Withdrawn Reactivity and Power

• Reactivity insertion results in corresponding increase in power 53 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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CRW: RCS Pressure and Temp

• RCS pressure and fluid temperatures increase as power increases

• Limiting MCHFR case occurs for case with high power, high temperature, high pressure conditions 54 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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CRW: RCS Flow and MCHFR

• RCS flow increases as power increases

• MCHFR margin increases after reactor trip 55 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analysis Results 15.5 Increase in reactor coolant inventory Sec. Event (1) Peak RCS Pressure Peak SG Pressure MCHFR (Acceptance criteria) ( 110% Pdesign: 2310 psia) ( 110% Pdesign: 2310 psia) ( limit: 1.284) 15.5.1 Chemical and volume control 2130 1418 2.379 system malfunction Significant margin to acceptance criteria 56 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Analysis Results 15.6 Decrease in reactor coolant inventory Sec. Event (1) Peak RCS Pressure Peak SG Pressure MCHFR Additional (Acceptance criteria) (< 110% Pdesign: 2310 psia) (< 110% Pdesign: 2310 psia)

(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.6.1 Inadvertent opening of reactor safety valve NA NA NA NA 15.6.2 Failure of small lines carrying primary coolant outside 2047 1368 NA Note 2 containment 15.6.3 Steam generator tube failure 2073 1806 NA Note 2 15.6.5 Loss of coolant accidents resulting from a spectrum of Result: 1.796 Minimum level postulated piping breaks NA NA above top of within the reactor coolant Acceptance core: 1.5 ft pressure boundary criteria: > 1.29 15.6.6 Inadvertent operation of Result: 1.41 emergency core cooling 1936 588 system (1) Acceptance criteria: > 1.13 (1) NuScale unique event (2) Mass release and iodine spiking time provided as input to radiological analyses SG tube failure maximum secondary pressure remains below design pressure Valve opening and LOCA events demonstrate margin to acceptance criteria 57 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: SG Tube Failure (SGTF)

• Sequence of Events for limiting mass release case 58 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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SGTF: Reactor Vessel and SG Pressure

• SG pressure increases after secondary side isolation

• DHRS on intact steam generator provides heat removal 59 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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SGTF: Pressurizer and SG Level

• Primary side inventory lost to environment through tube failure until secondary side isolation 60 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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SGTF: Level Above Core - RCS Mass Flow

• Primary side inventory remains well above the top of the core

• DHRS on intact SG provides decay heat removal 61 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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SGTF: Break Mass Flow

• Total primary inventory released from the module limited by secondary isolation 62 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: Inadvertent RRV Opening (IORV)

• Sequence of Events 63 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IORV: RPV and CNV Pressure - RPV Level

• RPV level remains well above top of active fuel 64 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IORV: Reactor Recirculation Valve Flow

• After ECCS valves open, recirculation flow is established from containment into the RPV 65 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IORV: RCS and Cladding Temps

• Cladding temperatures decrease from steady-state condition 66 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IORV: RCS Flow

• Rapid RCS depressurization causes voiding in the core and momentary decrease in RCS flow leading to reduction in CHFR 67 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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IORV: Reactor Power - MCHFR

• MCHFR occurs in first few seconds of the transient, margin increase after reactor trip

• MCHFR result of 1.41 remains above acceptance criteria 1.13 68 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Example: LOCA

• Sequence of Events: 10% injection line break (0.53 inch equivalent diameter) 69 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LOCA: RCS and CNV Pressure, Level

• DHRS heat removal conservatively not credited in LOCA EM:

• ECCS valve opening on IAB release pressure is delayed until break flow sufficiently depressurizes RPV

• RCS level decreases below equilibrium level before ECCS valves open.

