ML19210D649
| ML19210D649 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/2019 |
| From: | Michael Snodderly Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Snodderly, M, ACRS | |
| References | |
| Download: ML19210D649 (317) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 July 19, 2019 MEMORANDUM TO: ACRS Members FROM: Michael R. Snodderly, Senior Staff Engineer /RA/
Technical Support Branch, ACRS
SUBJECT:
CERTIFIED MINUTES OF THE MEETING OF THE NUSCALE SUBCOMMITTEE ON MARCH 20, 2019 The minutes for the subject meeting were certified on May 21, 2019, as the official record of the proceedings of that meeting. Copies of the certification letter and minutes are attached.
Attachment:
As stated cc w/o
Attachment:
L. Burkhart cc w/
Attachment:
ACRS Members
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 MEMORANDUM TO: Michael Snodderly, Senior Staff Engineer Technical Support Branch, ACRS FROM: Michael L. Corradini, Member /RA/
NuScale Subcommittee
SUBJECT:
CERTIFICATION OF THE MINUTES OF THE ACRS NUSCALE SUBCOMMITTEE MEETING ON MARCH 20, 2019, IN ROCKVILLE, MARYLAND I hereby certify, to the best of my knowledge and belief, that the minutes of the subject meeting are an accurate record of the proceedings for that meeting.
______R/A _____ _ _____________5/21/19 Michael L. Corradini, Member Date NuScale Subcommittee
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MINUTES OF THE ACRS NUSCALE SUBCOMMITTEE MEETING ON MARCH 20, 2019 The ACRS NuScale Subcommittee held a meeting on March 20, 2019 in TWFN 3D50, 11545 Rockville Pike, Rockville, Maryland. The meeting convened at 8:30 a.m. and adjourned at 12:29 p.m.
No written comments or requests for time to make oral statements were received from members of the public related to this meeting.
ATTENDEES ACRS Members Joy Rempe, Member Dennis Bley, Member Ron Ballinger, Member Charles Brown, Member Michael Corradini, Chairman Vesna Dimitrijevic, Member Walter Kirchner, Member Jose March-Leuba, Member Harold Ray, Member Gordon Skillman, Member Matt Sunseri, Member Steve Schultz, Consultant NRC Staff Mike Snodderly, Designated Federal Official Prosanta Chowdhury, NRO Amy DAgostino, RES Brian Green, NRR Amanda Marshall, NSIR Lauren Nist, NRR Maurin Scheetz, NRR David Desaulniers, NRR Chris Miller, NRR Marie Pohida, NRO Ian Jung, NRO Dabiw Ki, RES Dan Barss, NSIR Victoria Huckabay, NRR Jing Xing, RES Other Attendees Scott Harris, NuScale Doug Bowman, NuScale Ryan Flamand, NuScale Carrie Fosaaen, NuScale Steve Pope, NuScale Tim Tovar, NuScale
Nadja Joergensen, NuScale Power
SUMMARY
The purpose of the meeting was to review the staffs evaluation with open items of Chapter 9, Auxiliary Systems, and Chapter 16, Technical Specifications, to the NuScale design certification application. The meeting transcripts are attached and contain an accurate description of the matters discussed during the meeting. The presentation slides and handouts used during the meeting are attached to these transcripts.
SIGNIFICANT ISSUES Reference Pages Issue in Transcript
- 1. S. Harris of NuScale began the presentation. 6
- 2. Member Skillman and various Members asked about possible boron 8 stratification in the ultimate heat sink.
- 3. Member Rempe and various Members discussed the criticality evaluation and corrective actions taken by NuScale in response to errors 10 found with that evaluation.
- 4. Chairman Corradini asked about the Pool Leakage Detection System. 18
- 5. Member Bley asked about the design of the reactor building crane and 20 features to control its movement, including interlocks.
- 6. Various Members asked about the maximum lift heights by the reactor 24 building crane and the reactor building cranes design requirements.
- 7. Member Bley asked about the human system interface and the reactor building crane design. Chairman Corradini asked about an associated 32 ITAAC.
- 8. Various Members discussed the single failure proof aspects of the 38 crane.
- 9. Member Skillman asked how buoyancy of the module while moving in 42 the pool was considered.
- 10. S. Schultz discussed his concern of whether there is sufficient 47 guidance to implement the COL action items associated with the crane.
- 11. Member Rempe asked about the reactor component cooling water 51 system.
- 12. Member Skillman asked about the stainless steel lined reinforced 55 concrete reactor pool.
- 13. Member Rempe mentioned the open item concerning an exemption 58 request for the sampling system.
- 14. Member Skillman and various Members discussed various potential 63 failure modes for the chemical and volume control system (CVCS).
- 15. Member March-Leuba and various Members discussed the fire protection 82 system.
- 16. G. Tesfaye began the staffs presentation for Chapter 9. 87
- 17. Member Bley asked how they ensure sufficient boron mixing in the 93 reactor pool, including sampling.
- 18. Member Skillman asked about the ramifications of a torn liner and the 100 possibility of a common mode failure for the entire reactor pool.
- 19. Chairman Corradini asked about the General Design Criteria 33 exemption that relies on ECCS and containment isolation. He asked about opening containment to allow injection by CVCS. Member March-Leuba 108 asked about normal operations and the fact that reactor coolant levels will drop due to normal cool down and CVCS is required for injection and whether or not there is sufficient boron.
- 20. R. Snuggerud of NuScale began their Chapter 16 presentation. 117
- 21. Member Ray discussed how the refueling technical specifications 122 compare to those in operating plants.
- 22. Members Kirchner and March-Leuba asked for demonstrations of shutdown margin as part of the Chapter 4 presentation scheduled for April 129 17, 2019.
- 23. Member Sunseri asked about the definition of Mode Five. 136
- 24. Member Skillman asked about the technical specification limit for boron 144 in the reactor pools.
- 25. Member March-Leuba asked about mixing in the reactor pool. 146
- 26. C. Harbuck of the staff began his presentation of the evaluation of 151 Chapter 16.
- 27. Member Bley asked about technical specifications associated with 157 electrical systems.
- 28. Member Brown and the staff discuss the engineered safety feature 165 actuation system logic and verification.
- 29. Member Bley and the staff discussed the surveillance requirement 176 acceptance criteria for the valve leakage in the steam and feed systems.
- 30. Member Rempe asked for public comments. M. Lewis and S. Fields 180 made comments.
- 41. Member Rempe asked for comments from the Members. 182
- 27. Member Rempe adjourned the meeting. 188
ACTION ITEMS Reference Pages Action Items in Transcript
- 1. The staff or NuScale did not take any action items from the meeting. N/A Documents provided to the Subcommittee
- 1. U. S. Nuclear Regulatory Commission, NuScale Power, LLC, Design Certification Application - Safety Evaluation With Open Items for Chapter 9, Auxiliary Systems, Draft, January 3, 2019 (ML19055A001).
- 2. NuScale Power, Design Certification Application, Chapter 9, Auxiliary Systems, Revision 1, March 15, 2018 (ML18086A056).
- 3. U.S. Nuclear Regulatory Commission, SECY 94-084, Policy and Technical Issues Associated With the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs, March 28, 1994 (ML003708068).
- 4. U. S. Nuclear Regulatory Commission, NuScale Power, LLC, Design Certification Application - Safety Evaluation With Open Items for Chapter 16, Technical Specifications, Draft, February 21, 2019 (ML19050A439).
- 5. NuScale Power, Design Certification Application, Chapter 16, Technical Specifications, Revision 1, March 15, 2018 (ML18086A063).
- 6. NuScale Power, General Technical Specification, Design Certification Application Part 4, Volume, December 31, 2016 (ML17013A298).
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards NuScale Subcommittee Docket Number: (n/a)
Location: Rockville, Maryland Date: Wednesday, March 20, 2019 Work Order No.: NRC-0218 Pages 1-187 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 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5 (ACRS) 6 + + + + +
7 NuSCALE SUBCOMMITTEE 8 + + + + +
9 WEDNESDAY 10 MARCH 20, 2019 11 + + + + +
12 ROCKVILLE, MARYLAND 13 + + + + +
14 The Subcommittee met at the Nuclear 15 Regulatory Commission, Two White Flint North, Room 16 T3B50, 11545 Rockville Pike, at 8:30 a.m., Joy L.
17 Rempe and Gordon R. Skillman, Co-Chairs, presiding.
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2 1 COMMITTEE MEMBERS:
2 JOY L. REMPE, Co-Chair 3 GORDON R. SKILLMAN, Co-Chair 4 RONALD G. BALLINGER, Member 5 DENNIS BLEY, Member 6 CHARLES H. BROWN, JR., Member 7 MICHAEL L. CORRADINI, Member 8 VESNA B. DIMITRIJEVIC, Member 9 WALTER L. KIRCHNER, Member 10 JOSE MARCH-LEUBA, Member 11 HAROLD B. RAY, Member 12 PETER RICCARDELLA, Member 13 MATTHEW W. SUNSERI, Member 14 15 ACRS CONSULTANT:
16 STEPHEN SCHULTZ 17 18 DESIGNATED FEDERAL OFFICIAL:
19 MIKE SNODDERLY 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 Chapter 9, "Auxiliary Systems,"
4 NuScale Design Certification Application . . . . 10 5 Chapter 9, "Auxiliary Systems,"
6 Safety Evaluation with Open Items . . . . . . . . 87 7 Overview of Chapter 16, "Technical Specifications" 8 NuScale Design Certification Application . . . 117 9 Chapter 16, "Technical Specifications" 10 Safety Evaluation with Open Items . . . . . . . 151 11 Adjourn . . . . . . . . . . . . . . . . . . . . 187 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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4 1 P R O C E E D I N G S 2 8:30 a.m.
3 CO-CHAIR REMPE: Good morning, this 4 meeting will now come to order. This is a meeting of 5 NuScale Subcommittee of the Advisory Committee on 6 Reactor Safeguards. I'm Joy Rempe, and I'll be co-7 chairing the Subcommittee with Dick Skillman and Mike 8 Corradini.
9 Members in attendance today are Vesna 10 Dimitrijevic, Charlie Brown, let's see, I guess Ron 11 Ballinger, Jose -- sometimes known as Roland -- Jose 12 March-Leuba, Harold Ray, Dick Skillman, Mike 13 Corradini, Dennis Bley, and Walter Kirchner.
14 In addition, we're joined by our 15 consultant, Steven -- oh, excuse me. Matt Sunseri and 16 Pete Riccardella. I forgot my colleagues up here at 17 the front table. We also today joined by our 18 consultant, Steven Schultz. And we may be expecting 19 to be joined later today by Margaret Chu.
20 Mike Snodderly is the Designated Federal 21 Official for this meeting. And today the Subcommittee 22 will review the staff's evaluation of Chapter 9, 23 Auxiliary Systems, and Chapter 16, Technical 24 Specifications of the NuScale Design Certification 25 Application.
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5 1 We also expect to hear from members of the 2 NRC staff, as well as representatives from NuScale.
3 They'll be briefing our subcommittee.
4 The ACRS was established by statute and is 5 governed by the Federal Advisory Committees Act. The 6 rules for preparing today's meetings were announced in 7 the Federal Register on March 18, 2019. It will be an 8 open, and possibly then a closed, meeting, as we will, 9 may close the meeting after the open portion to 10 discuss proprietary material.
11 And presenters can defer questions that 12 should not be answered in the public session to that 13 time. And I'm going to ask for NuScale's help if some 14 of our questions during the open session do go beyond 15 what they should be and to just tell us you'd prefer 16 to wait.
17 No written statement or request for making 18 an oral statement to the Subcommittee has been 19 received by the public concerning this meeting. A 20 transcript of the meeting is being kept and will be 21 made available, as stated in the Federal Register 22 notice.
23 Therefore, we request that participants in 24 this meeting use the microphone that is located at the 25 front of the public seating area. And they need to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 first identify themselves, as well as speak with 2 sufficient clarity and volume so that they can be 3 readily heard when they do make comments.
4 We have a bridge line that's been 5 established for the public to listen to this meeting.
6 And to minimize disturbance, that line is being kept 7 in a listen-in only mode. To avoid disturbances, I 8 also want to request that attendees at this meeting 9 either turn off or mute their electronic devices.
10 And we're now going to proceed with this 11 meeting, and I'm going to call upon Scott Harris of 12 NuScale to start us off. Scott.
13 MR. HARRIS: Thank you. It's a pleasure 14 to be here today and present Chapter 9 on behalf of 15 NuScale. My name is Scott Harris, I'm a supervisor 16 for the Mechanical Systems Group.
17 Just to give background on myself, I 18 graduated from Kansas State University with a degree 19 in mechanical engineering. I started off my career as 20 a field engineer at Fort Calhoun Nuclear Station for 21 two years, and then transitioned to NuScale in 2012, 22 where I've been a member of the Mechanical Systems 23 Group ever since.
24 DR. NICHOL: My name is Corrie Nichol, I 25 have a PhD in mechanical engineering from Penn State.
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7 1 I spent 11 years as a research engineer at the Idaho 2 National Laboratory designing, among other things, 3 some of the remote handling equipment for the Yucca 4 Mountain Waste Package Welding Project. And now I'm 5 working for NuScale in the Remote Handling Group.
6 MS. FOSAAEN: Good morning, Carrie 7 Fosaaen, I'm a supervisor of licensing at NuScale 8 Power. I've been there for about three and a half 9 years. Involved in Chapter 9 ever since I joined the 10 company. I graduated from Perdue University in 2008 11 with a nuclear engineering degree, and then in 2010 12 with a masters in health physics.
13 MR. FIELDS: Good morning, I'm John 14 Fields. I'm a nuclear nomad. I've been all over the 15 industry for 35 years here, so I'm not going to go 16 into my degree and all that stuff. But all of that in 17 engineering and licensing.
18 MR. HARRIS: Just a high level review of 19 Chapter 9. It includes fuel storage handling, water 20 systems, HVAC, fire protection and other fire hazard 21 analysis.
22 DR. NICHOL: So we're going to start off 23 by talking about spent fuel storage and handling, then 24 we'll talk about new fuel storage, spent fuel pool 25 cleanup and cooling system, the fuel handling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 equipment, and the overhead heavy load handling.
2 So to start off the spent fuel racks, the 3 design of the NuScale spent fuel racks prevents 4 criticality without regard to loading patterns or 5 zones. The spent fuel pool itself is connected to the 6 ultimate heat sink, the operating pool of the 7 facility, which essentially means the volume is large.
8 The reason that's important is it prevents undetected 9 boron dilution in the pool.
10 Fuel management is controlled by 11 procedures, fuel storage and that sort of thing.
12 There are a couple tech specs related to fuel storage.
13 Tech spec 4.3 is fuel storage, tech spec 5.5.12 talks 14 about the neutron absorber monitoring program.
15 And there is a combined license item 16 related to this regarding the programs for fuel 17 movement and the programs and procedures that govern 18 that.
19 CO-CHAIR SKILLMAN: I'd like to ask a 20 question, please. In this seven million gallon pool, 21 you've got a number of fluid systems that feed to it.
22 What ensures homogeneity in the boron concentration in 23 that pool among the 12 module spaces, the refueling 24 space, and the maintenance bay?
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9 1 areas, and you've got a number of systems that touch 2 that water. What makes sure the boron concentration, 3 and I'm going to assume it's approximately 2,000 ppm 4 throughout, what makes sure it's homogeneous?
5 MR. HARRIS: So I can speak to that. So 6 there's two --
7 MEMBER CORRADINI: If I might just 8 interject. If we're saying things that are 9 proprietary, stop us.
10 MR. HARRIS: Okay, understood.
11 MEMBER CORRADINI: We can go and hold for 12 the, but sorry.
13 MR. HARRIS: Okay, so there's two main 14 systems that service the ultimate heat sink as far as 15 cooling is concerned, the spent fuel pool cooling 16 system and the reactor pool cooling system. Between 17 those two systems, they have a number of suction and 18 discharge points.
19 Suction points are taken off the spent 20 fuel pool and refueling pool then discharged back into 21 each individual operating bay. So that helps ensure 22 homogeneity.
23 CO-CHAIR SKILLMAN: Thank you.
24 DR. NICHOL: One of the unique features of 25 our plant, we don't have specific new fuel storage.
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10 1 The new fuel is brought in and stored in the spent 2 fuel racks. The racks, the pool liner, and the pool 3 itself are seismic category I, and they are protected 4 from non-seismic category I structures, systems, and 5 components.
6 There are a couple combined license items 7 related to this. The first one talks about the dry 8 cask. We haven't specified that, that's something 9 that the applicant will choose which model of dry 10 cask, and that will then be spelled out in a combined 11 license item.
12 There's also a combined license item for 13 the site-specific rack designs.
14 CO-CHAIR REMPE: So before you go on to 15 the next slide, on the prior slide, I know you've put 16 off to the COL applicant to do the criticality 17 evaluation, but in the staff's review, they noted that 18 some of the information you prepared that there were 19 some errors in some dimensions in that documentation, 20 which you've corrected.
21 But I guess what I'm asking about is 22 something that I saw a couple other places in the 23 staff's SER, where there were some errors that I would 24 have thought a QA program should have detected.
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11 1 staff identified, did you stop and think that there 2 might some sort of underlying root cause that might be 3 leading to such errors, and have you addressed those 4 errors with some sort of fixes to your QA program to 5 make sure that similar errors aren't existing that the 6 staff might have missed?
7 MR. HARRIS: I'm not familiar with the 8 particular errors, but I would say if they are 9 identified, they would have been put into our CAP 10 program, and depending on the severity, no matter the 11 severity, would have had a condition evaluation 12 performed.
13 CO-CHAIR REMPE: Okay, so as we go through 14 today's presentation I'll bring up some other cases 15 where the staff identified some things that I thought 16 that the QA program would have detected or should 17 detect and we'll see if you can, till I kind of get an 18 idea that you really have thought about hey, we don't 19 have those kind of things continuing to occur. Okay?
20 MR. HARRIS: Okay.
21 MS. FOSAAEN: And if I may just correct, 22 so we did present a criticality analysis in there.
23 The COL item is to confirm that it's still applicable 24 with the site-specific conditions.
25 CO-CHAIR REMPE: And in that analysis the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 staff found some errors, right?
2 MS. FOSAAEN: Yes.
3 CO-CHAIR REMPE: And so you understand 4 where I'm coming from, and yeah, okay, we all make 5 mistakes, but are there other mistakes? And should 6 the QA program have detected that?
7 MS. FOSAAEN: Yeah, I understand the 8 question. And like Scott said, we did enter all of 9 that into our corrective action system for evaluation, 10 but I'm not familiar with the outcome of those actions 11 and what actions we took, other than correcting the 12 errors identified.
13 CO-CHAIR REMPE: Okay, thank you.
14 MR. HARRIS: And since we're talking about 15 the --
16 MEMBER BLEY: Before you go ahead, I'd 17 like to follow up on Dick's question. Put on that 18 next slide, please. Yeah, thanks. There are a lot of 19 separate areas here. It makes sense the answer you 20 gave about why it ought to be mixed. But sometimes 21 things that make sense don't turn out to be right.
22 Where is the concentration monitored and 23 where will it be sampled and kind of how often? I 24 mean, have you played with this? Do we know for sure 25 that it's well mixed?
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13 1 MR. HARRIS: Well, I can't --
2 MEMBER BLEY: Or will we know for sure 3 when the plants operate?
4 MR. HARRIS: I can't speak to where it's 5 sampled specifically. I imagine it's sampled on 6 several different areas, in the spent fuel pool, 7 refuel pool, and each individual operating bay.
8 MEMBER BLEY: I didn't see where that's 9 spelled out. That's not part of the design. That's 10 going to be left to the --
11 MR. HARRIS: That'll be part of an 12 operating procedure.
13 MEMBER BLEY: Operating folks to figure 14 out where to sample.
15 CO-CHAIR REMPE: And how will they --
16 MEMBER CORRADINI: Is that an item that's 17 identified for the operator?
18 PARTICIPANT: No.
19 MR. FIELDS: Each applicant will have to 20 develop their own operating and maintenance 21 procedures. That's a COL item.
22 MEMBER BLEY: But there's nothing in COL 23 items that tells them, that speaks to this issue of --
24 MR. FIELDS: Not specifically to that, 25 your question, sir, no.
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14 1 MEMBER MARCH-LEUBA: Generally if you're 2 familiar with the criticality of the spent fuel pool 3 for what you said. Do we take credit for the boron in 4 the water, or are we not taking credit for it? For 5 the spent fuel pool criticality calculation. Did you 6 rely exclusively on the solid racks, or did you take 7 credit for the soluble boron?
8 DR. NICHOL: So I believe we did take 9 credit for the boron concentration.
10 MEMBER MARCH-LEUBA: In the water?
11 DR. NICHOL: For criticality control.
12 MEMBER MARCH-LEUBA: Okay, then these 13 questions are really relevant. It should not be left 14 over to the QA or the staff.
15 MS. FOSAAEN: Actually, we have someone in 16 the audience we're going to have help us out here real 17 quick with this question, if you're all right with 18 that.
19 MR. SHAVER: Good morning, my name is Mark 20 Shaver, I'm a supervisor of the Radiological 21 Engineering Group at NuScale Power. In our crit 22 safety analysis, the spent fuel pool is 1800 ppm of 23 boron, but we only take credit in the criticality 24 safety analysis for 800 ppm boron. And so that gives 25 us 1000 ppm margin.
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15 1 And also, the limits we work to is K 2 effective value of .95 at the 100 ppm, and then under 3 1.0 without boron. So we do it both with and without 4 boron, but the boron we do credit is conservative.
5 MEMBER MARCH-LEUBA: Do you take credit 6 for burnup, burnup credit?
7 MR. SHAVER: No, we do not take any burnup 8 credit.
9 MEMBER BLEY: All of this sounds pretty 10 good, but, you know, I've spent a little time around 11 the chemical process industry, and you sometimes see 12 very surprising results in tanks that are assumed to 13 be well mixed when you sample in different places.
14 MEMBER MARCH-LEUBA: Especially when you 15 have thermogradients. The spent fuel pool is likely 16 to need a hotter source -- to be insulated. And so 17 whenever you have thermogradients in such a large 18 pool, you have to worry about those things.