• Minimum level at time of ECCS valve opening 70 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 15.0 Radiological Analysis June 19, 2019 71 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Radiological Dose Consequences Event Summary and Acceptance Criteria Standard Acceptance Criteria Regulatory Guide Radiation Event Review Plan 1.183 Appendix(1) Source Section Exclusion Area Boundary And Control Low Population Zone(2)(3) Room(4)

(rem TEDE) (rem TEDE)

Loss-of-coolant accident 15.0.3 A 25 5 Damaged Fuel handling accident 15.7.4 B 6.3 5 fuel Rod ejection accident 15.4.8.A H 6.3 5 25 (Fuel damage or pre-incident Main steam line break 15.1.5.A E spike) 5 2.5 (Coincident iodine spike) 25 (Fuel damage or pre-incident Coolant activity F spike) 5 (with iodine Steam generator tube failure 15.6.3 spiking) 2.5 (Coincident iodine spike)

Primary coolant line break 15.6.2 n/a 2.5 5 Feedwater system pipe break 15.2.8 n/a 2.5(5) 5 (1) Appendices C, D, and G were not included because they are not applicable to the NuScale design.

(2) Based on 10 CFR 52.47 (LOCA), RG 1.183, and 10 CFR 20.1301.

(3) Individual at the EAB shall not receive dose limit for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period flowing the onset of release.

(4) Based on 10 CFR 52.47 and is for the duration of the event.

(5) Small fraction (10%) of regulatory dose reference value (25 rem TEDE).

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15.0.1 - Radiological Consequence Analyses Using Alternative Source Terms

• A modified methodology based largely on RG 1.183 alternative source term is used to evaluate radiological consequences of design basis events (DBEs) and the beyond-design-basis core damage event (CDE).

• Radiological consequences of the feedwater line break event are bounded by the consequences of a steam line break; radiological analysis not required

• Reactor building pool boiling radiological consequences evaluated to show negligible dose consequence

• Potential radiation shine exposures to operators within the control room following a radiological release event are evaluated to have no impact to total dose

• Doses remain below applicable limits for all events 73 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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15.0.2.4 - Radiological Analyses Methodology

• SCALE 6.1, TRITON, and ORIGEN-SCALE - used to calculate the time-dependent isotopic source term for radiological transport models

• NRELAP5 - used to provide event-specific thermal-hydraulic conditions for radiological transport models for DBAs, which dont have core damage.

• MELCOR - used to model the progression of severe accidents for the CDE

• ARCON96 - used to calculate onsite and offsite atmospheric dispersion factors for DBEs, and the CDE

• RADTRAD - used to estimate radionuclide transport and removal for the various DBEs and the CDE

• STARNAUA - used to perform aerosol removal calculations for the CDE

• NuScale pHT Code - used to calculate post-accident aqueous molar concentration of hydrogen ions for iodine re-evolution evaluation in the CDE

• MCNP6 - used for evaluating potential shine radiological exposures or doses to operators in the control room following a radiological release event.

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15.0.3 - Design Basis Accident Radiological Consequences

• Included in topical report TR-0915-17565 for completeness

• RG 1.183 Appendices C and D (BWR), and G (locked rotor) were not addressed because they are not applicable to the NuScale design

• RG 1.183 iodine spiking assumptions and decontamination factors for fuel handling accident were utilized

• No credit for iodine removal in piping and condenser

• Thermal-hydraulic response to an REA shows no resultant fuel failure; radiological analysis not required per RG 1.183 Appendix H

• Iodine spike DBST evaluated to serve as a bounding surrogate for design-basis loss of primary coolant into containment events

• ARCON96 atmospheric dispersion methodology used due to short distance to EAB and LPZ

• RADTRAD modeling techniques utilized consistent with RG 1.183 75 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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15.10 - Core Damage Event Radiological Consequences

• Detailed methodology provided in AST topical report TR-0915-17565

• Utilized NEI position paper on Small Modular Reactor Source Terms

• Event characteristics derived from a spectrum of MELCOR surrogate accident scenarios

• ARCON96 atmospheric dispersion methodology used due to short distance to control room, EAB and LPZ

• RADTRAD modeling techniques utilized consistent with RG 1.183

• STARNAUA used for modeling natural removal of containment aerosols for the CDE

• pHT evaluated to demonstrate negligible contribution of iodine re-evolution to dose consequences for the CDE 76 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Chapter 15 Radiological Dose Consequences Acceptance Criteria Dose Event Location (rem TEDE) (rem TEDE)

EAB 25.0 <0.01 Iodine Spike Design Basis Source Term(1)

LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 <0.01 EAB 25.0 <0.01 Iodine Spike Design-Basis Source Term(1)

LPZ 25.0 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 0.63 Core Damage Event(2) LPZ 25.0 1.37 CR 5.0 2.14 EAB 25.0 <0.01 Main Steam Line Break LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 0.01 EAB 2.5 <0.01 Main Steam Line Break LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 0.08 Steam generator tube failure LPZ 25.0 0.08 (pre-incident iodine spike)

CR 5.0 0.20 EAB 2.5 <0.01 Steam generator tube failure LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 6.3 0.02 Primary coolant line break LPZ 6.3 0.04 CR 5.0 0.08 EAB 6.3 0.55 Fuel handling accident LPZ 6.3 0.55 CR 5.0 0.89 (1) The iodine spike DBST is not an event, rather it serves as a bounding surrogate for design-basis loss of primary coolant into containment events.

(2) The CDE is a beyond-design-basis special event.

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 6.2.1 Containment Response Analysis June 19, 2019 78 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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6.2.1 CNV Response Analysis

• CNV designed to withstand full spectrum of primary and secondary system M&E releases with worst case single active failure and loss of power

• CNV response analysis methodology based on NRELAP5 described in TR-0516-49084 Rev 0

• NRELAP5 used to model integrated response of the RPV blowdown and CNV pressurization/heat removal

• Limiting event scenarios addressed:

- Primary side:

• LOCAs: CVCS discharge line, CVCS injection line, pressurizer high point vent line

• Valve opening events: Inadvertent RRV opening, Inadvertent RVV opening

- Secondary side:

• Main steam line break

• Feedwater line break 79 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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6.2.1 CNV Response Analysis

• Qualification of NRELAP5 code based on code qualification and NPM plant modeling approach described in :

- LOCA EM (for primary side pipe breaks and reactor valve opening events)

- Non-LOCA EM (for main steam line and feedwater line break events)

• In LOCA EM development, PIRT identified high ranked phenomena important for prediction of the containment response.

• In non-LOCA EM development, containment pressure response was considered as a FoM in the PIRT 80 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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6.2.1 CNV Response Analysis

• Containment response analysis based on models from LOCA EM or non-LOCA EM, with initial and boundary condition changes necessary to conservatively bias the mass and energy release and maximize the CNV pressure and temperature response

• In accordance with NuScale DSRS, initial RPV mass and energy maximized

• Maximum break sizes/valve sizes applied to maximize mass and energy release to the CNV

- Choked flow discharge coefficient Cd=1.0 applied

- Timing of ECCS valve opening evaluated based on level actuation setpoint and IAB release pressure to determine limiting condition 81 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Analysis Results Event Description CNV Pressure CNV Wall Temperature Limiting Case [Base Case] Limiting Case [Base Case]

(psia) (°F)

RCS discharge Break 943 [705] 510 [492]

RCS injection Line Break 959 [894] 526 [514]

RPV High point vent line 901 [554] 489 [471]

break Inadvertent RVV opening 911 [856] 486 [483]

Inadvertent RRV opening 986 [941] 512 [492]

Main steam line break 449 [< 449] 433 [< 433]

Feedwater line break 416 [< 416] 408 [< 408]

CNV design limits 1050 psia 550°F 82 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Pressure Case

• Initiated by inadvertent RRV Time (sec) Event opening 0 Inadvertent RRV opening