19 MR. HARRIS: Okay, so as we're talking 20 about the spent fuel pool, I'll just lay out the 21 ultimate heat sink configurations. So the ultimate 22 heat sink consists of the spent fuel pool, our 23 refueling pool, and the reactor pool. And they all 24 freely translate with each other.
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16 1 use to put our module in for maintenance operations 2 during refueling.
3 DR. SCHULTZ: Excuse me, a question. What 4 latitude are you providing to the COL applicant with 5 regard to the spent fuel pool racks, and what are they 6 working to achieve when they do their own spent fuel 7 pool criticality calculations?
8 MR. HARRIS: Do you want to speak to that?
9 DR. NICHOL: Sure, so if you actually go 10 back one slide, this is spelled out. The COL 11 applicant is responsible for submitting vendor-12 specific design that includes the seismic analysis, 13 the criticality analysis, and those sorts of things.
14 Does that?
15 DR. SCHULTZ: So they have the opportunity 16 to change rack design different than what you're 17 proposing at this point?
18 MS. FOSAAEN: Yes, they could change 19 design, but that would require a reevaluation by the 20 staff at the COLA stage. The COL item as represented 21 here is really intended to look at the seismic 22 structural aspect and ensure that the site seismic 23 characteristics are, you know, adequately evaluated.
24 And as part of that, we recognize that it could affect 25 the other characteristics.
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17 1 So this COL applicant item was really 2 intended to be relative to the seismic aspect. And 3 then if there was a change that wasn't bounded by what 4 was already performed, we recognized they would need 5 to look at these other things, and it could result in 6 changes that would need to be evaluated.
7 And the standard review plan 384 spells 8 out the criteria that they have to do this re-analysis 9 for. So the applicant would need to use that guidance 10 to ensure that they're appropriately selecting racks 11 if they were changing design.
12 DR. SCHULTZ: And is NuScale also 13 providing guidance based upon your analysis and your 14 approach that would help the COL applicant to 15 determine what they need to accomplish in those 16 analyses?
17 MS. FOSAAEN: So we've spelled out certain 18 criteria in the application that would need to be 19 confirmed against, yes.
20 DR. SCHULTZ: Okay, thank you. So this 21 really starts with things like external events that 22 would not be considered in the normal design practice, 23 based upon what you've assumed. And you've got racks 24 that are seismic one and you've got criteria that 25 match up with that.
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18 1 MS. FOSAAEN: Yes.
2 DR. SCHULTZ: Thank you.
3 MR. HARRIS: So as previously mentioned, 4 there's two systems which provide cooling for the 5 ultimate heat sink, that's the spent fuel pool cooling 6 and the reactor pool cooling systems. Both these 7 systems are not safety-related.
8 In addition to cooling, they also maintain 9 ultimate heat sink water level and provide a means for 10 chemistry control and also provide reactor pool 11 temperature information signals for post-exit 12 monitoring.
13 In addition to the ultimate heat sink 14 cooling systems, we have the pool cleanup system.
15 Again, non-safety related. This is specifically used 16 for removing impurities to reduce dose rates and 17 maintain both chemistry and clarity.
18 We also have the pool surge control 19 system, which is our system used to drain the dry dock 20 during refueling operations.
21 The pool leakage detection system is used 22 to monitor any leakage from the pool liner. It uses 23 a series of channels to flow any leakage to our sumps 24 within the reactor building for operator evaluation.
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19 1 that's situated, does that determine the location or 2 just gross loss?
3 MR. HARRIS: So you can isolate specific 4 channels to help identify the approximate area within 5 the reactor pool. And then from there you can use 6 borescopes to go up and down the channels and identify 7 any wet spots.
8 And those are my last two points there.
9 Fuel handling.
10 DR. NICHOL: So the fuel handling 11 equipment, essentially we're talking about equipment 12 that handles receipt of new fuel, refueling 13 operations, and loading of spent fuel into a dry cask.
14 Some of the design considerations, these pieces of 15 equipment are designed to the appropriate design 16 standards, including ASME and ANSI design standards.
17 Essentially the goal is to reduce the risk 18 of a drop or a criticality event or exposure to 19 radiation. The equipment is also designed to support 20 inspection of the equipment itself.
21 This is an illustration of the spent fuel 22 pool and the fuel handling equipment. You can see 23 starting on the left there is the new fuel jib crane.
24 The area of hook coverage with a new fuel jib crane is 25 illustrated there with a wedge cut out of the side of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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20 1 it that essentially prevents it from, by interlock, 2 prevents it from carrying loads over the spent fuel 3 racks.
4 MEMBER BLEY: Have you designed or 5 specified what kind of interlock that's going to be?
6 I mean, some interlocks are like a micro switch or 7 something sensing where it's moved. And if you got a 8 bad signal, it'll keep going. Other ones are actual 9 physical interlocks, so you just can't move.
10 DR. NICHOL: So it is not designed.
11 MEMBER BLEY: Is it specified? If not, 12 when will it be specified?
13 DR. NICHOL: That's a good question.
14 MEMBER BLEY: Thank you.
15 CO-CHAIR REMPE: So some background, for 16 some background information, the fact that you put 17 this interlock in was in result to the staff's review, 18 right?
19 MS. FOSAAEN: That's correct.
20 DR. NICHOL: So I would think at the 21 latest certainly at the time that the equipment is 22 designed it will be specified. It's a specification 23 to the equipment.
24 CO-CHAIR SKILLMAN: What is the 25 distinction on the medium left of this image and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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21 1 far right in this image where there is the green 2 crosshatching, and on the far right, the blue 3 crosshatching, and the combined crosshatching right in 4 the middle? What is the left green crosshatching 5 communicating to us?
6 DR. NICHOL: So there's a key at the 7 bottom of the image. The green crosshatching is the 8 coverage area that the fuel handling machine mast can 9 access. You can see that covers all of the storage 10 locations in the spent fuel racks, as well as a small 11 portion of the refuel pool, which gives it access to 12 the fuel when it's in the lower reactor pressure 13 vessel during refueling.
14 The blue crosshatching is the auxiliary 15 hoist, which is a piece of equipment on the fuel 16 handling machine. And that's used for various 17 operations, including handling control rod assemblies, 18 that sort of thing. So it's physically located in a 19 different spot on the fuel handling machine, so its 20 coverage area is different.
21 CO-CHAIR SKILLMAN: Is there a safety-22 related connotation to this crosshatching image?
23 DR. NICHOL: No.
24 CO-CHAIR SKILLMAN: No. This is just, 25 hey, this is where the equipment can reach, this is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 1 where the bridge and trolley are able to operate.
2 DR. NICHOL: Right.
3 CO-CHAIR SKILLMAN: It's simply that?
4 DR. NICHOL: Right.
5 CO-CHAIR SKILLMAN: Okay, thank you.
6 DR. NICHOL: The only other piece of 7 equipment that I haven't mentioned is the new fuel 8 elevator. You can see that illustrated there. The 9 new fuel jib crane of course brings new fuel 10 assemblies to the elevator. They're brought down to 11 the level that the fuel handling machine can them 12 access them and place them in racks.
13 The COL items associated with the fuel 14 handling equipment, there are procedures that need to 15 be developed related to transferring fuel to a spent 16 fuel cask, and also one regarding the inspection of 17 the, and periodic testing of the fuel handling 18 equipment itself.
19 The overhead heavy load handling system.
20 The main components of this system include the reactor 21 building crane, the nuclear power module, the NuScale 22 power module lifting fixture, which is the physical 23 structural elements that attach to the lift tabs. And 24 then the diagonal braces that attach to the module 25 itself.
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23 1 There's a module lifting adapter that 2 interfaces essentially between the reactor building 3 crane and the NuScale power module lifting fixture.
4 And there's a wet hoist, which is used for any of the 5 lifting operations that take place under water.
6 The reactor building crane specifically is 7 designed to be a single failure-proof piece of 8 equipment. It's designed to ASME NOG-1 Type 1 9 standards, that's the highest NOG-1 qualification, 10 highest NOG-1 design standard.
11 The movement of that system is controlled 12 via interlocks, and the path and the maximum lift 13 height are limited to prevent or to maintain shielding 14 when moving reactor modules. And there's a 15 communication system between the operator of the 16 reactor building crane and the Control Room.
17 MEMBER CORRADINI: What is the lift height 18 normally?
19 DR. NICHOL: The lift height of?
20 MEMBER BLEY: In other words maximum.
21 MEMBER CORRADINI: But normal, well, I 22 guess maybe both. When you move a module, it's how 23 far off the floor of the ultimate heat sink floor.
24 DR. NICHOL: A foot off the floor.
25 MEMBER CORRADINI: And then I guess Dennis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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24 1 is asking the maximum height?
2 DR. NICHOL: Correct.
3 MEMBER CORRADINI: What is that?
4 DR. NICHOL: The maximum height of the, so 5 I have the numbers in my head. The maximum hook 6 height is --
7 MEMBER CORRADINI: Assuming you can tell 8 us.
9 DR. NICHOL: Yeah, it's 145 foot off the, 10 I mean, that's the site elevation, which puts the --
11 MEMBER BLEY: You said there was a limiter 12 on how high it can lift. Where is that limiter set?
13 DR. NICHOL: Oh, so the lift height limit 14 is going to be based on the load that's being picked 15 up. So when a reactor module is being handled by the 16 crane, there'll be a limit imposed on how high it's 17 lifted.
18 MEMBER BLEY: So this is something an 19 operator will have to manually set.
20 DR. NICHOL: No, it will be handled by an 21 automatic system.
22 MEMBER BLEY: How does the automatic 23 system know what it's going to lift?
24 DR. NICHOL: The system looks at the load 25 that the crane is lifting and limits the maximum lift NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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25 1 height.
2 MEMBER BLEY: By, just by the weight.
3 CO-CHAIR SKILLMAN: What is the quality 4 level of that control system?
5 DR. NICHOL: That's a good question.
6 CO-CHAIR SKILLMAN: Yes, it is.
7 DR. NICHOL: So --
8 CO-CHAIR SKILLMAN: Okay, so I've got a 9 740-ton module I'm hoisting. I'm at 11 inches, I pass 10 through 12, now I'm at 13, now I'm at 14. You have no 11 module drop accident analyzed, and this control system 12 has not stopped hoist. What do you do?
13 DR. NICHOL: So as an operator what do you 14 do?
15 CO-CHAIR SKILLMAN: Yeah.
16 DR. NICHOL: So the operator will be 17 informed of what is appropriate --
18 CO-CHAIR SKILLMAN: Now I'm at 14 inches.
19 DR. NICHOL: And the operator has the 20 opportunity to hit a stop button.
21 CO-CHAIR SKILLMAN: How does he know to do 22 that?
23 DR. NICHOL: He monitors the operations.
24 CO-CHAIR SKILLMAN: Okay. But he had an 25 automatic system that he was or she was depending on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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26 1 to stop the hoist. So what's the quality of the 2 control system? Let's just leave it at that.
3 DR. NICHOL: Let me get back to you on 4 that.
5 CO-CHAIR SKILLMAN: Get back to us on 6 that.
7 DR. NICHOL: Back to all of you.
8 CO-CHAIR SKILLMAN: Thank you.
9 MEMBER MARCH-LEUBA: I'd like to follow up 10 on that because it's in the news all last week, all 11 these computer controls with single switches cause a 12 lot of problems. It really begs the question of what 13 else we have single input computer controlled for an 14 operation that could become dangerous if it runs away.
15 And that is not the responsibility of ACRS 16 or the responsibility of the staff, it's your 17 responsibility to find it. And we'll review what 18 you've done. You cannot rely on the staff or ACRS to 19 find all those problems.
20 It was on the news this morning that 21 another plane, well, that like here, the condition of 22 the crash was like there was another pilot. A day 23 before the crash, that he went to the crash the day 24 before because he knew how to handle it. And then the 25 day after, that guy was not flying, so it crashed.
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27 1 So having a crane handling one of these 2 modules, I'm saying, well, the operator will push the 3 red button.
4 MEMBER BALLINGER: So you say that it's 5 load control, so it's a set round number 700-ton 6 module. Most load cells 0.1% error, so that's about 7 7 tons. So what's the next heaviest device that the 8 crane would have to lift?
9 Is it way beyond the uncertainty of the 10 load system? In other words, is there something that 11 close, that's within the 7 tons error that you could 12 lift and make a mistake?
13 DR. NICHOL: No, no. I think the next 14 lightest component is well below 400 tons.
15 MEMBER BALLINGER: Okay.
16 MEMBER CORRADINI: I was going to just, 17 can I just generalize the question and then you can 18 think about it. Operational procedures of the crew so 19 that they know what's going on, even though they may 20 or may not believe the computer control I think is 21 kind of where we're going.
22 And so I'm curious, is there a, maybe 23 that's not in this chapter, but I think it's related 24 to the module. That's where we were a month ago, 25 talking to another group of the NuScale folks about NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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28 1 these topics. So I think that's the sense of what 2 you're hearing from some of the members.
3 MS. FOSAAEN: Right, and we do have a COL 4 item that identifies the need to develop procedures 5 and operator training specifically for the crane in 6 915.
7 And if I may follow up on the question 8 related to instrumentation, our software control will 9 follow Reg Guide 1.168, and that's described in 7.2.
10 So the crane has been identified, the crane operating 11 software has been identified as needing to follow that 12 information.
13 MEMBER RAY: Well, there's something I 14 guess was implicit in something Dick said earlier.
15 We've done a lot of heavy lifting in this business, 16 and it's not a unique or first-time event. But when 17 we did heavy lifting in the past, up until now as far 18 as I know, that could result in core damage, we always 19 off-loaded the core first. The exception of course is 20 the removal of a head.
21 But this is different, because you are 22 moving the core, basically. And so the questions 23 about the quality of the specified requirements, let's 24 put it that way, what are we certifying in this design 25 is the question. If we're just certifying that oh, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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29 1 you have to come up with a crane that'll move a heavy 2 load reliably, but that's the responsibility of each 3 COL holder to do that individually, that's one thing.
4 But if there are more requirements that 5 are placed in the design certification for this part 6 of the system, we'd like to know what it is 7 specifically.
8 DR. NICHOL: So I would actually like to 9 speak to that. So in a standard operator system where 10 the, you mentioned the core is offloaded before heavy 11 loads are handled over the top of the core, what we do 12 is different, and we recognize that. And we have 13 spent a good deal of time analyzing the potential for 14 core damage based on load drop scenarios.
15 Seen, though, from another viewpoint, it 16 is not uncommon to move containers containing spent 17 fuel in an operating reactor, in the operating reactor 18 fleet. So seen from that angle, essentially what 19 we're moving is a very large spent fuel container.
20 As I said, that doesn't mean that we're 21 dismissing the importance of this, and we have done a 22 lot of analysis looking at potential core damage 23 frequency and those sorts of things. But we are 24 essentially moving a large container of spent fuel.
25 MEMBER RAY: Well, I would disagree, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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30 1 this is not the place to debate that. It's not going 2 to sit there for five years before you load the spent 3 fuel cask. There are many reasons why I wouldn't 4 agree, but this is not the place to have that debate.
5 The question I'm trying to ask very 6 clearly is are you just going to tell the COL holder 7 they need to address this, or are you going to, is the 8 design certification going to provide specific 9 requirements that must be met by the COL holder. And 10 that extends over many, many issues from quality 11 requirements for the control system to operator 12 requirements and so on.
13 Because that has to do then with what is 14 the scope of the design certification. Okay?
15 CO-CHAIR SKILLMAN: I do want to go on 16 record and push this a little bit, because I find your 17 argument flawed. I concur, an 840-ton cask or an 850-18 or 870-ton module can be safely moved. Happens all 19 the time, big loads, shipyards, other parts of 20 industry move large loads safely.
21 What's different is you are moving a, the 22 current design moves a core that has abundant decay 23 heat generation, and it is being moved adjacent to 24 potentially 11 live cores. That's different.
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31 1 Chapter 19, module drop accident is not considered.
2 And so I find the combination of your response, it's 3 just a heavy load, kind of like a cask, combined with 4 the absence of a module drop accident, to be 5 problematic. I think NuScale has to own the notion 6 they're moving a very heavy load adjacent to live 7 cores.
8 Let me go one step further. I was 9 involved in moving a heavy load repeatedly over a dead 10 live core. And at TMI-2, we dropped equipment into 11 the open reactor vessel in which there was destroyed 12 fuel. So to the very best of our intention to have 13 the single failure-proof crane, guess what, a load got 14 away from us. It simply happens.
15 And so I guess I've got that orientation 16 that I'll never let go of that a crane accident can be 17 a very big problem. And the people at Arkansas 18 Nuclear 1 will verify that. This is a heavy load, and 19 it can have heavy consequences, and it's adjacent to 20 other live cores.
21 MS. FOSAAEN: If I may, we did not 22 evaluate it in Chapter 15 because of the guidance that 23 says if we design to single failure-proof. However, 24 we did evaluate model drop within our action with 25 nearby modules in Chapter 19. And there is an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 1 analysis that does look at that under PRA, with the 2 consequences.
3 MEMBER BLEY: There's one thing that kind 4 of bothers me, and I don't expect you to answer this 5 now. And I'm kind of guarding the staff, I'm going to 6 ask them as well. It came up in one of my earlier 7 questions about exactly what kind of limit device 8 would be on the jib crane.
9 In Chapter 18, NuScale responded to the 10 staff that details about the human system interface 11 and essentially potential operator errors is being 12 left to the designer of the crane, who will design to 13 your specification. I assume some of these other 14 issues are also in that ballpark. I don't know when 15 they get looked at, and I'm going to ask the staff 16 that when they come up.
17 If you've negotiated this and know about 18 that, that'd be great. But some of these issues are 19 potentially important, and putting it off on the 20 designer of the crane at the time the crane's going to 21 be built, I'm not sure where that fits in the review 22 process by the NRC. And will it be reviewed?
23 Or is it just, I mean, they're not even 24 right now reviewing your specification because they 25 don't think you have it yet.
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33 1 MEMBER CORRADINI: Can I ask Dennis's 2 question a little bit differently? If it's as he 3 said, is there an ITAAC then that has to come and be 4 cleared, such that the specifications have been 5 designated, met, and then checked by the staff? And 6 we can hold off and ask the staff this, but I'm just 7 kind of curious from your standpoint, because I think 8 he makes a valid point.
9 MS. FOSAAEN: So I can check, but I'm 10 quite sure that we actually do have an ITAAC for the 11 crane specifically. I don't recall off the top of my 12 head the specifics, but we can get back to that.
13 MEMBER CORRADINI: We'll come back.
14 CO-CHAIR REMPE: But I think Harold's 15 point was that did you at least give guidance on what 16 would be done? Not just have an ITAAC, but is there 17 some high level guidance of what is expected is where 18 I am at.
19 MS. FOSAAEN: We do have guidance in 20 Chapter 14 that helps understand what's required to 21 close the ITAAC. So there is guidance in 14 that 22 specifies what should be done to help resolve the 23 ITAAC.
24 CO-CHAIR REMPE: Thank you.
25 MEMBER BALLINGER: I couldn't find this in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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34 1 Chapter 9, but maybe I should look in other chapters.
2 And I'm quite familiar with the interaction between 3 electromechanical and electrohydraulic systems, 4 computer controls, and what happens when things can go 5 wrong and all of a sudden you get very high 6 temperature steam out into the laboratory.
7 And the way we solved the problem, invent 8 stuff. The way we solved the problem was to put a 9 mechanical limiter, mechanical limiter, in the system 10 that made it virtually impossible. The crane 11 operator, the computer guy could be stoned and try to 12 move something, and it hits a mechanical stop and it 13 just can't go any further.
14 Is there something like that with this 15 crane? I couldn't find it, but it's so easy to do.
16 MEMBER BLEY: They told me earlier that's 17 not specified.
18 MEMBER BALLINGER: I mean, that solves the 19 problem, or largely solves the problem.
20 DR. NICHOL: So there are no physical, I'm 21 trying to understand, you're saying, you're asking if 22 there's like a physical?
23 MEMBER BALLINGER: Module a foot, right.
24 But then somebody asked how high could you lift it.
25 You said 100 --
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35 1 DR. NICHOL: No, no, that's the top of 2 the.
3 MEMBER BALLINGER: Yeah, but how high can 4 you, you know, what limits how high you can actually 5 go?
6 MEMBER BLEY: Ron's point, you know, this 7 could be done by software. Looking at ovations and 8 positions, it could be done by something like limit 9 switches, electrical devices, or it could be done by 10 physical devices --
11 PARTICIPANT: We had all three.
12 MEMBER BLEY: To keep it from moving and 13 one works a whole lot more often than the other two.
14 Is successful, the whole.
15 MEMBER BALLINGER: And it's simple.
16 MEMBER KIRCHNER: So what is the range of 17 operation of the crane for other purposes? Isn't it 18 used for shielding?
19 DR. NICHOL: Maybe we can get into that 20 in, I have a figure coming up. I don't think it's the 21 next one, but we'll get back to the range of operation 22 of the crane. And that may answer some of the other 23 questions about limits of where the crane can operate.
24 MEMBER RICCARDELLA: Before we go on, 25 Corrie, you mentioned an ASME standard that I'm not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 familiar with. Could you tell us a little about that?
2 DR. NICHOL: I believe that was NOG-1, 3 nuclear --
4 MEMBER RICCARDELLA: NOG, N-O-G?
5 DR. NICHOL: Yeah, nuclear overhead 6 gantry, N-O-G-1.
7 MEMBER RICCARDELLA: Oh, okay.
8 MEMBER BALLINGER: But none of those 9 documents mention a live core. 550 whatever it is, is 10 two NUREGs and everything, there's nothing in there 11 that says this is for a live core. So that's what 12 makes it different, and I presume that your PRA, 13 you're getting around it by having it a very low 14 probability event in the PRA.
15 CO-CHAIR SKILLMAN: Well, no, the way they 16 get around it is they communicate this is a single 17 failure-proof crane and it won't happen.
18 MEMBER BALLINGER: Right, right, but the 19 NUREGs don't have anything in there that says if 20 you're lifting a live core, you got to do something 21 different. It's just --
22 MEMBER BLEY: Lifting over a live, 23 adjacent to it, lifting over a live core.
24 DR. NICHOL: If I could clarify, though, 25 that we do, in Chapter 19, we do consider module drop NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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37 1 accidents that would include core damage. We do look 2 at that.
3 MEMBER MARCH-LEUBA: Yeah, and I'm looking 4 at it. I'm not sure if it's proprietary or not, so I 5 won't mention numbers, but they are ludicrous.
6 MEMBER DIMITRIJEVIC: Well, in your 7 Chapter 19, that's the most important event from the 8 point of core damage. How you go around is that it's 9 crap and it's not going to be released and it's not 10 considered as a safety measure. And so there is not 11 really detailed analysis otherwise, as you tried to 12 mention.
13 This is an estimate of probability of this 14 happening, which is the low in ten to minus eight.
15 But it's still the most important event because this 16 is very safe plan. So this mostly, it seems this is 17 happening in the big pool, in the scrapping there is 18 no releases. That's what your estimate is that.
19 However, from the point of the importance 20 of the thing, even things will just to lead to core 21 damage without leading to releases should be 22 considered important.
23 MEMBER BROWN: I've got one question.
24 MEMBER BALLINGER: I'll say it again, a 25 mechanical stop cuts the head off of this snake.
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38 1 MEMBER BROWN: But let me ask a question 2 relative to that. Obviously if you've got a computer-3 controlled crane, you've got a processor of some kind 4 in it, right.
5 DR. NICHOL: Correct.
6 MEMBER BROWN: A microprocessor or 7 something. How do you handle a circumstance where the 8 processor locks up? You call it single failure-proof, 9 but what does that mean relative to the software? Do 10 you have redundant software that runs this that has to 11 be compared? Do you specify for that?
12 DR. NICHOL: So single failure-proof 13 refers to the portions of the crane responsible for 14 handling the load such that it isn't dropped.
15 MEMBER BROWN: Okay, well is that 16 software-controlled?
17 DR. NICHOL: The single failure-proof 18 portions of the crane are physical.
19 MEMBER BROWN: Are what?
20 DR. NICHOL: Physical. They're hardware, 21 they're redundant ropes and those sorts of things.
22 MEMBER BROWN: So the answer is no. Well, 23 what makes it move?
24 MEMBER MARCH-LEUBA: No, the answer is 25 digital means. Digital means, software.
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39 1 MEMBER BROWN: Software.
2 DR. NICHOL: Right.
3 MEMBER BROWN: So what, if it's moving, 4 what prevents it from failing if the software locks 5 up?
6 DR. NICHOL: So there will need to be a 7 redundant system that monitors the operation of the 8 existing software to ensure essentially that that 9 doesn't happen.
10 MEMBER BROWN: Well, I know, but you just 11 commented that the mechanical parts are designed to be 12 backed up so that you don't have it, it's a single 13 failure-proof.
14 DR. NICHOL: Right.
15 MEMBER BROWN: But you don't make any 16 comment about the software being backed up, the single 17 failure-proof.
18 DR. NICHOL: Right.
19 MEMBER BROWN: You kind of limited it 20 between the software-based systems and the mechanical-21 based systems. And if anything we've ever learned, 22 software locks up. It'll do it unexpectedly, you're 23 moving your mouse and all of a sudden it doesn't move 24 anymore, just randomly.
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40 1 are they going to use custom software, is it going to 2 be commercially dedicated software with all kinds?
3 You're going to get locked up if you don't provide 4 some safeguards, I don't, safeguard's the wrong word, 5 requirements to provide monitoring and/or backup 6 and/or prevent lockup of that software, and have that 7 lockup result in something that stops everything so 8 that you can take manual control.
9 If you don't tell them to do that, how do 10 you do it? Right now, it seems like there's a 11 division of responsibility right now. Here all these 12 things are going to be redundant and single failure-13 proof, but the software is just fine because it's not 14 a problem. It's kind of a hard, that's a hard nut to 15 follow. And I just throw that on the table and let 16 you deal with it.
17 DR. NICHOL: Okay.
18 MEMBER CORRADINI: My thought is we need 19 to move on, but I think you get the sense of it is 20 specifications that are given by NuScale, go to the 21 vendor, come back, are checked by NuScale and are 22 inspected by the staff might be what I'll call a 23 consensus opinion of the members.
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41 1 in terms of your system design.
2 MEMBER KIRCHNER: It goes back to Harold's 3 question, what are we certifying here, just a single 4 module? Or what's unique about this plant is the idea 5 of having multi modules and yet, the amount of detail 6 that's being provided for critical functions is scant 7 in this particular case. And yet, this is at the crux 8 of multi-mode operation. I just find it deficient.
9 That's one person's opinion.
10 MEMBER BROWN: You've really got to think 11 about a fail-safe circumstance for the software-based 12 systems, and that doesn't seem to be addressed based 13 on the conversations I've heard. I'll stop at this 14 point.
15 C0-CHAIR REMPE: So I think we've made our 16 point. And maybe if we haven't, we'll come back with 17 -- just let's go ahead. Unless you have some 18 clarifying thing like oh, you should look at chapter 19 such and such and you can see this, I think you need 20 to move on, but you can understand there's some 21 concern.
22 DR. NICHOL: Yes, let's move on.
23 C0-CHAIR REMPE: Okay.
24 CO-CHAIR SKILLMAN: Corrie, let me ask one 25 other question before we move on.
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42 1 DR. NICHOL: Okay.
2 CO-CHAIR SKILLMAN: I searched the word 3 buoyancy and I find that intriguing because the way 4 the documents are written, it seems to identify that 5 buoyancy might be variable from one place to another.
6 And I would just assert that if you're taking an 7 imaginary one cubic foot of something to the 8 Challenger Deep in the Marianas Trench or you put that 9 same cubic foot of imaginary in your bathtub, it's 10 still about 62.4 pounds of buoyant upward through the 11 centroid of that mass.
12 Why is buoyancy a concern here? And what 13 relationship would buoyancy have to the load-lifting 14 capability of the crane, if any? My thought is none, 15 but I'm asking the question because I searched on 16 buoyancy and it shows up a couple of places.
17 DR. NICHOL: So we credit the buoyancy of 18 the module when we assign a name plate capacity for 19 the crane.
20 CO-CHAIR SKILLMAN: And so you consider 21 the whole module voided?
22 DR. NICHOL: No. The condition that we 23 analyze is the condition that it will be in in 24 refueling which includes water in ECCS system and as 25 --
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43 1 CO-CHAIR SKILLMAN: So the containment is 2 flooded?
3 DR. NICHOL: Right.
4 CO-CHAIR SKILLMAN: So what would be 5 buoyant?
6 DR. NICHOL: We displace water with the 7 steel that makes up the reactor module. And we don't 8 flood the entire space in the containment vessel.
9 MEMBER CORRADINI: Is there a -- I figured 10 you were going to say that. Is there a reason for 11 that?
12 DR. NICHOL: To keep the -- I don't know, 13 is that proprietary?
14 MEMBER CORRADINI: We can hold off.
15 CO-CHAIR SKILLMAN: Let's talk about that.
16 I'm just curious because what is unique here is, if 17 you will, a buoyancy discussion for a module you don't 18 find that when you're refueling any other reactor.
19 DR. NICHOL: Right.
20 MEMBER CORRADINI: Let's talk about it in 21 the proprietary section, please.
22 DR. NICHOL: Right.
23 MEMBER CORRADINI: Thank you.
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44 1 in the full-level bases as an important function.
2 MEMBER CORRADINI: I know. That's why I 3 asked the question. That's where I went searching.
4 Thank you.
5 DR. NICHOL: Okay, so briefly, I want to 6 talk about the actual operating bay. In this image, 7 you can see the safe-load paths are illustrated so 8 maybe I can just talk through bringing a module in and 9 then refueling operations will, of course, be the 10 reverse of that.
11 So when a module is brought into the 12 plant, it's on its side. It's brought in where the 13 pointer is in the top left of the figure. It's 14 brought in horizontally. It's brought in and then 15 upended in the inspection rack.
16 C0-CHAIR REMPE: Is the crane used at this 17 point or is something else used?
18 DR. NICHOL: Once the module is upended, 19 the crane then picks up the upper module to move it to 20 the next station.
21 C0-CHAIR REMPE: So at this point it's 22 horizontal. It's not upended. So how are you 23 bringing it in is what I'm asking?
24 DR. NICHOL: It's brought in horizontally 25 and then upended.
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45 1 C0-CHAIR REMPE: By not the crane.
2 DR. NICHOL: Correct.
3 C0-CHAIR REMPE: By other devices.
4 DR. NICHOL: Correct.
5 C0-CHAIR REMPE: A tractor or something, 6 I don't know. And then a crane helps it upend?
7 DR. NICHOL: No, there's a piece of 8 equipment that's the module upender.
9 C0-CHAIR REMPE: Okay. It's brought in or 10 it always stays in there?
11 DR. NICHOL: It's a part of the refuel --
12 C0-CHAIR REMPE: Okay.
13 DR. NICHOL: Or the inspection rack.
14 C0-CHAIR REMPE: Okay. I didn't see this 15 in Chapter 9. Maybe I missed it. But it's there?
16 MS. FOSAAEN: The upender is described in 17 3.
18 C0-CHAIR REMPE: Three. Okay.
19 DR. NICHOL: So once the module, the top 20 portion of the module is brought in, I guess I should 21 preface this, the lower reactor pressure vessel and 22 the lower containment vessel are brought in separately 23 and upended separately and placed in the reactor 24 flange tool and the containment flange tool by means 25 of the wet hoist.
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46 1 Once the module is brought in and upended, 2 the reactor building crane picks it up and brings it 3 to the reactor flange tool where it's assembled to the 4 lower reactor pressure vessel. Then the assembly is 5 picked up and moved to the containment flange tool 6 where it's assembled to the lower containment vessel 7 and then it's moved to the appropriate operating bay.
8 The safe load paths are illustrated there.
9 They're the fairly narrow darker bands that 10 illustrates essentially where the module can be moved 11 by the crane. The other cross-hatched area is the 12 area of coverage for other operations that include wet 13 hoist operations and other auxiliary hoist operations 14 in the pool.
15 And I'll also point out there is a heavy 16 load exclusion zone where the reactor building crane 17 is prevented by interlock from the trolley traversing 18 over that space where the spent fuel racks are in the 19 spent fuel pool.
20 This is a chart that describes the design 21 codes that are applicable to different pieces of 22 equipment. You can see the reactor building crane 23 main hoist. I mentioned ASME NOG-1, Type I. That's 24 also the code that's used for the auxiliary hoist.
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47 1 module lift fixture are both designed to ANSI N14.6 2 which covers below the hook-lifting devices.
3 The wet hoist is an NOG-1 hoist and the 4 jib crane is designed to the ASME NUM-1, Type II 5 standard. NUM stands for Nuclear Underhung Monorail.
6 So it's a different configuration crane, so it's 7 covered by a different standard.
8 Also, noted on the chart are the seismic 9 category, the different pieces of equipment and the 10 maximum traverse and hoist speeds of the different 11 equipment.
12 I would mention there are a few COL items 13 on the crane. The first covers the process for 14 handling receipt of critical loads including the 15 module. The second one talks about spent fuel cask 16 handling equipment including the procedures for safe 17 handling of the spent fuel casks. The last one talks 18 about the governing procedures. We mentioned this, 19 the training and governing procedures that will be 20 required for operation of the heavy load handling 21 systems.
22 DR. SCHULTZ: It's these areas where I 23 find it -- I'll call it strange or surprising that 24 you're putting this to the COL applicant. I just have 25 to believe you'd want to provide very specific NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 1 instructions associated with these types of things.
2 Project that you've got five NuScale facilities or 3 you've got 10 or 100 NuScale facilities, to have COL 4 applicants doing different things at different places, 5 I just don't quite understand why you wouldn't want to 6 have very specific instructions, training programs, 7 procedures and so forth. These are very important 8 things.
9 You may be able to survive an accident or 10 a mishap associated with movement of a module, but if 11 that happens and it will be significant for the 12 NuScale industry, you don't want it to happen. So I'm 13 surprised that if you don't have them, you would put 14 off to a COL stage something that would be that 15 significant and be totally involved in the process 16 rather than have a COL applicant come up with this at 17 that stage of the process.
18 MS. FOSAAEN: And I would agree. We at 19 NuScale, we do intend to have a lot of services 20 available at future times. At this point, the focus 21 was on the design and we do have development in-house.
22 And so those would be services we would intend on 23 offering a future applicant.
24 And I think one thing that's important 25 here is that you mentioned five potential reactors.
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49 1 So there would be an advantage to standardization and 2 something we didn't really mention is that for us 3 refueling, I know you've said would be a frequent 4 evolution. So the advantage to that is we'll have a 5 lot of operating experience. We'll have crews that 6 are very familiar, as they'll be constantly performing 7 these evolutions, whereas in a traditional plant 8 you're only doing a refuel 18 months or 2 years. And 9 our cranes are always accessible as a result of not 10 being inside a locked containment. So there are some 11 advantages to our design. And I agree with the 12 points. And thank you for that.
13 MEMBER BALLINGER: I guess I didn't notice 14 this, but the last item on that slide, detailed 15 description of the safe load paths for movement of 16 heavy loads. I find that astounding because one of 17 your previous slides showed the safe load paths.
18 MS. FOSAAEN: Right.
19 MEMBER BALLINGER: That's got to be tied 20 to the PRA and every other thing.
21 MEMBER KIRCHNER: It has to be well 22 specified.
23 MEMBER BALLINGER: Has to be specified 24 just doesn't make any sense to me at all.
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50 1 have specified safe-load paths. That item was added 2 specifically in response to a staff question.
3 DR. SCHULTZ: Right. It was an RAI 4 response. I appreciate what you're saying, but you 5 say we can provide this as an offering to a COL. I 6 just think it's too important to NuScale to leave it 7 to the COLs.
8 I know the hubris of the industry is such 9 that oh, we can do this. We don't have to have 10 commonality, but think of the history of the industry 11 and at some point we determined that we needed a PWR 12 owners group and a PWR owners group activity in many, 13 many different areas.
14 It would be nice if we didn't have to do 15 that later, but if we did it now, and make it easier 16 for the first COL applicant and those that follow to 17 get this done.
18 MEMBER RAY: Not only that, Steve, but you 19 know, we're focused here on the certified design of 20 the scope and the content of that. The intentions 21 that exist for providing information, help assistance, 22 guidance and so on down the road, that's not part of 23 the certification. It's not something we can credit.
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51 1 the holder of that design may ultimately do years in 2 the future. And so we're just trying to get into the 3 certified design, the things that we think are 4 essential. Crediting what the present intentions are 5 in the future just can't be part of that process.
6 C0-CHAIR REMPE: I think we made our 7 point. Let's keep going. Thank you.
8 MR. HARRIS: Okay, on to 9.2 which is the 9 wire systems. This includes station service water is 10 not in NuScale design, but 9.2 includes reactor 11 component cooling water systems, demin water system, 12 potable and sanitary water, the ultimate heat sink, 13 condensate storage, site cooling water, chilled water 14 and utility water.
15 The majority of these systems are a 16 typical industry design, so in the interest of time 17 I'm going to only focus on 9.2.5, the ultimate heat 18 sink. Otherwise, the remainder of those systems are 19 non-safety related and non-risk significant.
20 C0-CHAIR REMPE: I'm going to stop you 21 here though since I think that means you don't plan to 22 talk about the reactor component cooling water system.
23 Is that a true assumption?
24 MR. HARRIS: I can talk about it if you'd 25 like.
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52 1 C0-CHAIR REMPE: I'd like it.
2 MR. HARRIS: Okay.
3 C0-CHAIR REMPE: Just briefly, at the last 4 month's meeting, I was told because of whatever, the 5 multi-module chapter that you have said clearly you're 6 going to install what's needed for when you have a 7 less than 12 modules.
8 So in this component cooling water system, 9 if you had 12 modules, you'd have 3 systems of 2 10 identical systems that each supports 6 modules and 11 then you have a third system that's a backup. So in 12 this case, even though I didn't see it clearly stated 13 in the DCA, I would assume you will do the backup and 14 the one system for whenever you even have one module 15 installed. Is that true?
16 MR. HARRIS: I can't speak to particular 17 construction activities about what, if you'd install 18 one or both reactor component cooling water systems.
19 I believe there's only -- should be only two reactor 20 component cooling water subsystems, one for each side, 21 north and south of the reactor cooling.
22 CO-CHAIR REMPE: Right, to the six, but 23 there's a backup. When I have module number 1 that 24 means I should have one of these two systems installed 25 and the backup, right?
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53 1 MR. HARRIS: I'm not sure what you're 2 referring to as a backup. It's a non-safety system.
3 There's no -- we have redundant pumps.
4 C0-CHAIR REMPE: The power train is not in 5 operation, but it's kept in stand by and started, if 6 needed. I'm quoting from the open DCA, okay? And so 7 to me, that means clearly that I've got to have the 8 backup as well as one of these systems installed from 9 day one if I have one module. And that's all I'm 10 asking because I mean that's what I think we read 11 according to this multi-module chapter that you do, 12 and I just am asking for clarification. Because 13 sometimes there's a system that supports like three or 14 four modules. Sometimes there's one that supports 15 six. But then there's like sometimes a backup system 16 and I'm just asking can we assume that, because we are 17 certifying a very -- sometimes fuzzy -- DCA.
18 So I'm getting into another question of 19 what are we certifying here, but just an example. And 20 I would have expected you have said immediately, oh, 21 of course, we'll have that back-up system when we --
22 before we start one of these modules up.
23 MS. FOSAAEN: I think it was just the 24 terminology that was confusing. The system would need 25 to have its basic functions. So if it was credited to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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54 1 have all three pumps installed, even for one module, 2 that would be required.
3 C0-CHAIR REMPE: I can assume, and this is 4 on the record, that again, before you start one 5 module, we'll have one RCCWS and its backup.
6 MS. FOSAAEN: And by backup, we mean the 7 full suite of pumps that would be in there.
8 C0-CHAIR REMPE: Okay. Okay. Go ahead, 9 now. Sorry.
10 MR. HARRIS: And so I think like Carrie 11 said, it's a difference in terminology, I think.
12 C0-CHAIR REMPE: Okay.
13 MR. HARRIS: So other than the ultimate 14 heat sink, none of these systems are an essential 15 source of water, nor are they required during or after 16 a natural phenomenon event. And anything in proximity 17 to a Seismic Category 1 component is designated 18 Seismic Category 2.
19 On to the ultimate heat sink. Like I 20 mentioned previously, the ultimate heat sink consists 21 of the reactor pool -- refueling pool and spent fuel 22 pool. It is a stainless steel lined reinforced 23 concrete pool filled with borated water.
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55 1 under what will be the controlled load path for the 2 module? Is it a one inch, two inch, a --
3 MR. HARRIS: It's less than one inch. I 4 can't speak specifically --
5 CO-CHAIR SKILLMAN: It's less than an 6 inch?
7 MR. HARRIS: Not off the top of my head.
8 CO-CHAIR SKILLMAN: Is it puncture proof, 9 if that module gets away from the -- from your main 10 hook?
11 MR. HARRIS: I wouldn't expect -- so -- I 12 just want to clarify, the pool liner is non-safety 13 related. So the safety-related structure is the 14 reactor building.
15 CO-CHAIR SKILLMAN: So it's the concrete 16 underneath.
17 MR. HARRIS: Right.
18 CO-CHAIR SKILLMAN: Got that. So now I've 19 got a 740 ton and I drop it, do I puncture that less 20 than one inch pool liner?
21 MR. HARRIS: I would expect if you dropped 22 a 900-ton load, it would puncture steel less than an 23 inch.
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56 1 under that safe-load path or something to make sure 2 that no matter what, that pool liner cannot be 3 punctured? Just think about it. You don't have to 4 answer.
5 MR. HARRIS: Okay. Thank you.
6 CO-CHAIR SKILLMAN: Thank you.
7 MR. HARRIS: The ultimate heat sink 8 consists of approximately seven million gallons of 9 water, so it's sized such that active cooling systems 10 are not required for accident conditions. And that 11 combined volume of water provides sufficient coolant 12 for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without additional make up.
13 The ultimate heat sink is also provided 14 with a seismically qualified make up line that can be 15 connected external to the reactor building if 16 additional water were to be added.
17 And during normal operations, spent fuel 18 pool cooling and reactor cooling system services this 19 body of water and in a design basis event involving 20 loss of AC power decay heat removed from the modules 21 and into the ultimate heat sink.
22 MEMBER CORRADINI: So let me ask just a 23 quick question. So is there a procedure as to the 24 replenishment after -- if there's change in level?
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57 1 but so many days before replenishment.
2 Is there a procedure for replenishment in 3 terms of at some point at some level?
4 MR. HARRIS: So the level is covered by 5 tech specs. So I believe the tech spec level is 6 greater than 68 feet, normal level, so anything below 7 that would trigger actions per the tech specs.
8 MEMBER CORRADINI: Thank you.
9 MR. HARRIS: And a number of COL items for 10 chemicals used, sanitary waste disposal, corrosion, et 11 cetera.
12 On to process auxiliaries. This consists 13 of the compressed-air system, process sampling, 14 equipment and floor drain system, chemical and volume 15 control system, stand by liquid control is not 16 applicable to our design, and also the containment 17 evacuation system and containment drain system for 18 9.3.6.
19 Similar to the other water systems, I'm 20 only going to focus on the items unique to our design 21 so with the exception of chemical and volume control 22 system. These are non-safety related, non-risk 23 significant systems.
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58 1 Category 2, and we have two COL items, leakage control 2 program and post-accident sampling.
3 With the chemical and volume control 4 system --
5 C0-CHAIR REMPE: I guess before you go 6 there, just for the record, apparently, the sampling 7 system, there's an issue about -- it's on hold because 8 of the source term issue, right?
9 MS. FOSAAEN: That's correct. There is an 10 open item from the staff in the review relative to 11 that. And NuScale recently did submit an exemption in 12 relation to that that is affecting that review.