• Loss of normal AC and EDSS power at event initiation Loss of normal AC and DC power

• Single failure of remaining RRV to open FW/MS isolation Reactor trip

• Low bias IAB opening pressure 0.4 High CNV pressure - CNV

• Fast release of non- isolation condensable gas into CNV 74 ECCS valve opening on IAB assumed release

• No DHRS operation credited 91 Peak CNV pressure 986 psia

• Maximum CNV pressure: ~1800 CNV pressure < 50% peak 986 psia < 1050 psia design pressure pressure 83 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Pressure Case RPV Pressure Containment pressure Break and ECCS flow Energy balance 84 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Pressure Case Containment pressure After peak CNV pressure, CNV continues to depressurize and pressure remains well below 50% of the design limit TR-0916-51299-P:

Ch 15 long term cooling Containment pressure analyses demonstrate Longer-term effective decay and residual heat removal 85 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Peak pressure margin assessment

• Maximum pressure 986 psia < 1050 psia design pressure

• Considering both external pressure and DHRS,

~59 psi additional margin not credited

• Internal pressures are conservatively evaluated with an assumed external pressure of 0 psia Atmospheric pressure and pool hydrostatic head ~ 22 psi additional margin not credited

• Decay heat removal system is a single-failure proof, safety-related system Sensitivity calculations indicate ~ 37 psi additional margin could be gained Containment response analysis provides assurance that the NPM design demonstrates sufficient margin to satisfy the requirements of GDC 16 and GDC 50 86 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary ACRS Subcommittee Presentation Chapter 15 Long Term Cooling June 19, 2019 87 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Regulatory Requirements

• DSRS 6.3

• 10 CFR 50.46(b): - Maintain safe shutdown

- (4) Maintain coolable geometry conditions for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without operator actions and without

- (5) Provide long term cooling nonsafety-related onsite or after successful initial operation offsite power of the ECCS

• DSRS 15.6.5

• PDC 35: Provide abundant emergency core cooling to - Address boron precipitation transfer heat from reactor core - Steam generator tube rupture such that fuel and clad damage events shall also be reviewed as part of the LOCA break that could interfere with core spectrum analysis. The cooling is prevented and clad reviewer shall review the metal-water reaction is limited potential coolant inventory to negligible amounts loss from reactor vessel to the secondary side.

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Long Term Cooling Acceptance Criteria Core cooling is provided to NRELAP5 used to remove decay and residual heat from the core evaluate collapsed liquid

• Collapsed liquid level in the RPV level and core inlet remains above the top of fuel temperature

• Cladding temperatures remain acceptably low Coolable geometry is Collapsed liquid level and maintained core inlet temperature

• Boron precipitation analysis used to evaluate margin ensures boron concentration in the core remains below solubility to boron solubility limits limit 89 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling TR-0916-51299-P

• NRELAP5 used to model integral response of the decay heat removal path from ECCS and CNV heat transfer to the reactor pool

- LTC analysis performed for design basis shutdown conditions

- Confirmation of SAFDLs during overcooling return to power addressed by separate analysis

• PIRT high-ranked phenomena for ECCS cooling addressed by

- LOCA EM validation

- Bounded input/methodology

- Validation against NIST-1 tests HP-19a, HP-19b

• Qualification of NRELAP5 code based on code qualification and NPM plant modeling approach described in :

- LOCA EM (for primary side pipe breaks and reactor valve opening events)

- Non-LOCA EM (for initial DHRS cooldown events) 90 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Boron Precipitation Analysis

• Simplified, conservative mixing volume approach

• No time dependence

- Assume all boron in RCS concentrated in liquid volume in core and riser region

- Assume perfect mixing of boron in the core/riser region

• Maximum allowable boron concentration 1800 ppm (HZP conditions)

• For given liquid volume during transient, confirm core inlet temperature greater than boron precipitation temperature 91 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Evaluation

• LOCA events transition to ECCS cooling

- Full LOCA break spectrum evaluated, includes inadvertent opening of RRV or RVV

• Non-LOCA events transition from DHRS to ECCS cooling

- Events considered are SGTF, loss of FW flow, and a general evaluation of DHRS cooldown

• Transient calculated for 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after ECCS actuation to capture minimum RPV level and long term level recovery

- LTC phase begins after ECCS cooling is established, however transient history can affect mid-term level response, and the initiating event phase is also modeled

• State point analysis calculates final conditions after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

- Decay heat set to constant value at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the model is run until temperatures converge

• Injection line break identified as overall limiting initiating event 92 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Evaluated Conditions Minimum Level Minimum Temperature Maximum Temperature Reactor Power (%) 102 13-102 102 Decay Heat (multiplier) 1.2 (LOCA) 1.0 (nonLOCA) 0.8 1.2 (LOCA) 1.0 (nonLOCA)

RCS Avg. T. (°F) 555 535 555 RCS P. (psia) 1780 1780 1920 PZR Level (%) 52 68 52 Pool T. (°F) 65 65 210 Pool Level (ft) 69 69 55 Non-condensable Gas (lbm) 0 0 ~131 ECCS Capacity (area and Cv) minimum maximum minimum Expansion Factor (Y) 0.7 1 0.7 Single Failures RVV/RRV none RVV/RRV DHRS Enabled No Yes No

• Minimum level confirms core remains covered

• Minimum temperature confirms boron precipitation precluded

• Maximum temperature confirms acceptably low cladding temperature 93 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LTC Results - Injection Line Minimum Level

• Minimum level case maximizes RPV energy and minimizes CNV energy 100% break ECCS

• Higher pressure difference actuates between the RPV and CNV increases coolant accumulation in the CNV 5% break ECCS which reduces level actuates

• Minimum level during LTC phase occurs 3-5 hours after ECCS actuation

• Long term, level recovers towards equilibrium as the pressure difference between the RPV and CNV is reduced with decreasing decay heat 94 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LTC Results - Injection Line Minimum Level Core inlet temperature Max cladding temperature Riser collapsed level RCS pressure above top of active fuel 95 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LTC Results - Injection Line Minimum Temperature

• Minimum temperature case minimizes initial RCS energy and decay heat while maximizing heat transfer to the reactor pool

• Core temperature used as input to boron precipitation analysis 96 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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LTC Results - Injection Line Maximum Temperature

• Maximum temperature Max clad temperature case maximizes initial RCS energy and decay heat while minimizing heat transfer to the reactor pool

• Results demonstrate clad temperature follows a decreasing long term trend

• Long-term level remains at Riser collapsed level equilibrium condition above top of active fuel 97 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Results

• All maximum temperature cases showed decreasing cladding temperature, with final clad temperature remaining well below operating temperature at full power

• All minimum temperature cases showed margin to boron precipitation

• All minimum level cases showed the core remained covered at all times during the LTC phase, and that boron precipitation is precluded during the time of minimum level when boron concentration is maximized

- The limiting level case is a 100% injection line break, with minimum level of 2.8 feet above the core occurring approximately 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after ECCS actuation 98 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary ACRS Subcommittee Presentation Loss of Shutdown Margin June 19, 2019 99 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Principle Design Criteria 27

• NuScale DCA includes an exemption request from GDC-27

- ECCS design does not include boron addition

- NPM design does align with precedent based compliance with GDC-27

• Principle Design Criteria 27

- Passive reactor GDC-27 equivalent

- Ensures the safety related reactivity control system is designed to achieve and maintain subcritical core

- Ensures fuel integrity for an extended overcooling in combination with a partial failure of reactivity system (stuck rod)

The reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

Following a postulated accident, the control rods shall be capable of holding the reactor core subcritical under cold conditions, without margin for stuck rods.