13 C0-CHAIR REMPE: So you've requested an 14 exemption, submitted all the documentation and the 15 staff is reviewing it at this time.
16 MS. FOSAAEN: That's correct.
17 C0-CHAIR REMPE: And that's where the hold 18 is right now. Thank you.
19 MR. HARRIS: All right, on to the chemical 20 and volume control system. It's used to purify 21 reactor coolant, maintain chemistry including boron 22 concentration. Also provides for make up and let down 23 and supplies pressurizer spray flow and also provides 24 a de-gas operation.
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59 1 system per module and during start up is used in 2 conjunction with the module heat system to raise the 3 reactor coolant temperature to generate natural 4 circulation. And also, it's used in conjunction with 5 the boron addition system which is a shared system for 6 all 12 modules.
7 MEMBER MARCH-LEUBA: Don't move. This 8 system, the CVCS is dear to all of us, we tend to 9 think about. Let's make sure I understood because 10 typically I am the one that nobody understands when he 11 speaks, but the acoustics in this room is so terrible 12 that I don't understand you.
13 MR. HARRIS: Okay.
14 MEMBER MARCH-LEUBA: You have to speak 15 more clearly. There is one CVCS system for the whole 16 plant. Is that correct?
17 MR. HARRIS: No. There's one per 18 operating module.
19 MEMBER MARCH-LEUBA: There's one CVCS 20 system --
21 MR. HARRIS: So there's 12 total.
22 MEMBER MARCH-LEUBA: So there is one BAS 23 system per module?
24 MR. HARRIS: No. One BAS per all 12 25 modules and --
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60 1 MEMBER MARCH-LEUBA: It doesn't appear on 2 the drawings. So each module has its own CVCS?
3 MR. HARRIS: Correct.
4 MEMBER MARCH-LEUBA: And they are not ---
5 other than to the BAS system, they're not cross tied?
6 MR. HARRIS: BAS is just -- so correct, 7 within the CVCS, they're not cross tied.
8 MEMBER MARCH-LEUBA: Completely 9 independent. So if one fails, you cannot bring --
10 MR. HARRIS: It will not impact.
11 C0-CHAIR REMPE: But there's a backup 12 power supply system, right, that includes -- there's 13 two -- the backup power supply system, I guess, 14 includes two redundant diesel generators and one 15 auxiliary alternating current power source.
16 How much of that is installed when you 17 have the first unit and six units and seven units, et 18 cetera? Is all of that installed from day one?
19 Because that is needed for the CVCS and I assume it's 20 given credit for it.
21 MR. HARRIS: So any system, whether it's 22 -- if it's shared or not, required for a module 23 operation will be installed and as you bring other 24 modules on line, module-specific systems would be 25 installed for those modules. So for your example, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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61 1 backup diesel generator, that would be installed to 2 support that initial first module coming on line.
3 C0-CHAIR REMPE: So the two -- for modules 4 less than six, would you have one diesel generator, 5 but you'd have the alternating current power source 6 installed? Or would you put both diesel generators in 7 from day one?
8 MR. HARRIS: I'm not sure of the specifics 9 of the backup diesel generator, if there's one or two.
10 C0-CHAIR REMPE: Maybe it's an unfair 11 question to ask you, but do you think it's documented 12 in the DCA somewhere other than a vague statement?
13 MEMBER CORRADINI: Your question is where 14 is it documented?
15 MS. FOSAAEN: I know that 20.1 discusses 16 at a high level the multi-module in bringing 17 additional modules on line, but I don't believe the 18 level of detail you're requesting is specified in the 19 DCA.
20 C0-CHAIR REMPE: So I'm just kind of 21 wondering when they start building this, and I know, 22 the economic case depends on being able to produce 23 power with a few modules while you're still bringing 24 in the other modules. But to make sure that they've 25 appropriately done DCA, it seems to me some sort -- I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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62 1 think we've looked at that test in chapter -- was it 2 20, 21, whatever it was you said. And it's very high 3 level and it kind of commits, yes, we're going to 4 bring in everything that's needed for the modules.
5 Okay, I might have accepted when you said well, we'll 6 put one of the diesel generators, but that back up 7 system, it seems like it's got to be there.
8 Somehow or other, I'd like to see a plan 9 that provides some guidance on how it's going to be 10 added on because we get a lot of single units sites in 11 the U.S. where they've got a concrete pad and it's 12 like the second unit never got installed.
13 CO-CHAIR SKILLMAN: Let me raise an issue 14 here on CVCS. When I review, let's back up. CVCS 15 provides the ability to maintain the chemistry of the 16 reactor coolant system. It's a back up for adding 17 boric acid to your seven million gallon pool. It 18 serves a host of key functions.
19 If I'm an operator and I wish to add boric 20 acid to the reactor coolant system, I would have to 21 light off my boric acid system. I've got to work my 22 way through a bunch of valves, find my way to the 23 make-up pump, and at the make-up pump increase the 24 pressure from 152 psi up to 2200. And only then do I 25 get that boron, boric acid, into the core.
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63 1 From Table 3.2-1, everything in CVCS is 2 quality group D. It's 2 Bravo, excuse me, Bravo 2 out 3 of .69 and it's all Seismic Category 3 for both CVCS 4 and boric acid.
5 My point is a system upon which I very 6 significantly depend is basically commercial grade, no 7 augmented QA equipment. Why should I be comfortable 8 with that? It's a key system.
9 MR. HARRIS: I agree. It's a key system.
10 I do want to clarify a point. There are two safety 11 related demineralized water isolation valves which are 12 safety-related.
13 CO-CHAIR SKILLMAN: That's great. I'm not 14 for putting in un-borated water. I agree with that.
15 MR. HARRIS: I agree.
16 MS. FOSAAEN: I think the key point --
17 CO-CHAIR SKILLMAN: I want a lot of boron 18 and I want it now.
19 MS. FOSAAEN: Right. And our Chapter 15 20 analysis does not credit CVCS for any boron addition.
21 CO-CHAIR SKILLMAN: I know. Isn't that 22 great. I want boron. I'm an operator. I want shut 23 down. I want as much as negative reactivity as I can 24 throw at that core for whatever reason. And I have no 25 way to do it unless I go through a daisy chain.
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64 1 Doesn't that spark a little bit of uncomfort? I would 2 think it would to those who are people who have held 3 reactor operator licenses. And every other 4 application, you push a button, you get copious boric 5 acid. And on the Bs (phonetic) you've got slick. And 6 on this plant, I don't have that capacity. Is that 7 something that ought to be looked at more thoroughly?
8 MR. HARRIS: I'd also like to add that 9 there are redundant make-up pumps, redundant 10 recirculation pumps, so there's redundancy within the 11 system if an event a pump failed.
12 CO-CHAIR SKILLMAN: I still have a daisy 13 chain to get boric acid to my core. I'll just leave 14 it at that, okay?
15 MR. HARRIS: Okay.
16 MEMBER MARCH-LEUBA: And I want to second 17 that opinion.
18 MR. HARRIS: Thank you.
19 MEMBER MARCH-LEUBA: We will see in 20 Chapter 15. Thank you.
21 MR. HARRIS: So the chemical and volume 22 control system is also the only system with 23 connections to the reactor coolant system. The piping 24 runs outside containment, so it is the total scope of 25 the inner system LOCAs for consideration in NuScale NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 1 design.
2 The containment isolation function for 3 CVCS is also part of the containment system which is 4 covered in Section 6-2.
5 As I mentioned, CVC is equipped with two 6 automatic safety related fail closed demineralized 7 isolation valves to ensure no inadvertent boron 8 dilution concentration. Those are covered by Tech 9 Spec 346 for containment isolation valves.
10 Section 936 is the containment evacuation 11 system and containment flooding and drain system.
12 These systems are used to transfer liquids and gases 13 between the containment vessel free volume and other 14 plant systems.
15 The containment evacuation system 16 establishes and maintains a vacuum in the containment 17 vessel by a moving water vapor and non-condensable 18 gases in the CNV. It also has a function to perform 19 leakage detection for a reactor coolant system.
20 The containment flooding and drain system 21 is used to flood the containment vessel with borated 22 water after shutdown in preparation to removing -- to 23 moving the module and it's also used to drain the 24 module prior to start up.
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66 1 system also has a function to provide borated cooling 2 inventory to the containment vessel during a beyond 3 design basis event.
4 The isolation function, containment 5 isolation function is covered in Section 6-2 and in 6 tech specs for these systems are 347, 349, and 373.
7 And here's just diagrams of the 8 containment evacuation system. It consists of two 9 redundant vacuum pumps. These pumps are operating 10 continuously. Any liquid that's removed from the 11 vessel is condensed and collected and quantified in 12 the sample vessel and the non-condensable gases are 13 sent to either the reactor ventilation system or 14 gaseous rad waste system.
15 And the containment flood and drain system 16 consists of two redundant pumps. There's one 17 containment flood and drain system per six modules, 18 north and south of the reactor building, so they each 19 serve six modules apiece.
20 On to 9.4 which is air conditioning, 21 heating and cooling, and ventilation system. This 22 covers control room area ventilation, reactor venting 23 and ventilation in the spent fuel pool, ventilation 24 radiative waste building ventilation, turbine building 25 ventilation. Section 945 engineering, continuous NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 1 safety-feature ventilation system is not applicable to 2 the NuScale design.
3 A common element to these systems are they 4 are non-safety related or risk significant, not 5 credited for public dose mitigation. Any components 6 which could adversely affect Seismic Category 1 are 7 designated Seismic Category 2 and we had a COL item to 8 specify periodic testing inspection program 9 requirements.
10 CO-CHAIR SKILLMAN: Let me ask this on the 11 ventilation systems, one characteristic that people in 12 the plant and the plants know real well is that 13 ventilation systems are systems where you don't get 14 wet and you don't get irradiated. And so most of all, 15 you ignore them and they sit there and they turn away 16 until the force draft fan of the blower breaks an axle 17 and starts making noise, if the maintenance doesn't 18 put some grease like they should. So it's squeaking 19 and you say something needs to be fixed.
20 But the ventilation systems provide the 21 negative pressure barriers and the contours for 22 ventilation to make sure that the operators are safe.
23 So here are these ventilation systems that are almost 24 not too different than a ventilation system in an 25 airplane hangar in this application.
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68 1 What makes sure that these ventilation 2 systems are performing the way they're supposed to?
3 Because they actually do perform -- you might not call 4 it a safety function, but it's certainly an important 5 function for the operators.
6 MR. HARRIS: There's a number of assurance 7 that you would use to ensure system performance flow, 8 pressure, pressure across filter banks. And you're 9 going to be continuously monitoring those systems and 10 even non-safety alarms tied to anything that would 11 identify improper performance.
12 CO-CHAIR SKILLMAN: Thank you. Right?
13 MR. HARRIS: On to 9.5. It includes 14 communication system, lighting system, fire protection 15 program, and fire hazards analysis. So the 16 communication system includes components for inter-17 plant and plant off-site communication. This includes 18 private branch exchange, public address, general alarm 19 system, sound power telephones, distributed antenna, 20 and radios for point-to-point communication. Serves 21 no safety-related risk significant functions. It's 22 not credited for design basis accidents or safe 23 shutdown. And there are two COL items attached to the 24 system.
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69 1 is your fire protection program. Chapter 9 is void of 2 any description of your firefighting system. No 3 identification of plumbing. I don't know how many 4 diesel driven fire pumps or electrically driven fire 5 pumps you have. I don't know where your fire mains 6 are. You do communicate in Chapter 3.2 or Table 3.2 7 item 1. It's all non-safety and it's not seismic.
8 Fire protection is probably one of the 9 most important systems you have in the plant. But 10 there isn't anything in the design cert. application 11 that points to this. What it does communicate is that 12 some fire piping is likely Seismic II over I. And in 13 that case, it won't be Seismic III. It will be 14 something else. But that's a real loose requirement 15 for what is at least in my judgment a critical system 16 in this design.
17 Why isn't there any treatment of the fire 18 system in Chapter 9?
19 MR. HARRIS: So the fire protection system 20 is described in 9.5.1. We'll be getting to that. We 21 kind of just reversed order so we could do the fire 22 hazards analysis together in this presentation.
23 CO-CHAIR SKILLMAN: I would suggest that 24 it's mighty sparse.
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70 1 of compliance with Reg. Guide 1.189. And there's 2 where you get all of the specifications for how much 3 water to be delivered to given areas. How many stand 4 pipes, what the yard configuration is, the separation 5 of the yard piping, and all that kind of stuff. All 6 that is included in the compliance to Reg. Guide 1.189 7 table in Section 9.5.1.
8 MEMBER BLEY: It sounds as if the system 9 is not designed yet. Is that right?
10 MR. FIELDS: I don't know how -- what full 11 extent it's designed.
12 MR. HARRIS: It's designed to the point 13 necessary for DCA.
14 MEMBER CORRADINI: Say that again?
15 MEMBER BLEY: There's no schematic.
16 There's no --
17 MS. FOSAAEN: There is, in 9.5.1, there's 18 Figure 9.5.1-1 and it shows the pumps, the tanks. So 19 there are high-level schematics in 9.5.1 at the tail 20 end.
21 MEMBER MARCH-LEUBA: Can you say again 22 what table?
23 MS. FOSAAEN: 9.5.1-1 and 9.5.1-2.
24 MEMBER CORRADINI: But you made a point 25 and maybe I didn't understand it. You said it's not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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71 1 required for DCA?
2 MR. HARRIS: I'm not saying that. I'm 3 just saying that it's designed up to the point 4 required for design specification application. It's 5 not in detailed design.
6 MEMBER CORRADINI: Thank you.
7 C0-CHAIR REMPE: So I didn't come back and 8 give you examples where I found the QA program might 9 have a problem. I think that they throughout the 10 staff's review, they noticed the seismic categories 11 were incorrectly identified. There were plant layout 12 drawings that had been updated and not reflected in 13 the DCA.
14 I'm just kind of wondering again when you 15 have something that's only designed up to a certain 16 extent, is there going to be a good process in place 17 that will say oh, I need to change something because 18 of the fire protection system design, a wall needs to 19 be thicker and you'll have noted it there, but it 20 won't get updated other places in the DCA. And I'm 21 bringing this up not to pick on you, but we've 22 actually had examples for a certified design had not 23 been updated and it's a real costly endeavor to change 24 a certified design. And so that's why I'm wondering, 25 do you have confidence you've improved your QA system NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 1 that you'll be able to track requirements from 2 different systems that if that's the actual plant 3 design and say oh, I need to change it here, too, 4 because whatever system?
5 I mean how -- we noticed some problems and 6 say oh, yeah, we need to fix it and we have done that 7 or you just don't think it's a problem yet?
8 MR. FIELDS: Yes, we've gone through a 9 very painstaking process over the last year to ensure 10 that the DCA is under configuration control and we've 11 captured changes that we had missed in previous 12 changes --
13 C0-CHAIR REMPE: That's the answer I was 14 looking for.
15 MR. FIELDS: Yes, we've been going through 16 that process to go ensure that and as we grow and as 17 we change into customers and other applications, we're 18 setting up systems to capture that so that it doesn't 19 cross pollinate or doesn't cross boundary lines for 20 one design versus another.
21 C0-CHAIR REMPE: Thank you.
22 MEMBER BROWN: I would presume the COMS 23 systems that you talk about in your first paragraph 24 there include general alarm systems, branch, public 25 address, and it's intra-plant as well as plant to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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73 1 offsite.
2 Is there any definition of how those 3 communication systems are ensured that they don't 4 provide false information to plant people since your 5 external to internal plant information? For instance, 6 a general alarm could be sounded based on a cyber-7 attack that comes in and turns on the alarm and 8 everybody puts their hair on fire, or is that say, you 9 don't specify anything relative to that through the 10 COL anywhere?
11 Are you putting it off or are you trying 12 to isolate critical functions that could provide 13 general alarms throughout the plant, so as to not 14 inadvertently done by hacking. I'm sure your system 15 is computer controlled. They're all done that way.
16 MR. HARRIS: Yes, if I may, I'd like to 17 Jeff Ehlers, our electrical supervisor, who is on the 18 phone?
19 MEMBER BROWN: Your who?
20 MR. HARRIS: Jeff Ehlers. He is our 21 electrical supervisor on the phone address that 22 question.
23 MEMBER CORRADINI: Jeff, are you out 24 there?
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74 1 off a minute. So yes, I would figure any software, 2 any software that we use for our communication system 3 will have to be under the software QA program and meet 4 the same requirements as all the other software in the 5 plant for cyber security.
6 And other than that, communication systems 7 are all specified to be encrypted to add another layer 8 of protection there.
9 MEMBER BROWN: By encrypted, what do you 10 mean? It sounds good, but I'm -- encryption means if 11 I send something somewhere, you have to have a fancy 12 algorithm on the other end to decrypt it. And you 13 have to encrypt it to start with. I presume -- I'm 14 not a designer, but I assume that's the right thought 15 process.
16 MR. EHLERS: Yes. I believe you're 17 correct. We have not specifically made -- stated any 18 specifics about that encryption, but there is a 19 requirement in the DCA for all communications to be 20 encrypted.
21 MEMBER BLEY: I've got a question. I 22 spent time in the Navy and we used sound powered 23 phones all the time. I don't usually see them in a 24 commercial plant. Will they be throughout the plant 25 to be used when the other com systems don't work well?
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75 1 MR. HARRIS: Jeff, can you answer that 2 one?
3 MR. EHLERS: Sorry, the beginning of the 4 question --
5 MEMBER BLEY: Yes. I haven't seen sound 6 powered phones in commercial plants. Of course, the 7 Navy has had them forever.
8 Two questions about it. One, are they 9 rigged throughout the plant so operators can use them 10 when other communication systems are down? And two, 11 all the ones I'm familiar with have really hideous 12 frequency response and the first time you use them, it 13 just sounds like gobbledy-gook and you have to 14 practice for a week or more and pretty much routinely 15 to be able to understand people. You can understand 16 perfectly well when you're used to them.
17 And is there anything about training 18 people on using these sound powered phones or maybe 19 they're better than they used to be.
20 MR. EHLERS: Well, I think your first 21 question is about the industry. They're already used 22 in the industry. They were used at my Houston plant.
23 MEMBER BLEY: Oh, okay. I haven't seen 24 them, but go ahead.
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76 1 nuclear industry. They do use them. There are 2 designed -- they are jacks installed in various 3 locations to be able to plug one in and use -- as long 4 as you have a proper path that you can be able to use 5 it.
6 MEMBER BLEY: Okay, so the jacks are 7 around the plants such that you could use them for any 8 operations you need to do?
9 MR. EHLERS: Correct. If you lose all the 10 other battery pack communication systems, either 11 wireless, fiber optic, or whatever, then yes, you have 12 the sound powered to use.
13 MEMBER BLEY: Okay. What about legibility 14 and practice on them?
15 MR. EHLERS: Legibility and practice?
16 MEMBER BLEY: Yes, the ones I'm used to, 17 if you don't use them regularly, you can't understand 18 what anybody is saying. You have to practice kind of 19 routinely to be able to understand. Frequency 20 response, or at least used to be horrible on them.
21 MR. EHLERS: We don't have any 22 specification for, you know, how they would practice, 23 but I'm assuming would be in their ERO drills and 24 stuff like that where they would practice that kind of 25 drill where they have to go to their sound powered NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 phones. I know we have.
2 MEMBER BLEY: Okay, so that's left up to 3 whoever is going to operate the plant?
4 MR. EHLERS: Yes, it's part of the 5 emergency preparations they do.
6 MEMBER BLEY: Okay.
7 MR. HARRIS: On to 9.5.3, our lighting 8 system, plant lighting system provides artificial 9 illumination for buildings, room spaces, and outdoor 10 areas of the plant. The functions include normal 11 plant lighting, emergency plant lighting, normal and 12 emergency main control room lighting. It's in 13 accordance with NUREG post-700. We have emergency 14 lighting fed from our DC power system for a minimum of 15 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and a battery pack for one and a half hours 16 for egress or exiting in accordance with NFPA 804.
17 And the plant lighting system also 18 provides emergency lighting with battery pack for 19 eight hours for post-fire, safe shutdown activities 20 outside of the main control room.
21 Line 5.1 is our fire protection program.
22 The objective of our fire protection program is to use 23 defense in depth to achieve the required degree of 24 reactor safety. Our safe shutdown relies on passive 25 fire protection and redundant safe shutdown of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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78 1 equipment is separated by three-hour fire barriers.
2 So part of that defense in depth is prevent fires from 3 starting, to rapidly detect, control, and extinguish 4 fires that do occur, and provide protection to 5 components important to safety.
6 9.a, fire hazards analysis. Our fire 7 hazard analysis demonstrates that the plant maintains 8 the ability to perform its safe shutdown functions and 9 minimize radioactive material release in the event of 10 a fire. So our hazard analysis considers in-situ and 11 transient fire hazards, determines the effect of any 12 fire -- of a fire in any location in the plant and its 13 ability to safely shut down the reactor. Also 14 specifies measures for fire protection prevention, 15 detection, and suppression.
16 Within 9.a, our special cases are a main 17 control room fire area. We have switches that provide 18 back-up control of systems, automatically controlled 19 by module protection system.
20 In the event of a main control room fire, 21 we have manual switches associated with module 22 protection functions to isolate those actuations.
23 Those switches are located outside the main control 24 room.
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79 1 unlikely due to our containment is held at a vacuum.
2 Our electrical conductors are made of noncombustible 3 material and a lot of the conduit is inaccessible 4 during normal operation. And also during shutdown, we 5 flood our containment which also helps prevent 6 ignition or the spread of fire.
7 For the top of module fire area, also 8 considered highly unlikely. It's enclosed in our bio 9 shield and cabling under the bio shield is in a 10 conduit or three hour rated cable. This top of module 11 area is inaccessible during normal power operations.
12 The hydraulic hood used in our containment isolation 13 valves is specified to be noncombustible.
14 For our containment isolation valves and 15 decay heat removal system actuation valves, these 16 valves move to their safe position using stored 17 pressurized nitrogen accumulators.
18 And between modules, there is a three hour 19 fire barrier with the exception of the bio shield 20 while facing the reactor pool.