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Compliance with PDC-27

• Immediate shutdown margin is sufficient to protect RCPB and SAFDLs with appropriate margin for the worst rod stuck out of the core

• Cold shutdown is achieved with all control rods fully inserted

• Loss of Shutdown Margin Consequences Benign

- Worst case OCRP not challenging to SAFDLs

- Critical power level does not challenge DHRS or ECCS heat removal capabilities

• Probability of the combination of conditions that results in a loss of shutdown return to power with a single rod stuck out of the core is small 101 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Return to Power Mechanisms

• Moderator overcooling

- ECCS and DHRS designed to removed decay and residual heat

- Under cold reactor pool conditions DHRS or ECCS can cause a fairly rapid temperature decrease and increased moderation (reactivity insertion)

• Fission product decay

- Xenon decay causes a slow post shutdown reactivity insertion

• Boron redistribution

- Boiling/condensing systems cause boron redistribution

• Boron concentration in boiling region

• Boron dilution in condensing region

• Limited boron acid volatized and carried with vapor

-

Conclusion:

Boron redistribution during extended ECCS operation increases boron and SDM in the core 102 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Evaluation of Return to Power

• DHRS or ECCS passive heat removal can be characterized as having a de-energizing phase and equilibrium phase

- At higher RCS temperatures passive heat removal capabilities well exceed decay heat generation (de-energizing phase)

- At lower RCS temperature conditions, passive heat removal nearly matches decay heat (equilibrium phase)

• Under maximum cooldown conditions, the equilibrium phase is reached in about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for either DHRS or ECCS

• Xenon decay inserts reactivity from 12-72hrs at an average rate of

<0.02pcm/s

• WRSO criticality is achieved <200°F 103 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Return to Power Characterization

• Characterize the steady state RCS temperature a function of available decay heat for both DHRS and ECCS heat removal.

• Calculate the WRSO critical power level as a function of RCS temperature.

Conclusions:

- Loss of SDM achieved >40hrs with zero Boron in RCS after Xenon decay

- RCS temperature must be <200F

- ECCS cooling is more limiting

- Pool temperature >140F precludes WRSO criticality

- Simple pool boiling CHF analysis demonstrates large margin 104 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Return to Power Results 250 Covered Riser ECCS Actuated Uncovered Riser Average Coolant Temperature (F) 225 Critical Power 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> DH 200 175 150 0.25 0.50 0.75 1.00 1.25 1.50 Decay Heat (%RTP) 105 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Bounding Dynamic Return to Power Analysis

• Purpose

- Demonstrate no challenge to SAFDLs, RCPB or CNV

• Assumptions

- Extended DHRS cooldown event without AC or DC power

- Zero boron EOC core conditions are applied to HZP RCS operating conditions (MDC curve applied)

- Pool temperature minimized to maximize cooldown rate

- Xenon decay consider as a reactivity source without consideration of time dependent worth

• Evaluation Procedure

- Evaluate maximum core peak power response (power overshoot)

- Calculate local peaking with core physics codes

- Repeat system calculation with IAB release (ECCS actuation) assumed at time of power peak

- Evaluate transient MCHFR to confirm acceptability

- Confirm acceptable RCS and CNV pressure response 106 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Return to Power Analysis DHRS 107 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Return to Power Analysis ECCS 108 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Acronyms

• AOO - Anticipated Operational

• EPRI - Electric Power Research Institute Occurrences

• oF - degrees Fahrenheit

• FIV - Flow Induced Vibration

• ASME - American Society of Mechanical Engineers

• FSAR - Final Safety Analysis Report

• ASTM - American Society for Testing and

• ft - feet Materials

• FW - Feedwater

• BPVC - Boiler Pressure Vessel Code

• FWIV - Feedwater Isolation Valve

• CES - Containment Evacuation System

• HZP - Hot Zero Power

• CIV - Containment Isolation Valve

• ISI - Inservice Inspection

• CNV - Containment Vessel

• LOCA - Loss of Coolant Accident

• COL - Combined License

• LTOP - Low Temperature Overpressure

• CRDM - Control Rod Drive Mechanism Protection

• CVCS - Chemical and Volume Control

• MPS - Module Protection System System

• MSIV - Main Steam Isolation Valve

• DHRS - Decay Heat Removal System

• MSS - Main Steam System

• ECCS - Emergency Core Cooling System

• MWt - Megawatts thermal

• EFPY - Effective Full Power Years 109 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Acronyms