21 And during shutdown, the bio shields are 22 moved and administrative controls placed on that area 23 for combustible -- for transient combustibles.
24 CO-CHAIR SKILLMAN: Where are the valves 25 that are hydraulically actuated located?
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80 1 MR. HARRIS: All of our containment 2 isolation valves are hydraulically actuated and heat 3 removal actuation valves are also hydraulically 4 actuated. They're all located on top of the actual 5 module.
6 CO-CHAIR SKILLMAN: So if you had a 7 hydraulic leak, where would the hydraulic fluid leak 8 to?
9 MR. HARRIS: Assuming it's gravity 10 draining, just leak on top of the module and into the 11 pool.
12 CO-CHAIR SKILLMAN: Into the pool. And 13 what is the fluid? What is the hydraulic fluid? Is 14 that like automatic transmission fluid?
15 MR. HARRIS: I can't speak to the 16 specifics of the hydraulic fluid. It is 17 noncombustible.
18 CO-CHAIR SKILLMAN: But it is a 19 hydrocarbon.
20 MR. HARRIS: Again, I can't speak to the 21 specifics of it.
22 MR. FIELDS: I don't know if it's been 23 specified fully at this point. It's just that it's 24 noncombustible 25 MEMBER CORRADINI: Do you specify it this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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81 1 way?
2 MR. FIELDS: I don't believe it's fully 3 specified what we're going to use, but it's just a 4 noncombustible hydraulic fluid.
5 MR. HARRIS: And just to reiterate, to go 6 to the safe position for all these valves I mentioned 7 does not require any hydraulic fluid. It's all done 8 by nitrogen spring.
9 So there's a vessel on top or connected to 10 each containment isolation valve and to actuate the 11 valve you actually remove, you vent that hydraulic 12 path and it has that hydraulic path vents during 13 nitrogen spring -- moves the valve to the safe 14 position.
15 CO-CHAIR SKILLMAN: At what pressure is 16 the hydraulic fluid?
17 MR. HARRIS: I'm not prepared to answer 18 that question.
19 CO-CHAIR SKILLMAN: I suggest you take a 20 look at the operating experience from TMI II. We 21 leaked hydraulic fluid and we ended up with leafy 22 green vegetables because whatever is in that water 23 that may be organic will go after that hydraulic fluid 24 and it will grow.
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82 1 place to identify if hydraulic pressure is reducing 2 and thus identifying a leak.
3 CO-CHAIR SKILLMAN: I didn't realize there 4 was hydraulic fluid over the pool water. So I'll 5 leave that with you. Thank you.
6 MEMBER MARCH-LEUBA: Let's continue on 7 that. I'm looking at this figure 9.5.1-1 and what I 8 see here about it, fire protection system is two big 9 water tanks and three pools that spray water into 10 whatever is inside the plant, right? So am I to 11 assume that you are spraying water on top of all of 12 the control room computers and control system and the 13 operators? Or is there a second fire protection 14 system for the control room and the controllers?
15 What it specifies here is that you're 16 going to be spraying water inside the control room and 17 obviously, make it completely inhabitable and 18 nonworking.
19 MR. FIELDS: If Ed Siener's on the line we 20 could have him answer that question for you.
21 MR. SIENER: Could you repeat the 22 question, please?
23 MEMBER MARCH-LEUBA: What type of fire 24 suppression system do you have in the control room and 25 all the controllers, computers, electrical system?
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83 1 MR. SIENER: In the control room 2 specifically, there's no automatic water suppression.
3 It's a continuously operated control room so we rely 4 on detection and manual suppression in the control 5 room.
6 In electric rooms, it's somewhat similar 7 depending on the fire loading. I would say generally 8 in an electrical room, there is not automatic 9 suppression. There's detection with manual 10 suppression located per code.
11 MEMBER MARCH-LEUBA: Is that specified in 12 the CVA somewhere or it's a COL item?
13 MR. SIENER: Well, one place it is located 14 in is the 9A where we have every fire area in the 15 plant and it says specifically whether it does or does 16 not have automatic suppression.
17 MEMBER MARCH-LEUBA: Okay, I'll look it up 18 now that I know where to look.
19 Second related question, when you look at 20 the drawing of the plant, most of it is underground 21 and most of it -- well, one would consider it water-22 tight rooms, for example, the battery room. It's 23 underground and fairly water tight. When you turn on 24 all those sprinklers in the event of a fire, what 25 happens to the water that accumulates on the floor?
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84 1 Because it's underground, so the only way you can 2 drain that water is with an active pump system. Does 3 nothing exist or do you just want to make a swimming 4 pool of the battery room?
5 MR. SIENER: No, the NFPA code 6 specifically address drainage and any of those rooms 7 with automatic water suppression would have drainage 8 that would not allow de-pooling of water. Most 9 electrical equipment, other important equipment would 10 probably be elevated so that we could deal with small 11 puddling on the floor. But the drainage would take 12 care of any buildup of water.
13 MEMBER MARCH-LEUBA: These are underground 14 rooms, so it would have to be active draining. You 15 have to have power and you have to have a pump 16 working. But the point that mostly Dick is making 17 which he has a lot of experience on this is that fire 18 protection is probably the most important thing you 19 can do in this plant for safety and it looks a little 20 light on the CVA.
21 MR. SIENER: A huge advantage NuScale has 22 is compartmentalization. There's a lot of rooms and 23 we're able to separate all the safe shutdown 24 equipment. We're using three hour fire barriers 25 except where notified in those special cases. So our NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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85 1 suppression -- we assure safe shutdown by separation 2 of the equipment, so the suppression is really more 3 for defense in depth and human safety.
4 CO-CHAIR SKILLMAN: But just remember, 5 Browns Ferry was designed with that same, if you will, 6 mental framework. TVA intended for those plants to be 7 extremely safe and they underwent a terrible fire.
8 And my point would be that sometimes fires 9 happen and they get away from you and they're 10 insidious because they actually have a mind of their 11 own. You can think you're going to drive that fire 12 and the fire will drive you and you've got a couple of 13 choices. You can prevent it from happening by having 14 no combustibles or you can make the compartments so 15 small that there is no real threat to the rest of 16 plant, but now you are confined basically like a rat 17 in a cage. You can't get to it.
18 And so the whole issue of firefighting is 19 one that warrants a kind of a front lobe attack at the 20 design stage to make certain that the fire cannot get 21 to where it shouldn't get to. And where it's really 22 problematical is in the control room and your switch 23 gear and particularly where your cable spreading 24 occurs. You don't need the sermon, but just recognize 25 again it happened at Browns Ferry and those of us who NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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86 1 have had maritime experience, the worst thing that can 2 happen is that you're so compartmentalized you can't 3 get to it which means you have to have ample either 4 carbon dioxide or Cardox or wintergreen or whatever 5 you're going to have or at Browns Ferry copious 6 amounts of water readily available which means your 7 headers and your hoses have to be ready to go.
8 C0-CHAIR REMPE: Colleagues, we're already 9 like 30 minutes behind, so let's let them go through 10 and then we need to decide whether we're going to 11 bring the staff up or have a break before. But we are 12 way behind, okay?
13 Go ahead.
14 MR. HARRIS: This is our last slide, just 15 listing out our COL items, or no COL items for fire 16 protection, no unique COL items for fire protection.
17 We have generic items for fire protection program 18 development, procedure development, implementation, QA 19 program, training. And no unique COL items for the 20 fire hazards analysis, just COL items to provide site 21 specific fire hazards analysis and that's it.
22 C0-CHAIR REMPE: Do we want to have a 23 quick break and come back? So let's do that and it's 24 10:16 by my clock, so let's get back here by 10:25, 25 okay? And that will give us staff time, but I didn't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 want to cut into the chapter 16.
2 (Whereupon, the above-entitled matter went 3 off the record at 10:16 a.m. and resumed at 10:25 4 a.m.)
5 CO-CHAIR REMPE: Okay. We're back. It's 6 10:25 and I'm going to bang the gavel. So, please sit 7 down if you're in the near area. And the rest of the 8 members will come as they can.
9 So let's restart. And I'm not sure which 10 of the staff is beginning, but okay, please start.
11 MR. TESFAYE: Good morning everyone. My 12 name is Getachew Tesfaye. I am the NRC Project 13 Manager. Can you all hear me?
14 I'm the Project Manager for Chapter 9 of 15 Auxiliary System. The staff has completed the Phase 16 Two safety evaluation for the open items and submitted 17 it to ACRS.
18 This report contains four one items. And 19 one of them, one of the open items has since been 20 closed. And the remaining open items will be 21 addressed in Phase Four of this review.
22 Chapter 9 comprised of five major areas, 23 review areas and 26 sections. And our presentation 24 slides have 41 slides, but due to time limitation, 25 we'll focus our presentation on Section that have open NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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88 1 items and areas where we believe there is ACRS 2 interest.
3 Those who are presented are -- who are 4 presenting are sitting here with me. And the rest of 5 the technical staff is up in the audience.
6 With that, I'll ask the presenters to 7 first introduce themselves.
8 MS. SIWY: Alex Siwy, NRO Reactor Systems.
9 MR. STUTZCAGE: Ed Stutzcage, Radiation 10 Protection Reviewer.
11 MR. NOLAN: Ryan Nolan, Reactor Systems.
12 MS. SACKO: Fanta Sacko, Electrical 13 Engineering Branch in NRR.
14 MR. TESFAYE: And the rest of the 15 contributors to this 41 slide package are Alissa 16 Neuhausen, Raul Hernandez, Hanry Wagage, Chang Li, 17 Angelo Stubbs, Bob Vettori, Tony Gardner, Alexander 18 Chereskin, Andrew Yesnick, Nan Chien, and Dawnmathews 19 Kalathivettil.
20 So, with that, if it's okay, we'll do the staff 21 observation.
22 MS. SIWY: Thanks Getachew. Again, my 23 name is Alex Siwy. And I'll walk you through the 24 staff's review of Section 911.
25 The main objective of the review of this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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89 1 Section is to verify that fuel will remain sub-2 critical during all credible storage and handling 3 conditions in accordance with 10 CFR 50.68 and GDC 62.
4 The main guidance that the staff used for 5 its review is SRP Section 911. And there are several 6 topics in this review area as are listed on this 7 slide.
8 As a result of the staff's review, the 9 staff finds that the applicant's criticality 10 calculation methodology is acceptable and adequately 11 benchmarked. And the applicant's criticality models 12 correctly incorporate design information and use 13 appropriate assumptions.
14 There is an open item related to the 15 analysis. And that is that the applicant is making 16 the structural analysis of the racks, a COL item.
17 The implications of this COL item on the 18 criticality analysis is that a COL applicant would 19 have to show that any rack deformation or relocation 20 due to a seismic event or a fuel assembly drop would 21 be bounded by the existing criticality analysis.
22 And if it weren't, then the COL applicant 23 would have to re-analyze. Right now this is an open 24 item pretty much because the staff is working with the 25 applicant and OGC on the wording of the COL item.
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90 1 Other than that, the staff generally 2 accepts the approach.
3 So, other than the analyses that are 4 related to that COL information item, the staff finds 5 that the applicant's criticality analyses do comply 6 with 10 CFR 50.68 and GDC 62.
7 CO-CHAIR SKILLMAN: Alex, let me ask this 8 question, please.
9 MS. SIWY: Sure.
10 CO-CHAIR SKILLMAN: On the safety 11 evaluation on page 12, the safety evaluation 12 communicates about two thousand PTM.
13 And then on page 13 is the statement, DCA 14 Part Two, tier two, section 911 states that the large 15 volume of water, about seven million gallons by the --
16 in the UHS prevents an undetected boron dilution 17 accident from the normal -- I think it's nominal, 18 nominal boring concentration of about 18 hundred to 19 below eight hundred.
20 My question, is it two thousand or 18 21 hundred?
22 MS. SIWY: I'd have to double check the 23 table in the DCA. I think --
24 CO-CHAIR SKILLMAN: I thought it was two 25 thousand.
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91 1 MS. SIWY: Yeah. I think they're aiming 2 for two thousand. But I think there is one table in 3 the DCA, in chapter 9 that says at least 18 hundred.
4 CO-CHAIR SKILLMAN: Thank you. All right.
5 Thanks.
6 MS. SIWY: Okay. Yes, next slide. So, 7 the staff also conducted an independent confirmatory 8 analysis that support the applicant's conclusions with 9 regards to the acceptance criteria.
10 The staff also notes that the tier one 11 information is complete and consistent with tier two.
12 And that the Tech Specs related to fuel storage 13 adequately protect the assumptions in the criticality 14 analysis.
15 With regard to materials for fuel storage, 16 the staff notes that NuScale has not specified a 17 neutron absorbing material. And there is a related 18 open item in which the staff requested additional 19 details on the impacts of manufacturing on neutron 20 attenuation, materials qualifications for the spent 21 fuel pool environment, and manufacturing process 22 controls.
23 The staff and applicant recently agreed on 24 a path forward for this open item by creating a COL 25 item to address the staff's concerns. And like with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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92 1 the other open item, the staff and the applicant are 2 working on wording of this right now.
3 The staff also notes that the applicant is 4 implementing a neutron absorber monitoring program 5 that's consistent with TSTF-577, Revision 1. And 6 utilizes the staff approved NEI 16-03 guidance for 7 monitoring a fixed neutron absorber and spent fuel 8 pools.
9 And this provides the staff reasonable 10 assurance that degradation of neutron absorbing 11 material can be adequately detected so that the 12 neutron absorber continues to provide the credit 13 criticality control.
14 That concludes my presentation. Are there 15 any questions?
16 MR. TESFAYE: Thank you Alex. And Section 17 9.1.2, new spent fuel storage, there was one open item 18 related to the fuel gage protection system. The 19 information we received was provided by NuScale. And 20 we have since closed that open item.
21 And for the next sections, we don't have 22 any open items. The slides are presented. If you 23 have any questions, we'll be glad to address them.
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93 1 I would like to jump to that slide.
2 MEMBER CORRADINI: No. I think you're 3 going the wrong direction.
4 MEMBER BLEY: On 9.1.2-1, RAI response, 5 you say that one's closed now. Did you investigate 6 how they ensure they have mixing in the -- in fuel 7 pool so the boron concentration is the same 8 everywhere?
9 MR. TESFAYE: Okay. Raul, would you 10 please address that question?
11 MEMBER BLEY: And hot just the pumping, 12 but how they do sampling to make sure it really is 13 mixed.
14 MR. HERNANDEZ: This open item is not 15 related to the boron solution issue. The open item in 16 front of us is related to leakage detection.
17 And the spent fuel pool and the ultimate 18 heat sink having an adequate leakage channel behind 19 the walls to detect leakage. And the latest response 20 they added.
21 For the boron solution issue, I don't have 22 the exact location of the sampling point. So I cannot 23 talk about that.
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94 1 different points on the pool that would allow the flow 2 of water to increase mixing and help prevent having 3 stagnant areas of water.
4 MEMBER BLEY: How are you convinced they 5 don't have any stagnant areas?
6 I've seen lots of tanks where people were 7 surprised when they finally sampled at a different 8 place then they'd been sampling and found very 9 different concentrations.
10 MR. HERNANDEZ: Based on the drawings and 11 -- I have not -- I don't expect there to be one. I 12 don't have any specific details to give you right now 13 at this moment.
14 MEMBER BLEY: So you don't know of really 15 having mixing.
16 MR. HERNANDEZ: At this moment I cannot 17 answer the question specifically.
18 MEMBER BLEY: Okay.
19 MEMBER CORRADINI: Is there --
20 MEMBER BLEY: I'm sorry.
21 MEMBER CORRADINI: Oh.
22 MEMBER BLEY: Go ahead Mike.
23 MEMBER CORRADINI: Well, I just wanted to 24 ask, there's a COL information item, 9.1-1. I guess, 25 and it was about procedures relevant to handling of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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95 1 storage of spent and new assemblies and criticality 2 control.
3 This isn't related. But in some sense, 4 their leak -- as I understood when we were asking 5 NuScale, they're leaving it to the operator to develop 6 procedures to determine that they have good sampling.
7 And it is homogeneous.
8 That's how I interpreted.
9 MEMBER BLEY: I didn't hear them say that.
10 They said, yeah, the operator will have to figure out 11 where they're going to sample.
12 But, nothing about we have been mixing, we 13 have checked for stagnant areas. We make sure you've 14 got the uniform solution.
15 I didn't hear anybody say that that was a 16 requirement that was laid out.
17 MEMBER MARCH-LEUBA: I guess my question 18 is more for Alexander who is the technicality person.
19 Is there any place in the UHS where we make ready for 20 the boron other than the fuel suppression at the spent 21 fuel pool?
22 MS. SIWY: I specifically focused on the 23 spent fuel pool part. The other instance I could 24 think of is during refueling.
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96 1 don't think there is any transient that results in 2 water around the pool that inside the vessel.
3 And the containment is always intact.
4 MS. SIWY: Um-hum.
5 MEMBER MARCH-LEUBA: So, just a 6 recommendation, if I were going to sample it, I would 7 sample it in the spent fuel pool drain line, the 8 suppression line. That's where I would make sure that 9 the boron is within specs, because that's where it's 10 needed.
11 But that's my opinion.
12 CO-CHAIR REMPE: I believe there's someone 13 from the audience who had a comment? Oh, okay.
14 Sorry.
15 MR. TESFAYE: Okay. Are we ready to move 16 on?
17 CO-CHAIR SKILLMAN: Getachew, what 18 challenge, if any, did the staff raise about the 19 adequacy of the thickness of the pool liner floor?
20 MR. TESFAYE: Yes. We have that question 21 for. Raul, do you have the answer to that?
22 MEMBER BLEY: If we speak more directly in 23 the microphones, everybody can hear you better. It's 24 surprising how bad the acoustics are.
25 MR. TESFAYE: Yes. Okay, I'm sorry.
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97 1 MR. HERNANDEZ: Can you repeat the 2 question, please?
3 MR. TESFAYE: That how confident are we --
4 CO-CHAIR SKILLMAN: The question is, what 5 consideration did the staff give to the adequacy of 6 the thickness of the pool liner floor?
7 There's a stainless steel liner down 8 there. It sounds like it's a quarter of an inch, 9 maybe half an inch. And there are going to be 10 multiple 730 ton loads --
11 MR. HERNANDEZ: Okay, --
12 CO-CHAIR SKILLMAN: Transported over or 13 dragged over that very thin membrane.
14 MR. TESFAYE: Okay.
15 MR. YESHNIK: This is Andrew Yeshnik. I'm 16 a materials engineer in the Office of New Reactors.
17 CO-CHAIR REMPE: Speak up.
18 MR. YESNICK: Closer? Oh, sorry. That 19 was not credited. The liner itself is not credited 20 with any sort of structural function in the case of a 21 drop.
22 The drop is precluded so it's not part of 23 the design.
24 MEMBER MARCH-LEUBA: So if it drops, it 25 breaks? Or it has a high probability of breaking? Or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 1 at some point --
2 MR. YESHNIK: It will probably deform, 3 yeah.
4 MEMBER MARCH-LEUBA: Huh?
5 MR. YESHNIK: I mean, it's a very heavy 6 load.
7 MEMBER MARCH-LEUBA: The difference --
8 MR. YESHNIK: I have not seen the analysis 9 of what that load specifically is, but.
10 MEMBER MARCH-LEUBA: I heard you say 11 earlier something I liked very much. There is a leak 12 detection system.
13 MR. YESHNIK: Um-hum.
14 MEMBER MARCH-LEUBA: I'm pretty sure there 15 is one on the spent fuel pool. But is there one on 16 the UHS?
17 MR. YESHNIK: Yes.
18 MEMBER MARCH-LEUBA: So there are channels 19 underneath, if it breaks we will see it before the 20 pool starts dropping?
21 MR. YESHNIK: No. Those channels are 22 designed to detect any leakage of the wells. Not the 23 entirety of the UHS.
24 MEMBER MARCH-LEUBA: Oh, so it's not an 25 out leakage for this in use?
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99 1 MR. YESHNIK: Um --
2 MEMBER MARCH-LEUBA: Ignore me.
3 MEMBER CORRADINI: If the well is cracked, 4 it's swept along the weld line.
5 MR. YESHNIK: That's correct.
6 MR. HERNANDEZ: I just wanted to add that 7 and this is not unique to this design. There are in 8 other designs, the liner itself, it doesn't have the 9 function of retaining the water.
10 The function is basically to protect the 11 concrete. The concrete structure is the one that is 12 holding the water and is designed to maintain the 13 integrity.
14 We do monitor the wells between the plates 15 to ensure that there is no long term degradation of 16 the concrete because of all known leakage.
17 But, like Andrew just said, the liner is 18 not credited to retain the water. It's the concrete 19 structure.
20 CO-CHAIR SKILLMAN: I assure you my 21 question was not aimed at suggesting the liner has any 22 structural component whatever. The liner is nothing 23 more than a membrane.
24 But if that membrane is injured because of 25 whatever reason, that is a common mode failure for 12 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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100 1 live core reactors.
2 Now let me say that again. If you tear 3 that liner, if you drop a module, if you drop a fuel 4 assembly, if you drop a fixture, if you drop a reading 5 device onto a one-quarter inch thick membrane that is 6 well supported, I suspect that there will be a 7 puncture wound.
8 And you will have to do something to 9 repair that liner. And that's a common mode failure 10 on a single pool that's cooling potentially 12 live, 11 160 megawatt reactors.
12 That's a problem. And that means that 13 floor liner should probably be more then a one-quarter 14 inch membrane.
15 It should be either armored for the load 16 path. Or otherwise precluded from sustaining a 17 puncture wound.
18 One man's opinion.
19 MR. TESFAYE: I think we've taken a note.
20 So we'll look into that.
21 CO-CHAIR SKILLMAN: So the answer to my 22 question is, the staff did not consider beyond a 23 quarter of an inch or whatever that membrane thickness 24 is.
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101 1 the answer is no.
2 MR. TESFAYE: No.
3 CO-CHAIR SKILLMAN: Okay. Thank you.
4 MR. TESFAYE: Thank you. Any other 5 questions on this?
6 (No response) 7 MR. TESFAYE: If there are no questions on 8 the next Section 9.1.3 spent fuel pool cooling and 9 clean up system and light load handling and systems, 10 I'd like to move on to the next section that has an 11 open item.
12 We have no open items on water systems.
13 Okay. Ed, let's move onto.
14 MR. STUTZCAGE: Hi, I'm Ed Stutzcage.
15 I'll be present 9.3.2. Which is process and post-16 accident sampling systems.
17 The purpose of the process imposed 18 accident sampling system is similar to other designs.
19 It is to collect and analyze the chemical and 20 radiochemical conditions from the primary and 21 secondary systems and other systems including the 22 containment atmosphere. Next slide, please.
23 There's no dedicated post-accident process 24 sampling system used in the NuScale design. The 25 normal process sampling system is used during post-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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102 1 accident conditions.
2 I'll be focusing mostly on post-accident 3 sampling, because we had no open items for normal 4 sampling.
5 For post-accident sampling, we actually 6 had seven open quest -- open items in Chapter 9. And 7 nine open questions in Chapter 12.
8 They're all related to the post-accident 9 sampling process, which is unique to NuScale's design.
10 And you can go to the next slide, please.
11 So in NuScale's design to collect liquid 12 samples during an accident, you'd have to un-isolate 13 containment, and send the fluid through the chemical 14 and volume control system. Through the sampling line 15 where it would, you know, it would be sent through the 16 sampling point and collected.
17 For gaseous samples, it would also have to 18 -- the containment would have to be un-isolated.
19 Liquid would have to be sent through the containment 20 evacuation system to the sampling system and then 21 returned to the containment through the containment 22 flood and drain system.
23 There were numerous staff concerns related 24 to this. Particularly related to the dose, to workers 25 taking the samples and also the system operation and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 1 so on and so forth.
2 NuScale has recently requested an 3 exemption from the requirements of 10 CFR 4 50.34(f)(2)(8), which are requirements to take post-5 accident samples and within specific dose limits.
6 Instead they would -- they're requesting 7 to rely on continuously planning to take samples if 8 they needed to.
9 And part of their basis for this is, they 10 have other means of providing the inf -- of getting 11 the information that sampling would provide, such as 12 under the bio shield radiation monitors for 13 radiochemical conditions, they have hydrogen monitors.
14 However, NuScale's hydrogen monitors are 15 on their sample lines. So they'd have to use a 16 similar process of opening the containment evacuation 17 system to monitor for hydrogen.
18 The exemption request is somewhat similar 19 to relaxations applied to the operating fleet in 20 several topical reports. But like I said, there are 21 some unique features in the NuScale design that are 22 being considered.
23 So, we're still evaluating the exemption 24 requests.
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104 1 one? Because I didn't understand it. It's because 2 you're operating at low pressure within the 3 containment, and that creates an issue of how to take 4 the sample?
5 MR. STUTZCAGE: There's different 6 challenges depending on what the conditions are inside 7 containment. There's a possibility and my 8 understanding that you would have to send nitrogen 9 into the system to push the gaseous fluid through.
10 And there's also potential challenges of 11 liquids. If the water in the containment is too low, 12 depending on where you're at in the accident, and what 13 the accident is, you may have to do things to raise 14 the water level.
15 So, there's -- like I said, there's -- it 16 kind of depends on the situation and conditions.
17 MEMBER CORRADINI: So, their contingency 18 -- would you help me out here? I should have asked 19 them and I forgot.
20 Their contingency is to do what in terms 21 of sampling? Or is it yet to be determined because 22 we're in the middle of the RAI and the open items?
23 MR. STUTZCAGE: NuScale's approach is that 24 they have other means to detect the information that 25 sampling could provide them. However, if they really NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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105 1 needed to, and they chose, it would be beneficial to 2 take a sample.
3 They can, you know, they have the 4 capability to do it. It's just they're only going to 5 rely on it if they determine it was, you know, 6 completely -- if they determine it to be beneficial.
7 They don't want to have a requirement to 8 have to be able to do it within, you know, a short 9 amount of time following an accident.
10 MEMBER CORRADINI: But this is -- this RAI 11 is still going to remain open. This is --
12 MR. STUTZCAGE: Yeah --
13 MEMBER CORRADINI: Is 90.44.
14 MR. STUTZCAGE: All of the -- yes. All of 15 the -- right. All of those questions remain open as 16 we evaluate the exemption request.
17 And then yeah, we make a determination on 18 that -- determination on that and we'll figure out 19 what to do with that RAI.
20 MEMBER CORRADINI: Thank you.
21 DR. SCHULTZ: Is there a schedule for that 22 Ed?
23 MR. STUTZCAGE: I can't say we have a 24 definitive schedule at this point. Yeah, it came in 25 at the very end of January and we've -- yeah.
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106 1 DR. SCHULTZ: Got you. Thank you.
2 MR. STUTZCAGE: Any other questions?
3 (No response) 4 MR. TESFAYE: Okay. Thank you. Next 5 slide is on CVCS. Ryan Nolan will be presenting that.
6 MR. NOLAN: Yes. My name is Ryan Nolan.
7 I was one of the reviewers for the CVCS system.
8 And we don't have any open items in this 9 review area. However, there is an exemption request 10 that I'll spend a little bit of time discussing, as 11 well as some previous ACRS questions that we received.
12 I'll discuss some of those as well. So if you go to 13 the next slide.
14 So NuScale requested an exemption to GDC 15 33 which requires having a reactor coolant makeup 16 system to address for small breaks in reactor coolant 17 pressure boundary.
18 And NuScale's design doesn't rely on any 19 injection of ECCS from external tanks for adding 20 inventory to the RCS. Instead, they rely on isolation 21 of the system, inventory retention.
22 And then the ECCS and the containment 23 vessels maintain the core coolability to meet the 24 SAFDLs. And the way the staff approached this is we 25 looked at it from both normal operation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 considerations, as well as off normal.
2 And so for normal operations, we would 3 assume that CVCS is available to control inventory.
4 And then for off normal considerations NuScale relies 5 on the safety actuation isolation signals that would 6 isolate CVCS as well as potentially containment. And 7 then rely on ECCS and containment to meet the SAFDL 8 requirements.
9 That's the finding we made on the 10 exemption to 33.
11 MEMBER CORRADINI: If -- this is a kind of 12 an off question. If they chose to un-isolate 13 containment and they isolate a wait use list, is that 14 selective?
15 Or is it un-isolated in all valves such 16 that if they wanted to operate ECC -- CVCS selectively 17 open the valves in there.
18 MR. NOLAN: I'm not as familiar in like 19 recovery operation aspects of how they would actually 20 do that. I would have to go back and look at the 21 design.
22 MEMBER CORRADINI: Yeah. But it's not in 23 the plan of how they want to address any sort of 24 accident situation. They go to isolation and --
25 MR. NOLAN: Correct. And rely on --
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108 1 MEMBER CORRADINI: To delay the high 2 pressures and low pressure ECCS.
3 MR. NOLAN: Yes. And rely on ECCS for the 4 long term cooling requirements.
5 MEMBER MARCH-LEUBA: Yeah, we -- we would 6 look at these events during Chapter 15 review a little 7 more. Because Chapter 15 is missing a few events.
8 MR. NOLAN: Um-hum.
9 MEMBER MARCH-LEUBA: One of them is normal 10 shutdown. And under normal shutdown, without CVCS 11 completely isolated, the reactor will start cooling 12 down by itself. And it will start shrinking.
13 MR. NOLAN: Um-hum.
14 MEMBER MARCH-LEUBA: The water level will 15 drop to 63 percent of volume. And I would like to see 16 what happens then.
17 I mean, during normal operation they will 18 have CVCS maintain the inventory in the vessel. And 19 the vessel is going to drop significantly in liquid.
20 And I have not seen that analysis to 21 ensure that it is acceptable.
22 MR. NOLAN: Well, we'll go to the next 23 slide. We've received some questions --
24 MEMBER MARCH-LEUBA: Um-hum.
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109 1 flow boron mixing phenomenon that we've seen in 2 boiling water reactors.
3 And I guess the phenomenon is at really 4 low flows, when you put boron into the lower vessel, 5 the boron has a tendency to settle out and not make 6 its way into the core.
7 And so in this drawing, what I'm trying to 8 illustrate here is that the CVCS injection line is 9 actually above the core. It injects eight feet above 10 the core.
11 And so if there is a phenomenon of low 12 flow or boron potentially setting, it actually has to 13 go into the core first before.
14 And this is just sort of teeing up a 15 discussion in Chapter 15. This is a thermal hydraulic 16 phenomenon that we're currently looking at.
17 Boron dilution in long term, that's 18 evaluated in Chapter 15. And then the mixing 19 methodology that NuScale has used for the dilution 20 events is evaluated in 15.4.6.
21 MEMBER MARCH-LEUBA: My point is, of 22 course the vessel will never reach the temperature on 23 the UHS. But, it might get close.
24 Because after, a mounting there eventually 25 will reach. If you start at 1850 psi, the density of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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110 1 the vessel is 0.63 grams persistent. At the end of 2 the operation, this is 1.
3 So if you have complete isolation of 4 containment the water level will drop about 63 percent 5 of what it was inside the vessel. So the operator 6 will start CVCS eventually. And they will put a lot 7 of water.
8 And that water better not be un-borated.
9 And I don't see anywhere that that event is described.
10 I don't see any event in Chapter 15 or Chapter 19.
11 Maybe it is in Chapter 19.
12 I'm just putting it on the record, that 13 there are events that I have not seen anywhere. Thank 14 you.
15 MR. NOLAN: One more point I wanted to 16 make. Member Skillman, you mentioned earlier this 17 morning about -- you questioned the capability of 18 putting boron into the core and why is CVCS not safety 19 related, or questioning the pedigree.
20 And I'll just point out that that's a --
21 the staff looks at that as part of GDC 26 and 27. And 22 that's eval -- GDC 26 is evaluated in Chapter 4. So, 23 we'll have more discussions on that.
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111 1 using diverse principals for controlling reactivity.
2 First is control rods, which are safety related.
3 And then the second one is for controlling 4 rate of reactivity changes. And that's for our plan 5 in normal operations.
6 And so in this case, CVCS could be 7 credited for meeting that portion of the GDC. It 8 doesn't specify that it has to be safety related.
9 And then GDC 27, NuScale is seeking an 10 exemption for that. And that's evaluated in Chapter 11 15.
12 CO-CHAIR SKILLMAN: Yeah. And what kind 13 of drove my comment this morning is my recognition 14 that you -- that NuScale is seeking an exemption from 15 GDC 27 and 33.
16 And those are basically key structural 17 components for PWRs in terms of shutdown. And so here 18 we have a proposed new design where the designer is 19 saying, I really don't need that boron.
20 And guess what, I really don't need a make 21 up system for a small break. I don't need the boron 22 because I've got another way to shut it down. And I 23 don't need Generalized Design Criteria 33 because the 24 only small breaks I'm going to have are in the steam 25 system.
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112 1 And I'm choking on that. It would seem to 2 me that a responsible reactor operator would say, I 3 don't care what those words are. I want a backup 4 shutdown system. And I want a pump that's going to 5 put that boron in there when I tell it to go in there.
6 It would just seem to me that that would 7 have been the construct for a passive design. And the 8 addition of the boron does not need to be high volume.
9 It just needs a little bit.
10 That's why we have the slip. That's why 11 the newer reactor designs have direct vessel 12 injection.
13 Well, to me it's just counterintuitive.
14 And it seems like a very simple adjustment to what is 15 a relatively robust design. Thank you.
16 MR. NOLAN: I understand. A lot of the 17 long term review areas, that's something that we look 18 at in Chapter 15.
19 And so, I'm sure we'll have lots of 20 discussions on that in the near future.
21 CO-CHAIR SKILLMAN: Thank you.
22 MR. TESFAYE: Okay. Any other questions 23 to Ryan?
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113 1 we don't have any other open items in the rest of 2 Section 9.3. And Section 9.4 HESE, we don't have any 3 open items.
4 So, if you don't have any questions on 5 those sections, I'd like to move onto the next section 6 that has open items. Which is in Section 9.5.3, 7 lighting. Fanta?
8 MS. SACKO: Okay. My name is Fanta Sacko, 9 and I'll be discussing Section 9.5.3, lighting 10 systems.
11 The NuScale design plan lighting system is 12 a non-class one system that is not safety significant.
13 Not risk significant and non-safety related.
14 And it includes normal pin lighting, 15 emergency can lighting, normal and emergency main 16 control lighting.
17 The normal pin lighting provides 18 illumination on the plant site and for plain 19 buildings. The emergency plan lighting provides 20 illumination outside the control room or for loss of 21 normal lighting.
22 The normal and emergency main control room 23 lighting provides illumination under all operating 24 MENs testing and emergency conditions in the MCR.
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114 1 whether the plan levels are adequate in all ten areas.
2 And whether the PLS can operate without adversely 3 impacting the operational control and maintenance of 4 FSCs.
5 We conducted these reviews in accordance 6 with SRP Section 9.5.3 and NUREG 0700. In summary, we 7 had one open item related to Reg Guide 175, which is 8 the physical separation between non-safety related 9 lighting circuits and safety-related circuits.
10 And we had one confirmatory item. So the 11 completion of the staff's review of the plain lighting 12 system is awaiting the completion of Chapter 8 open 13 item regarding GDC 17 and 18 exemptions.
14 So, like I said, the open item in this 15 section relates to the physical separation between the 16 non-safety lighting and the safety-related circuits.
17 Regarding the staff verifying that the 18 design does not require a safety-related power. The 19 confirmatory item of plan illumination levels will be 20 addressed in phase four of the FSAR.
21 This concludes this slide presentation.
22 MR. TESFAYE: Thank you Fanta. Any 23 questions to Fanta on lighting systems?
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115 1 presentation. But I'd like to add a couple of things 2 on schedule. There was a question on when are we 3 going to be completing the pass exemption requests.
4 Pass exemptions as all the other 5 exemptions, are reviewed in the Chapter that it's in.
6 And it's a phase four activity.
7 And phase four has begun for this chapter, 8 of it will be completed in accordance to the Chapter 9 9 schedule, Chapter 9/Chapter 12 schedule for phase 10 four. We do have a schedule.
11 CO-CHAIR REMPE: Thank you. And thank you 12 for a very succinct presentation. Because we're back 13 on schedule. I hadn't expected that.
14 MEMBER MARCH-LEUBA: I wanted to make one 15 question. Just to put it on the record, because you 16 don't know the answer of this.
17 I am looking at the design diagram of the 18 CVCS. Which is Figure -- it's on page 9.3. That's 19 84, Figure 9.3.4-1. Which is a diagram of the 20 chemical boron control system.
21 And it shows that there is a continuous 22 sampling line on the discharge that comes from the 23 vessel. So, we know what the concentration is in the 24 vessel by measuring this.
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116 1 the injection line, unless it's one of these heating 2 ones that isn't a sample.
3 MR. NOLAN: It's just downstream of the 4 regen heat exchanger. It's a normally closed.
5 MEMBER MARCH-LEUBA: The one for PSS?
6 MR. NOLAN: Yeah.
7 MEMBER MARCH-LEUBA: Okay. I take it 8 back.
9 CO-CHAIR REMPE: Thank you. If there 10 aren't any other questions, let's invite NuScale to 11 come back and discuss Chapter 16.
12 (Off mic comments) 13 CO-CHAIR REMPE: So Ross, are you ready to 14 begin?
15 MR. SNUGGERUD: Yes. All right. Hello, 16 good afternoon. My name is Ross Snuggerud. I have a 17 Master's degree in Nuclear Engineering.
18 I have been in the industry for 25 years.
19 I spent 15 of those at a PWR in Michigan. And I've 20 spent the last almost 11 years at NuScale.
21 MR. GROSS: My name is Carl Gross. I've 22 been in the industry for 39 years now, including old 23 licensing under Part 50 at Palo Verde.
24 I've worked in the operations and 25 licensing.
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117 1 CO-CHAIR REMPE: Carl, you're going to 2 have to get closer.
3 MR. GROSS: Worked in operations and 4 licensing out there. And have been mostly in 5 licensing and operating programs since then.
6 Worked at about 25 commercial reactors and 7 five or six DOE sites in nuclear safety and then TSR 8 development there.
9 MEMBER BLEY: Please try to stay closer to 10 the mic.
11 MR. GROSS: Sorry.
12 MEMBER BLEY: Because you disappear as 13 soon as you back off a little.
14 MS. FOSAAEN: Again, Carrie Fosaaen, 15 NuScale Licensing Supervisor for Chapter 16.
16 MR. SNUGGERUD: All right. So today we 17 are going to talk about NuScale's Chapter 16. Which 18 generally relates to the development of the technical 19 specifications that will be used by the COLA applicant 20 to submit specific technical specification for their 21 facility.
22 The technical specifications -- who runs 23 the slides? Do I? The technical specifications that 24 we've built are designed to support individual 25 modules.
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118 1 So, the tech specs will behave for the 2 operator much like they do in the current industry.
3 There would actually technically be 12 copies of the 4 technical specifications. Our intent would be that 5 they are all the same.
6 But they deal with the technical 7 specifications on a module by module basis. Shared 8 systems like the pool would potentially cause 9 actuations of the technical specifications across 10 multiple modules. But they would be addressed 11 individually.
12 Some of the things that are unique about 13 the development of technical specifications for the 14 NuScale design are the fact that we have a simpler, 15 natural circulation design without the primary coolant 16 pumps.
17 We have passive cooling systems. Both the 18 ECCS and the DHR system. We have a completely digital 19 module protection system.
20 We don't require any safety related AC 21 power or any need for offsite power which has an 22 impact on the tech specs. We don't have any credited 23 HVAC systems and no active support systems.
24 And we do move the reactor when we refuel.
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119 1 developing the tech specs.
2 Because the design is simpler and because 3 we have fewer systems that are incorporated in the 4 technical specifications, the ultimate document that 5 we generated is roughly half the size of the typical 6 set of instructions for a more conventional PWR 7 design.
8 In developing the technical 9 specifications, we used the criteria of 50 -- 10 CFR 10 50.36 to evaluate all the plant designs. Which was 11 largely a review of Chapter 15.
12 And we developed a topical report that 13 went through our evaluation of all of the designs.
14 This was unique for us because NuScale doesn't have a 15 predecessor plant.
16 So we didn't have a predecessor set of 17 tech specs that we could start and work to modify. We 18 started with a blank piece of paper.
19 So, we did a complete evaluation of the 20 plant. We used the existing NUREGs for the base 21 designs, both PWRs and BWRs as source input where 22 appropriate.
23 And we used an operations group that 24 incorporates -- I want to make sure I get the numbers 25 right, but we have 18 previously licensed SROs with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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120 1 over 560 years of experience in operating nuclear 2 facilities.
3 And that experience went into helping both 4 the development of the technical report and the 5 reviews of the technical specifications.
6 In addition, we followed the guidance of 7 the standard tech spec writer's guide. And we 8 monitored industry actions.
9 So, in addition to the standard NUREGs 10 that exist, we also followed tech spec task force 11 work. Both approved travelers and upcoming work that 12 is available to the public.
13 And evaluated whether or not those changes 14 were appropriate to the NuScale technical 15 specifications. And incorporated what was appropriate 16 to our design where that occurred.
17 One thing I didn't mention on the slide, 18 but there were -- there's four criteria for 50.36.
19 The fourth one is a criteria that's used because it's 20 assumed to be prudent, not because it necessarily 21 meets any of the prescribed requirements.
22 And there are two features of the NuScale 23 design that we included for prudency. That's the 24 manual actuation function and the remote shut down 25 function.
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121 1 I mentioned earlier, and the committee is 2 clearly aware that we do move the module during our 3 refueling process. And so that did change the way we 4 went about developing our mode definitions.
5 Sometimes the mode definitions get 6 confused because they're used very heavily in plant 7 operations. But, the reality is, the mode definitions 8 are chosen to make applicability of the Section Three 9 tech specs easier for the operator to apply.
10 So you want to pick transition points that 11 logically follow the plant design, and make it easier 12 for the operator to know which sets of technical 13 specifics apply to any given mode of operation.
14 As a result, we got rid of the classic 15 mode two. Which is the low power hot mode. Because 16 at NuScale there were no unique requirements that only 17 applied to that mode of operation. So that mode was 18 removed.
19 We added the mode transition which applies 20 to the reactor module while it's being moved and prior 21 to decoupling any of the bolts associated with the 22 lower portion of the reactor vessel.
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122 1 associated with the reactor flange, that's the 2 transition mode.
3 And then finally, we have a refueling 4 mode. Which is not unlike the existing industry.
5 Just for clarity, the CRA is a control rod 6 assembly, see --
7 MEMBER RAY: Wait a minute. You have a 8 refueling load that's not unlike existing industry?
9 MR. SNUGGERUD: At the point that --
10 MEMBER RAY: What do you mean by unlike --
11 MR. SNUGGERUD: At the point that we're in 12 refueling, the primary things that the tech specs are 13 concerned about, are monitoring for reactivity changes 14 in the core.
15 MEMBER RAY: Just tell me what you meant 16 by that statement.
17 MR. SNUGGERUD: So, the kinds of things 18 that are required in our Mode Five, look a lot like 19 the kinds of things that are required in a refueling 20 Mode Five for other plants.
21 Obviously, ours is compressed. But it's 22 talking about the same types that evolutions is 23 talking about, picking up and moving fuel inside a 24 core. And what's necessary to ensure that that's a 25 safe activity.
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123 1 I understand that my choice of words may 2 not have been appropriate.
3 MEMBER RAY: Those words were not 4 appropriate. Period.
5 MR. SNUGGERUD: From what basis? Help me 6 understand why you feel that strongly.
7 MEMBER RAY: Well, you've done refuelings 8 before. You're moving fuel elements one at a time in 9 and out.
10 This is moving an entire core and a 11 reactor and its vessel with a core, you know, the 12 entire core inside of it. That's in and of itself 13 completely different.
14 MR. SNUGGERUD: I agree. And that's the 15 transition mode. Which is unique and new to NuScale.
16 I transitioned from the transition mode, 17 which incorporates all those activities you just 18 talked about, to the refueling mode, which looks much 19 more --
20 MEMBER RAY: Oh, after you get this --
21 MR. SNUGGERUD: Yes.
22 MEMBER RAY: I apologize. My fault.
23 MR. SNUGGERUD: I just want to make sure 24 we are talking trans --
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124 1 moving the reactor vessel when you said that.