• NDE - Non-destructive Examination

• RCPB - Reactor Coolant Pressure Boundary

• NEI - Nuclear Energy Institute

• RCS - Reactor Coolant System

• NPM - NuScale Power Module

• RG - Regulatory Guide

• NPS - Nominal Pipe Size

• RPV - Reactor Pressure Vessel

• OD - Outside Diameter

• RSV - Reactor Safety Valve

• OE - Operations Experience

• RTNDT - Reference Temperature for Nil-

• psia - pounds per square inch absolute ductility Transition

• P-T - Pressure and Temperature

• RVV - Reactor Vent Valve

• PTS - Pressurized Thermal Shock

• SG - Steam Generator

• PWR - Pressurized Water Reactor

• TRV - Thermal Relief Valve

• PWSCC - Primary Water Stress-Corrosion

• TS - Technical Specifications Cracking

• TT - Thermally Treated

• PZR - Pressurizer

• UHS - Ultimate Heat Sink

• RCCWS - Reactor Component Cooling Water System

• USE - Upper Shelf Energy 110 PM-0619-65981 Revision: 0 Copyright 2019 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 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 111 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Backup Slides -

Provide to Members 112 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Temperature Case

• Initiated by CVCS Time (sec) Event injection line break 0 CVCS line break

• Loss of normal AC Loss of normal AC power power at event initiation FW/MS isolation

• Single failure of one 3 High CNV pressure:

RRV to open CNV isolation Reactor trip

• Low bias High CNV Level setpoint 952 ECCS actuation on high CNV level

• No DHRS operation credited 955 ECCS valve opening 978 Peak CNV temperature 526°F

• Maximum CNV wall temperature: ~2500 CNV pressure < 50% peak 526°F < 550°F design pressure temperature 113 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Temperature Case Containment pressure RPV Pressure Energy balance Break and ECCS flow 114 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Containment Response Limiting Temperature Case

• Maximum wall temperature results occur following ECCS actuation Max CNV Wall Temperature

• Long-term maximum temperature not reduced due to modeling conservatism of adiabatic wall boundary condition Max CNV Wall Temperature applied above pool Longer term level 115 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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Long Term Cooling Results

• State-point results after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Core Inlet Temperature Collapsed Riser Level Boron Precipitation Margin

(°F) (ft) (°F)

Case Description Transient State-point Transient State-point Transient State-point at 12.5 Hours at 72 Hours at 12.5 Hours at 72 Hours at 12.5 Hours at 72 Hours Maximum Temperature IL Break 292.8 270.4 8.9 9.1 208.9 187.8 Minimum Temperature IL 152.8 140.4 10.0 10.4 73.1 62.3 Break Minimum Level IL Break 165.3 154.5 7.3(1) 8.0 76.2 69.0 Maximum Temperature IL Break, 45 ft reactor pool - 280.3 - 9.2 - 197.6 level (1)

Minimum Temperature IL Break, 13% initial power (1) - 94.3 - 10.4 - 16.6 Minimum Temperature SGTF, 13% initial power (1) - 112.1 - 10.1 - 33.3 Minimum Temperature DHRS cooldown, 13% initial power - 116.8 - 10.4 - 39.2 (1)

(1) A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> transient simulation for these cases was not performed. Limiting conditions are only important at the end of the LTC phase at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

116 PM-0619-65981 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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