2 MR. SNUGGERUD: Yeah. I'm sorry.
3 MEMBER RAY: Okay.
4 MR. SNUGGERUD: The transition mode is 5 completely unique to NuScale.
6 MEMBER RAY: Okay.
7 CO-CHAIR SKILLMAN: Far right column.
8 Indicated reactor coolant temperature, you reach 9 Fahrenheit.
10 What is the distinction between all versus 11 any? What do those words mean, please?
12 MR. SNUGGERUD: So, there is clarification 13 in the basis document. We're talking about all safety 14 related T colds and T hots.
15 Or any safety related T cold or T hot 16 since you can't always guarantee that the hots will be 17 hotter then the colds.
18 CO-CHAIR SKILLMAN: Okay. Thank you.
19 MEMBER MARCH-LEUBA: And mostly you said 20 that any temperature indicated that this is less than 21 420? Is that --
22 MR. SNUGGERUD: All temperatures have to 23 be less the 420 in order for you to be in Mode Three.
24 MEMBER MARCH-LEUBA: Okay.
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125 1 pieces of equipment that are no longer required. But 2 you have to prove that you've reduced the energy to 3 below 420 at all locations. All operable indications.
4 MEMBER MARCH-LEUBA: And do you also have 5 to guarantee that you -- the effect is .99? Because 6 when you get really, really cold, you may not satisfy 7 .99.
8 MR. SNUGGERUD: So, the --
9 MEMBER MARCH-LEUBA: How do you know where 10 you are? Because what I'm saying is, I'm an operator 11 and would like to know what temperature would be a 12 threshold that would make me unsafe.
13 MR. SNUGGERUD: So we would have a 14 guidance document not unlike what the PWR that's used 15 in the industry that would tell the operators for a 16 given temperature what the required boron 17 concentration is.
18 It is demonstratable that under most cases 19 the boron concentration necessary to meet this 20 requirement is consistent with the operating boron 21 concentration. Assuming that all of the control rods 22 have fully inserted.
23 For refueling though, obviously we haven't 24 --
25 MEMBER MARCH-LEUBA: At any temperature?
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126 1 MR. SNUGGERUD: Right. You can go to a 2 much lower temperature. And our expectation is that 3 there would be the addition of boron to the system to 4 do a normal plant shutdown to meet that requirement.
5 MEMBER MARCH-LEUBA: I got that impression 6 too.
7 MR. SNUGGERUD: Yeah.
8 MEMBER MARCH-LEUBA: What we're concerned 9 is, what equipment are you going to use to accomplish 10 it?
11 MR. SNUGGERUD: We will be using the CVCS 12 system.
13 MEMBER MARCH-LEUBA: A non-safety related 14 CVCS?
15 MR. SNUGGERUD: Yes. Just like existing 16 facilities.
17 MEMBER MARCH-LEUBA: You might want to put 18 that on the record.
19 MEMBER KIRCHNER: I'm just puzzled. Why 20 do you say our expectation? Why isn't this more 21 definitive?
22 You could be at the end of a burn up cycle 23 of a module and down to, if I remember right from 24 Chapter Four, somewhere close to 10 ppm when you 25 finish that cycle.
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127 1 So, I would say that boron injection is 2 required, not expected. Because your -- beyond the 3 criteria --
4 MR. SNUGGERUD: The design is designed 5 such that when all the rods insert, we maintain 6 shutdown margin at all temperatures.
7 But, there are specific cases, and if you 8 assume a rod missing and those types of things, where 9 you would need slightly more margin.
10 In addition, our refueling process will 11 involve removing control assemblies. And some of 12 those assemblies will have control rods.
13 So the process includes margin for that 14 boron. And because we want to ensure that the boron 15 at the bottom of the vessel meets the boron 16 requirements for the refueling operations, we would be 17 adding boron during the shutdown process to ensure 18 that.
19 I'm trying to be a little bit careful 20 because the reactivity --
21 MEMBER KIRCHNER: But we asked -- where 22 there --
23 MR. SNUGGERUD: That's where the 24 expectation comes in. Because technically --
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128 1 requirement to do it safely. It's not an expectation.
2 MR. SNUGGERUD: I understand your point.
3 MEMBER KIRCHNER: Yeah. And therefore I 4 would ex -- there I would use the word expected to be 5 specifically clear either in tech spec space or 6 procedures. Preferably tech spec.
7 MR. SNUGGERUD: The tech specs do provide 8 requirements for concentrations of boron in shutdown 9 modes.
10 MEMBER KIRCHNER: Right.
11 MR. SNUGGERUD: And what the required 12 margins are. So if those couldn't be demonstrated 13 with the boron concentration that had been exemplified 14 during operation, additional boron would need to be 15 added.
16 And we have the capability of doing that 17 using the chemical volume control system.
18 MEMBER KIRCHNER: By using tech spec 19 space.
20 MR. SNUGGERUD: It is in tech spec space.
21 MEMBER KIRCHNER: So there's no longer an 22 expectation, it's defined.
23 MR. SNUGGERUD: Fair enough. Yes. The --
24 I mean, the mode definition is a piece of that. The 25 boron requirements within tech spec are an additional NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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129 1 piece of that.
2 MEMBER MARCH-LEUBA: We will review and --
3 we will be reviewing Chapter Four next month, I 4 believe.
5 And I'm just giving you something -- maybe 6 you could use some feedback to your colleagues. We 7 hear that there will be a lot of questions about 8 certain margins.
9 And it would be nice if they have some 10 backup calculations that we can see. So, some actual 11 numbers and actual calculations with actual 12 assumptions. Not just statements of expectation.
13 MR. SNUGGERUD: Understood. You'd be more 14 interested in what those numbers look like under 15 different circumstances with different assumptions.
16 MEMBER MARCH-LEUBA: Not just the numbers, 17 but how you arrived at those numbers. Because 18 shutdown margin is the most important thing in a 19 nuclear reactor.
20 I guess being an operator, you know that.
21 Right? It cannot be an expectation. And it has to be 22 well analyzed, very -- and how with uncertainty 23 determination for it.
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130 1 up there?
2 I was just going to clarify that CRA 3 stands for control rod assembly. CVCS you probably 4 recognize as chemical volume control system. And CFDS 5 is the containment flood and drain system.
6 Part of what those are touching on are 7 things that could potentially change the concentration 8 of the water inside the module after you've 9 established the boron concentration you intended to 10 have in place for transition.
11 CO-CHAIR SKILLMAN: Is there a need --
12 this is an open question. And I don't -- I don't have 13 an expectation for an answer.
14 But, is there a -- should there be an 15 additional mode for NuScale where one of the modules 16 or several are somehow compromised or in casualty?
17 These are operating modes. Most of us who 18 have been around the industry kind of recognize this.
19 But, in this plant with 12 live cores, is 20 there another mode that is -- ought to have its own 21 unique name tag so that the leadership of the plant 22 and the operators, the men and women that are working 23 there, recognize this is different.
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131 1 ventilation. Some other kind of requirement.
2 This is not a norm -- that's not the right 3 word. This is so different. Does it deserve some 4 kind of a moniker where you say, you know, this one's 5 -- it's not in mode one, two, three, four, or five.
6 This is in mode 12. And mode 12 means special rules 7 apply.
8 MR. SNUGGERUD: So for the unaffected 9 modules, the answer is no.
10 CO-CHAIR SKILLMAN: And I agree with that.
11 Yeah.
12 MR. SNUGGERUD: So, for the affected 13 module, and this is true for existing facilities, when 14 you're in an accident, then you're in your emergency 15 procedures and you're following your emergency 16 procedures.
17 The technical specifications are designed 18 to ensure that the plant started prior to the accident 19 in a condition that allowed it to be in an analyzed 20 position for the accident that occurred.
21 To the extent they can, operators will 22 continue to try and maintain the plant within the 23 requirements of technical specifications.
24 But, once you get into an accident, the 25 function of the technical specifications to some NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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132 1 extent, has been met.
2 CO-CHAIR SKILLMAN: But accident has a 3 unique meaning also.
4 MR. SNUGGERUD: Absolutely.
5 CO-CHAIR SKILLMAN: So, maybe it's not an 6 accident. Maybe you've got a module that's stuck.
7 Your cable's failed.
8 Where the module is located is precluding 9 some other operation. Or it's over in one of the 10 storage stands. Your cable's unreaved.
11 You've got six of a number of bolts or 12 detention. So, it isn't a full casualty. You don't 13 have an accident. You might not be leaking 14 radioisotopes.
15 But there is an alert in the plant where 16 folks say, things are not right right now. We've got 17 to get this situation modified. But it's none of 18 these modes.
19 Don't answer.
20 MR. SNUGGERUD: You have a --
21 CO-CHAIR SKILLMAN: I'm just --
22 MR. SNUGGERUD: I've kind of given what I 23 can answer.
24 CO-CHAIR SKILLMAN: Yeah. There's a --
25 you're in a different world in this plant.
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133 1 MR. GROSS: Yeah. The role of tech specs 2 though is not off normal conditions such as that.
3 It's to keep the plant, as he mentioned, aligned with 4 the assumptions before the event occurs and so that it 5 will respond as appropriate if it hap -- if something 6 occurs.
7 There is one different one, and that's of 8 course to identify leakage from the RCS system. The 9 -- those, the conditions you're describing would be 10 off normal emergency procedures.
11 By definition if it affected the tech --
12 if it affected the safety analysis, we would have to 13 take action under the emergency plan of procedures.
14 If a module is just not movable, it really 15 doesn't have an effect. There's no special actions to 16 take before we began moving or entering or changing 17 conditions, all the other modules would have to comply 18 with their tech specs.
19 And for example, the shutdown margin, the 20 mode four transition definition, there --
21 MEMBER KIRCHNER: I think there might be 22 a -- I don't know if you were thinking this. But, I'm 23 thinking that if you build your plant and say install 24 half the module, so you have six.
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134 1 four in. There are a few instances where you do have 2 shared systems. Boron addition comes to mind and 3 such.
4 So, where you have special requirements in 5 terms of isolation as you tap into those systems. So 6 --
7 MR. GROSS: I don't believe any of those 8 systems other then UHS are credited.
9 MEMBER KIRCHNER: Um-hum.
10 MR. GROSS: So, as long as we maintain 11 water in the pool with the appropriate boron and the 12 right temperatures, we've met our requirements for our 13 safety analysis, which is what Chapter -- or what tech 14 specs are driven to protect.
15 CO-CHAIR REMPE: Well, which mode is when 16 you add another module? Is it transition?
17 Where am I when I'm bringing in a new one?
18 MR. SNUGGERUD: Yeah. So when you're 19 bringing in a new module, it's not fueled yet.
20 So, it wouldn't technically be in 21 technical specifications at that point. Once you put 22 the lower pieces into the plant, and assuming you had 23 completed whatever hot -- or start up testing and 24 validation you had completed.
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135 1 be in refueling.
2 CO-CHAIR REMPE: So, the other -- you've 3 divided it up. And I've forgotten the term you had.
4 MR. SNUGGERUD: It's in the inspection 5 station. Yeah.
6 CO-CHAIR REMPE: And then you put the fuel 7 in there and --
8 MR. SNUGGERUD: No.
9 CO-CHAIR REMPE: Then bring it down. Are 10 you bring -- what --
11 MR. SNUGGERUD: So, when you bring a new 12 module in, obviously there's some testing that we're 13 going to do before we ever put any fuel in it. Right?
14 We're going to do some training work and 15 those kinds of things. But you get to the sixth, 16 seventh module, you bring the components in.
17 You bring the lower containment vessel in.
18 Low lower reactor vessel in, and then the upper 19 reactor vessel and the upper containment.
20 The lower reactor vessel goes in its tool.
21 The lower containment goes in its tool. At that 22 point, with the upper module in the inspection bay, 23 there's no difference between that and a refueling.
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136 1 entering mode five for refueling. And the equipment 2 required to support refueling activity would have to 3 be operable.
4 CO-CHAIR REMPE: And then you transition 5 it back to its station. But -- and you hook things up 6 just like you would the others.
7 MR. SNUGGERUD: And then before you left 8 mode three, the required equipment would have to be 9 available to support hot shutdown and so on.
10 CO-CHAIR REMPE: Okay. Thank you.
11 MEMBER SUNSERI: So the way you -- the way 12 you have that mode five defined up there, it is 13 equivalent to at least in my view, what plants call at 14 all times right now.
15 Is that what was your intention?
16 MR. GROSS: Yeah. I believe it's slightly 17 different. At all times is whether there's fuel in 18 the core or not.
19 MEMBER SUNSERI: But that says less than 20 one bolt fully tension.
21 MR. GROSS: So he's --
22 MEMBER SUNSERI: You might look at that.
23 MR. GROSS: I'll take a -- thanks. I'll 24 take that back.
25 MR. SNUGGERUD: It's a good --
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137 1 MEMBER BLEY: Karl said something that I'm 2 not sure I read anywhere. And I'm not sure I heard 3 you correctly.
4 But I think you were saying, if you have 5 a module that's in a -- some kind of failed condition, 6 that before you do anything with that module, all the 7 others have to be meeting their tech specs.
8 Is that what you said?
9 MR. GROSS: All the others will meet their 10 tech specs contin -- they need to -- everything needs 11 to stay within its tech spec bound limits all the 12 time.
13 MEMBER BLEY: Yeah.
14 MR. GROSS: If there's fuel in the pot 15 basically. The -- if there's an upset module, it's by 16 definition outside a -- pretty much outside of tech 17 specs.
18 Not all -- that's not always true. But in 19 general, that's where it is. And not fully out 20 necessarily, but.
21 MR. SNUGGERUD: But the operators continue 22 to apply tech specs to the --
23 MR. GROSS: That's right. Exactly.
24 MR. SNUGGERUD: Other parts of the plant.
25 I think again, maybe part of your question goes to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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138 1 mode definitions are for the applicability of the 2 technical specifications.
3 They're not intended to describe all 4 potential plant conditions. They're described --
5 intended to describe the required equipment in given 6 plant conditions.
7 So, if I was in the condition where I had 8 un-tensioned some bolts and six bolts I couldn't get 9 un-tensioned, I know where I'm at on this table. The 10 plant might be in some unique configuration that would 11 require response of the operations organization and 12 the maintenance organization.
13 But, I have defined where I'm at in the 14 table. And that tells the operators what equipment is 15 required under those conditions.
16 And then you get the multi-module issue, 17 which you kind of hinted at as well. But now the 18 other modules can't get to the refueling station 19 because I have a module in the refueling station.
20 Those types of recovery plans are things 21 that we're talking about in the design and development 22 of our recover -- of our refueling equipment to make 23 sure that the challenges to the other plants aren't 24 experienced.
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139 1 preclude you from wanting to get into transition. So 2 you would stay in your safe shutdown condition.
3 And that would define the requirements for 4 that unit while it was sitting in its module space 5 waiting for the other issue to be resolved.
6 CO-CHAIR SKILLMAN: I was just thinking 7 back to what we got into when we started refueling at 8 TMI-2, and we thought we had it all sewed up nice and 9 clean, nice and tidy, and every time we turned around, 10 we were in a new situation, trying to figure out where 11 to go from here.
12 And a lot of the stuff, we were doing on 13 the fly. We were writing procedures, we were writing 14 stuff like this, trying to navigate our way through a 15 safe and defendable path.
16 And then, here we are at the front end of 17 this unique design, maybe a little more effort would 18 stave off what would be really a complex situation in 19 the future.
20 MR. SNUGGERUD: And it is our intent and 21 our hope that the effort that we've put into this, 22 I've been working on tech specs, at least part time, 23 for all ten years I've been at NuScale, would, to a 24 large extent, do that. I can't claim that we have 25 thought of everything, because that would be a false NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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140 1 claim.
2 But I did try and bring out the fact that 3 we do have a substantial staff of previously licensed 4 operators, and we have a very competent technical 5 staff in the Licensing Department. It won't catch 6 everything, but we have made a very big effort.
7 Like I said, we didn't just go through the 8 PWR tech specs, we looked at the BWR tech specs, we 9 looked at foreign types of things, to get input to 10 make sure that we were evaluating this new design in 11 as many ways as we could.
12 CO-CHAIR SKILLMAN: Thank you.
13 MR. SNUGGERUD: Let me get back into my 14 presentation here. So, this is going into the 15 sections of the technical specification that actually 16 hold the limiting conditions of operation. Section 17 3.1 and 3.2 are not terribly different from things 18 that you would normally see, so we have left those 19 out, and I've jumped to Instrumentation.
20 I mentioned in my opening that we have a 21 completely digital module protection system. And as 22 a result, that changed, to some extent, the way we've 23 organized the Instrumentation section of our technical 24 specifications.
25 The manual actuation function is there, at NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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141 1 3.3.4. That was one that we don't credit in our 2 Chapter 15 analysis, but as an operator, and our 3 operating staff, we agreed that the operators need to 4 have a way that's independent of the digital system to 5 affect the actuation of the engineered safety features 6 and the reactor trip system. I don't know if there's 7 anything else you want to add on this slide?
8 Passive Core Cooling is another section 9 that is unique to NuScale, in that our cooling systems 10 don't involve the pumps. So, we have the emergency 11 core cooling system, the decay heat removal system, 12 and we put the ultimate heat sink under passive 13 cooling because it obviously supports those two 14 systems. There are elements of the ultimate heat sink 15 that also support refueling operations, but they're 16 covered under 3.5.
17 Containment Systems, again, is a little 18 bit different. You're not going to find personnel 19 hatch controls for our technical specifications.
20 We're largely looking at the containment itself, its 21 function as an isolation vessel, and then, the valves 22 associated with isolating the containment.
23 Plant Systems, again, a smaller set. We 24 have main steam and main feed water, because the 25 isolation of the steam generators is important to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 function of DHR and for the mitigation of steam 2 generator tube ruptures.
3 And then, we have the in-containment 4 secondary piping leakages, which is a specification 5 that we added after working with the staff to ensure 6 that the assumptions of our leak-before-break 7 methodology are in place during normal operations.
8 3.8 typically is Electrical. For NuScale, 9 we don't have that section. What we do have is 10 Refueling Operations, which is normally 3.9.
11 MEMBER CORRADINI: So, is 3.8 then the 12 transitional and refueling?
13 MR. SNUGGERUD: So, it's --
14 MEMBER CORRADINI: I'm trying to understand 15 where movement fits into this and limiting conditions 16 --
17 MR. SNUGGERUD: The movement is mostly 18 controlled by the early technical specification on 19 Reactivity Requirements. So, when we're moving the 20 module, we ensure, prior to disconnection, that 21 adequate boron is in place, and we ensure that 22 adequate decay time, which is in 3.8, is in place.
23 And we ensure that the module temperature 24 and systems are aligned as designed, to ensure passive 25 cooling during transition. But in transition, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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143 1 specifically, there aren't any technical 2 specifications, beyond ensuring that those things that 3 were set up aren't changed.
4 MEMBER CORRADINI: And the conditions you 5 want?
6 MR. SNUGGERUD: Yes, correct.
7 MEMBER MARCH-LEUBA: So, is the nuclear 8 instrumentation live instrumentation? I mean, you 9 have a cable pulled in for the incore detectors?
10 MR. SNUGGERUD: So, the incore detectors 11 are only meaningful at power. So --
12 MEMBER MARCH-LEUBA: But that's better than 13 not having nothing?
14 MR. SNUGGERUD: The types of self-powered 15 incore detectors that we are using would not provide 16 any indication.
17 MEMBER MARCH-LEUBA: So, do you add 18 transition source level detectors?
19 MR. SNUGGERUD: So, we have source level 20 detectors at the operating position and at the 21 refueling position. Between those two, it is our 22 position that reactivity changes are precluded.
23 MEMBER MARCH-LEUBA: Okay. So, the nuclear 24 instrumentation specs don't say any -- say you don't 25 need anything --
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144 1 MR. SNUGGERUD: That's correct.
2 MEMBER MARCH-LEUBA: -- during transition?
3 CO-CHAIR SKILLMAN: Do the tech specs 4 require that the reactor be borated to the 1,800 or 5 2,000 PPM before it is disconnected and then moved?
6 I mean, clearly, you're depending on the rods to be 7 bottomed, but are you depending on full-load boron as 8 the preventative, so that you do not need neutron 9 monitoring during module movement?
10 MR. SNUGGERUD: So, there is no means of 11 changing out the fuel during the transition.
12 CO-CHAIR SKILLMAN: Oh, I understand that.
13 MR. SNUGGERUD: There is advantageous to 14 have the boron concentration at the bottom of the 15 vessel, at full boron concentration, prior to starting 16 refueling activities, if your intent is to load new 17 fuel, which requires the 1,800 PPM. The fuel that 18 you're taking out, doesn't.
19 The current tech specs would require that 20 you know that you have 1,800 PPM borated water in the 21 lower reactor vessel. It doesn't currently specify 22 the details of how you do that.
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145 1 would be to find a way to ensure that you had flushed 2 that vessel prior to moving anything and taking 3 samples appropriately.
4 We haven't specified that level of detail 5 in the tech specs, other than to ensure that the 6 requirements for the new core would be met prior to 7 the new core being inserted into the lower reactor 8 vessel.
9 CO-CHAIR SKILLMAN: So, you're basically 10 allowing credit for burnup, as reactivity depletion --
11 MR. SNUGGERUD: Yes.
12 CO-CHAIR SKILLMAN: -- and not necessarily 13 requiring full-load boron before you move the module?
14 MR. SNUGGERUD: That's correct. And the 15 existing fleet could do the same. Again, they know 16 they're going to put fuel back in their modules and 17 they don't want their cold legs sitting there with 18 boron concentrations that don't meet the refueling 19 boron concentrations for the new core.
20 So, they borate to refueling 21 concentrations, not because the core they're taking 22 out needs them to, but because the core they're 23 putting in requires it. Same thing applies to 24 NuScale. And the NuScale facility will have the same 25 expectations that the existing facilities have.
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146 1 CO-CHAIR SKILLMAN: Yes, except that you 2 borate it up before you take the head off, and so, 3 you're fully borated before you touch your first fuel 4 assembly.
5 MR. SNUGGERUD: That's correct. At our 6 plant, when you take the head off, you're going to be 7 exposing it to a seven million gallon pool that meets 8 all boron concentration requirements.
9 MEMBER MARCH-LEUBA: And in the operating 10 fleet, they have pumps that they can kind of mix the 11 boron.
12 MR. SNUGGERUD: Yes, but the pumps don't 13 hit all portions of the shutdown plant, depending on 14 the plant's design. Which, again, is why typically, 15 and at the plant that I operated at, we borated to the 16 new core's boron concentration before starting 17 refueling processes.
18 MEMBER MARCH-LEUBA: But I've been reading 19 in-between lines that you expect the boron 20 concentration in the vessel to be non-uniform. And --
21 MR. SNUGGERUD: That's not what I'm saying.
22 MEMBER MARCH-LEUBA: -- how do you know 23 what you --
24 MR. SNUGGERUD: What I'm saying --
25 MEMBER MARCH-LEUBA: -- do you say the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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147 1 bottom of the vessels or --
2 MR. SNUGGERUD: The easy way to do it would 3 be to do the same thing at NuScale, right? It would 4 be to borate the module to 1,800 PPM on the way over.
5 As you had pointed out, during shutdown, 6 there's going to be ample space made during the 7 cooling process, that you could add as much boron as 8 you chose to. I don't want to tell you that that's 9 what they're going to do, because the tech specs don't 10 force them to do that, and technically, it isn't 11 required.
12 What is required is, before they touch a 13 bundle and start putting new fuel into the reactor, 14 that they ensure that that lower pole of the reactor 15 that's sitting in the pool is at the boron 16 concentration that they intended.
17 MEMBER MARCH-LEUBA: How they do that if 18 instrument is on the discharge line of the CVCS?
19 MR. SNUGGERUD: So, as I suggested, the 20 easy way to do that would be to borate the entire 21 module prior to ever doing it.
22 MEMBER MARCH-LEUBA: That was my --
23 MR. SNUGGERUD: But there are other ways of 24 doing it --
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148 1 expectation --
2 MR. SNUGGERUD: -- and the tech spec --
3 MEMBER MARCH-LEUBA: -- before you started 4 talking.
5 MR. SNUGGERUD: Yes. The tech specs don't 6 tell them how they have to do that, they tell them 7 what they have to do. I would agree with you, that 8 would be the easy way.
9 MEMBER MARCH-LEUBA: Yes, if you don't have 10 a sensor up there, that's the only way.
11 MR. SNUGGERUD: I can envision means by 12 which you could ensure turnover of that lower module.
13 It's not a very big space, you have hot fuel in it.
14 It's going to be pulling water down the down comer and 15 pushing it up through the core, just because of the 16 flow design of that module.
17 I don't want to tell you that somebody in 18 the future won't do the calculations to make that 19 argument and do the demonstrations to prove that 20 that's appropriate, but I haven't done them.
21 MEMBER MARCH-LEUBA: I have a running 22 concern with the staff about how well the boron mixes 23 when you don't have any flow, and indeed, experiments 24 for BWR show the boron does not mix and it goes all to 25 the bottom of the vessel. And this will have to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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149 1 addressed somewhere --
2 MR. SNUGGERUD: Yes.
3 MEMBER MARCH-LEUBA: -- I expect Chapter 4 15.
5 MR. SNUGGERUD: And I've heard your 6 concerns and I know NuScale is aware of them.
7 So, this is just a brief, the staff will 8 cover this in more detail, but we have received 9 approximately 60 Chapter 16 RAIs. A significant 10 number of them have been responded to. We do have 11 some outstanding and we do have some that were 12 recently provided.
13 I will also mention that the number of 14 RAIs isn't exactly the same as the number of 15 questions, because one example up there, a single RAI 16 had 80 subparts. So, we have more than 60 questions 17 about the tech specs. The staff has -- well, I guess 18 I have another slide that kind of covers some of the 19 high level things.
20 There were a lot of discussions about how 21 we chose to implement standard tech spec content into 22 our design. And so, there were some formatting and 23 some deliberations about the most appropriate way to 24 do that, to ensure that the understanding is 25 consistent with the existing fleet.
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150 1 We got a number of questions related to 2 instrumentation, which makes sense, again, as NuScale 3 has a unique and newly designed module protection 4 system, which is the subject of a Topical Report. And 5 that spawned additional questions related to response 6 time testing, because the barriers or the designations 7 between where you do your testing are a little bit 8 different for our design.
9 We discussed leak rate monitoring. Our 10 containment is different, it's dry, it's in a vacuum.
11 So, what are we going to use to give us evidence and 12 make the operators aware of any leakage before it 13 becomes a problem? We had discussions about that 14 system.
15 And then, we added the RCS-specific 16 activity limit that addresses dose rates for 17 operators, which is something that the staff asked 18 for.
19 And then, lastly, we added a tech spec, 20 most recently, related to any containment secondary 21 piping leakage that supports our leak-before-break 22 methodology.
23 So, in conclusion for my side, NuScale has 24 spent a lot of time to develop technical 25 specifications and bases that reflect our design and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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151 1 the analysis and the operations as we envision them.
2 We've tried to ensure the safety in 3 licensing basis will be implemented as designed, 4 aligned with the standards that the industry uses, 5 take into account industry experience, and prepare a 6 document that will be easy for the COLA applicant to 7 finalize, consistent with the intent of both the staff 8 and NuScale.
9 CO-CHAIR REMPE: If there aren't any more 10 questions, we're again running behind, and so, I'm 11 going to ask the staff to come up.
12 MR. SNUGGERUD: Okay.
13 CO-CHAIR REMPE: Thank you.
14 MR. TESFAYE: Good morning. I'm also the 15 Chapter 16 PM. Good morning, again. Craig Harbuck is 16 my to my right and he's the technical reviewer, as you 17 may all know. Greg's not here, he's our lead PM.
18 Staff has completed a Safety Evaluation 19 Report and submitted it about a month ago. This 20 report contains several open items when it was issued, 21 but a good portion of the open items have since been 22 closed.
23 The remaining open items will be addressed 24 in Phase 4 of the review. Craig will discuss in 25 detail the open items that have been closed and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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152 1 remaining issues that will be addressed in Phase 4 and 2 his overall technical findings.
3 Now, at this time, Craig has asked me, if 4 it is okay with Committee, he would like to go through 5 the presentation and then, the Q&A, if that's okay 6 with you, in the interest of time.
7 CO-CHAIR REMPE: Thank you.
8 MR. TESFAYE: Okay.
9 MR. HARBUCK: I'll be discussing open items 10 that are still open.
11 CO-CHAIR REMPE: Okay. So, please scoot 12 very close to the mic.
13 MR. HARBUCK: Okay. Can you hear me now?
14 CO-CHAIR REMPE: Yes.
15 MR. HARBUCK: It's good? Okay. My name is 16 Craig Harbuck. I'm in the Technical Specifications 17 Branch and have been for a long time. Next -- no, 18 previous slide. Previous slide.
19 This is the people who I was able to 20 identify that were participating in supporting this 21 review. If I've left anyone out, I apologize. Next 22 slide.
23 The outline of this presentation, it 24 consists of an overview of the tech specs and then, a 25 number of technical topics that are mostly related to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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153 1 open items.
2 A brief overview of the tech spec chapter, 3 sections, and subsections. This outline shows that 4 the tech spec chapters and sections are mostly 5 organized consistent with a PWR STS NUREG.
6 Notice that the NuScale tech specs are 7 missing the LCO that allows operation for a limited 8 period of time with an inoperable piping support 9 snubber. That's normally LCO 308, but since they're 10 not having that one, that's what it will be.
11 Also, NuScale tech specs do not have a 12 Chapter 3 section for electrical power, for reasons we 13 all know. Next slide.
14 This slide lists the Chapter 3 LCO 15 subsections that were the result of applying the LCO 16 Selection Criteria 50.36. NuScale's evaluation of its 17 design and safety analyses against the LCO Selection 18 Criteria is described in their report entitled 19 Technical Specifications Regulatory Conformance and 20 Development. Currently, we're up to Revision 1 on 21 that report.
22 There's two LCOs that satisfy both 23 Criterion 2 and Criterion 3. That would be boron 24 dilution control and the ultimate heat sink.
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154 1 differs from the equivalent STS LCO by only including 2 verification of reactor coolant system flow resistance 3 within limits at the beginning of each operating 4 cycle, right after refueling.
5 And this is due to the use of natural 6 circulation, where coolant flow is a function of core 7 power, and therefore, we don't include a 12-hour 8 surveillance that you typically have for verifying RCS 9 flow.
10 Also, as mentioned previously in NuScale's 11 presentation, they're using a leak-before-break 12 evaluation, to ensure an in-containment secondary 13 system line break is essentially precluded from 14 occurring. To ensure this remains valid, LCO 3.7.3 is 15 established, to specify a limit on the rate of 16 secondary leakage into containment. Next slide.
17 This slide lists the sections and 18 subsections of Chapter 4, Design Features, and Chapter 19 5, Administrative Controls.
20 In the Procedures Section, that's 5.4.1, 21 NuScale agreed to add a requirement for procedures to 22 control the availability and reliability of systems 23 covered by the Owner Control Requirements Manual, 24 because these systems are not covered by the Reg Guide 25 1.33 required procedures or the Quality Assurance Plan NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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155 1 procedures.
2 And examples of systems that we would 3 expect to see covered by the OCDM are control room 4 habitability, ventilation, the CVCS system, 5 demineralized water system, electrical power, the 6 containment evacuation system, the core flood and 7 drain system, and a number of others.
8 We'd also like to point out that the 9 NuScale tech specs include a surveillance frequency 10 control program, which is Risk-informed Tech Spec 11 Initiative 5B, which is implemented using Traveler 12 TSTF-425. Next slide.
13 CO-CHAIR SKILLMAN: Craig, if I could 14 please, on that same slide.
15 MR. HARBUCK: Okay.
16 CO-CHAIR SKILLMAN: Under the procedures 17 that you just enumerated, which is 5.4.1, Frank, F, 18 would you also expect the procedures for the operation 19 of the heavy load handling system, the crane, to be 20 binned in that same category?
21 MR. HARBUCK: I'm not sure, that's not 22 normally included in the procedures that are required.
23 Well, let me put it this way, if that equipment was 24 deemed to be safety-related, perhaps it would be 25 covered already by --
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156 1 CO-CHAIR SKILLMAN: Do you think it's 2 safety-related?
3 MR. HARBUCK: -- Reg Guide 1.33.
4 CO-CHAIR SKILLMAN: Do you think it's 5 safety-related? The crane?
6 MR. HARBUCK: Oh, I do think so, yes.
7 CO-CHAIR SKILLMAN: So, do you think the 8 procedures --
9 MR. HARBUCK: But --
10 CO-CHAIR SKILLMAN: -- may be in here?
11 MR. HARBUCK: Well, everything they do at 12 the plant should have procedures. What I'm concerned 13 about in the tech spec arena is to ensure the things 14 that we've normally included in tech specs, but that 15 are not being called out as meeting any of the 16 criteria, that there be some means of -- I mean, a 17 tech spec requirement for there to be procedures.
18 Otherwise, it's hard to know what level of control 19 we're going to have, since none of these things fall 20 within the other non-tech spec controls.
21 CO-CHAIR SKILLMAN: Okay, I'll leave it 22 there. Thanks.
23 MR. HARBUCK: Okay.
24 MEMBER BLEY: I need a little help, I'm not 25 well-versed here. Your Slide 5 is the Chapter 16 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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157 1 sections, yes?
2 MR. HARBUCK: Yes, that's correct.
3 MEMBER BLEY: And the last two slides you 4 showed, your 6 and 7, are from Part 4, which as I 5 think I understand it is going with the Generic Tech 6 Specs in the NUREG and seeing which ones apply, is 7 that right?
8 MR. HARBUCK: Well, what Slide 6 does, it 9 lists the LCOs that were identified as meeting the 10 criteria and one that doesn't, in 50.36(c)(2)(ii).
11 MEMBER BLEY: And if all these meet the 12 criteria, except for one, shouldn't they all be in 5?
13 I'm just a little confused about the relationship 14 between --
15 MR. HARBUCK: Oh, the --
16 MEMBER BLEY: -- Part 4 and Chapter 16.
17 MR. HARBUCK: We refer to tech specs as 18 having chapters, sections, and subsections.
19 MEMBER BLEY: Yes.
20 MR. HARBUCK: On Slide 5, I've gone down to 21 the section level.
22 MEMBER BLEY: Okay. So, you've gone a 23 layer deeper?
24 MR. HARBUCK: That's right.
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158 1 electric power?
2 MR. HARBUCK: There are no LCOs for 3 electrical power, except for the Class 1E isolation 4 devices to protect the MPS instrumentation.
5 MEMBER BLEY: And the reason there's not?
6 I mean, there's something in here on CVCS --
7 MR. HARBUCK: It is --
8 MEMBER BLEY: -- is non-safety.
9 MR. HARBUCK: Well, that had to do with 10 NuScale agreeing to add another provision for 11 procedures --
12 MEMBER BLEY: Okay.
13 MR. HARBUCK: -- so we would have something 14 to point to.
15 MEMBER BLEY: I'm linking back to things 16 that aren't --
17 MR. HARBUCK: Yes.
18 MEMBER BLEY: -- in your area here, but --
19 MR. HARBUCK: Okay.
20 MEMBER BLEY: -- when we reviewed the 21 chapter on electric power, they, for several reasons, 22 decided not to have 1E electric power system and to 23 have a highly reliable electric power system.
24 MR. HARBUCK: Yes, DC power.
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159 1 there ought to be something like a tech spec 2 requirement on it.
3 MR. HARBUCK: The conclusions, I believe, 4 on Chapter 8, were that there were no safety-related 5 electrical systems. No electrical power is required, 6 since everything that actuates in accidents does so by 7 removing electrical power. So, it's my understanding 8 that none of the criteria have been satisfied by the 9 electrical power.
10 MEMBER BLEY: So, with this highly reliable 11 electric power system, they could have it shut down 12 and it would not affect the tech specs? It could be 13 turned off?
14 MR. HARBUCK: Well, they might have a hard 15 time operating if they didn't have that.
16 MEMBER BLEY: Yes, they might.
17 MR. HARBUCK: I had a thought --
18 MEMBER BLEY: Go ahead. Chapter 8 isn't 19 finished, by the way, there were a lot of outstanding 20 things.
21 MR. HARBUCK: Okay. Anyway, this reflects 22 our understanding of the current status of Chapter 8 23 and its role in the safety analysis.
24 MEMBER MARCH-LEUBA: Chapter 8? I've been 25 saying Page 8.
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160 1 MEMBER BLEY: Sorry.
2 MR. HARBUCK: Okay. That's the end of the 3 overview. Now, we're going to start on the technical 4 topics. And before I begin, just let me follow up on 5 something from the previous presentation.
6 The mode definition for Mode 4, that 7 transition mode, requires shutdown margin to be 8 whatever the reactivity condition, k-effective less 9 than 0.95, which is going to dictate what your boron 10 concentration is going to have to be set at before you 11 start moving the module.
12 And then, when you disassemble the module, 13 the requirements on the pool boron concentration in 14 3.5.3 would dictate the reactivity condition to 15 maintain that requirement.
16 MEMBER MARCH-LEUBA: Do tech specs say you 17 have to have a value for k-effective? Or do they say 18 --
19 MR. HARBUCK: No, it says k-effective, I 20 think that's --
21 MEMBER MARCH-LEUBA: Which you cannot 22 measure during operation.
23 MR. HARBUCK: It's correlated with -- it's 24 calculated. You know what the boron concentration is, 25 you estimate --
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161 1 MEMBER MARCH-LEUBA: It's --
2 MR. HARBUCK: -- how long since you shut 3 down and --
4 MEMBER MARCH-LEUBA: I wanted to save it 5 for the end, but she will make me shut up, because 6 we'll run out of time.
7 MR. HARBUCK: Okay.
8 MEMBER MARCH-LEUBA: First principles, if 9 you insert all rods, at any point in the cycle, but 10 mostly towards the end of the cycle, when boron 11 concentration is very low and the monitored 12 temperature coefficient is very negative, you will 13 have a temperature, a relationship between boron 14 concentration and core temperature, which you regain 15 criticality. So, for every boron concentration, you 16 have a temperature you become critical again.
17 Why is this not on tech specs? There 18 should be a three-dimension table that tells me 19 temperature below which I cannot go, as function of 20 boron concentration and burnup. And the statement, it 21 says, you cannot reach is.
22 Tell me what it is and put it in tech 23 specs, because the operator doesn't know what k-24 effective is, if he can measure the temperature.
25 MR. HARBUCK: Well, at the moment, these NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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162 1 values are not in the tech specs, they're in the Core 2 Operating Limits Report, which we haven't seen.
3 MEMBER MARCH-LEUBA: Which you have not 4 seen?
5 MEMBER CORRADINI: Because that's typical 6 of most PWRs?
7 MR. HARBUCK: That has been the case since 8 the 1990s, since we adopted this Core Operating Limits 9 Report.
10 MEMBER MARCH-LEUBA: When will be able to 11 see the COLR?
12 MR. HARBUCK: I would have to let NuScale 13 answer that.
14 MEMBER MARCH-LEUBA: So, it's --
15 MR. HARBUCK: It's probably not something 16 that's part of the DCA.
17 MEMBER MARCH-LEUBA: But do we know -- can 18 we specify that something like this must exist in the 19 COLR?
20 MR. HARBUCK: Yes, because the tech spec on 21 shutdown margin says, you have to meet the values that 22 are in the COLR. And then, there's a definition of 23 shutdown margin, which provides the other assumptions 24 you have to make in calculating what your shutdown 25 margin is.
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163 1 MEMBER MARCH-LEUBA: Okay. We'll talk 2 about this at some point.
3 MR. HARBUCK: Okay. All right. So, enough 4 of that.
5 Last thing I wanted to say was, the 6 definition of mode also includes a phrase that says 7 there's got to be fuel in the vessel. So, when you 8 remove all your fuel, you're not in a mode. Okay.
9 I'll now discuss several topics of 10 interest for which the staff and NuScale have not yet 11 reached resolution. And first up are several defined 12 terms.
13 The sentence that NuScale proposes to add 14 to the pressure boundary leakage definition, it's 15 indicated by the blue underlined text there, matches 16 an STS change previously proposed by industry that the 17 staff did not accept.
18 We do not believe that this relaxation, to 19 permit leakage through an isolation device of a fault 20 in the reactor coolant pressure boundary, is 21 warranted. We are also not convinced that there are 22 locations in the NuScale reactor coolant pressure 23 boundary where this allowance could even be used.
24 Next slide.
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164 1 understand that a little better.
2 MR. HARBUCK: Okay.
3 MEMBER CORRADINI: Okay, thank you.
4 CO-CHAIR REMPE: We can go on, Mike, or do 5 you have a question?
6 MEMBER CORRADINI: No, I just was trying to 7 catch up.
8 MR. HARBUCK: The next one is operability.
9 And we believe that the indicated changes of the STS 10 operability definition are appropriate for NuScale.
11 This is a markup of the STS definition.
12 The most noteworthy difference with the 13 STS definition is the removal of the phrase normal or 14 emergency to describe electrical power. This makes 15 sense, because, as we've previously mentioned, NuScale 16 has no safety-related emergency electrical power 17 sources and, thus, no LCO section for electrical power 18 systems.
19 Also, notice this definition includes two 20 lists of the various ways tech specs refer to a 21 redundant part of a supported system. These lists 22 should match, and I've added some markup to do that.
23 The second list should include the term 24 Separation Group, as does the first list, it's a 25 NuScale term. And then, both lists should include NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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165 1 Channel and Division, because the MPS instrumentation 2 has four channels and the MPS RTS and ESFAS actuation 3 logic functions have two divisions. Next slide.
4 NuScale proposed to admit the STS 5 definitions for reactor trip system response time and 6 engineer safety feature response time, because the 7 digital portion of the MPS will contribute a 8 conservatively allocated response time to the total 9 ESF and RTS response times, and NuScale thereby 10 concludes that these STS definitions are not needed to 11 effectively state the Surveillance Requirements for 12 verifying the I&C response times are within limits, as 13 explained in the bases of the following Surveillance 14 Requirements.
15 SR 3.3.1.3 is MPS implementation response 16 time measurement. This surveillance verifies for each 17 instrumentation function the response time of each 18 channel from when the process variable exceeds its set 19 point until the output from the channel analog logic 20 reaches the input of the MPS digital logic. So, the 21 response time for this surveillance is just handling 22 the input to the digital system.
23 And then, SR 3.3.2.3 verifies for each 24 reactor trip system actuation logic division, that the 25 division response time from the output of equipment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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166 1 interface module until the reactor trip breakers are 2 opened. So, it goes to the other end of the 3 instrument loop and handles the output from the 4 digital system.
5 And again, the digital system, they 6 provide a conservative allocated time in calculating 7 the total response time. Next slide.
8 MEMBER BROWN: Go backwards a minute.
9 MR. HARBUCK: Slide 11?
10 MEMBER BROWN: Is the stuff in blue 11 replacing the lines above it or --
12 MR. HARBUCK: No --
13 MEMBER BROWN: -- is that just an 14 explanation --
15 MR. HARBUCK: Well, what --
16 MEMBER BROWN: -- that we're going to get 17 rid of the RTS and ESF terms?
18 MR. HARBUCK: Well, what this is meant to 19 indicate, what I'll get to on the next slide, these 20 are quotes from the bases and they are meant to 21 indicate some material that was taken from the defined 22 terms, the definitions, that have been put into the 23 bases now. And I had a point to make about that.
24 CO-CHAIR REMPE: Let's let him get to the 25 next slide, okay? I think it explains it.
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167 1 MR. HARBUCK: Okay. Now, SR 3.3.3.2 2 verifies, and this is kind of interesting, so listen 3 carefully, it verifies the pressurizer heater trip 4 ESFAS function logic and actuation division response 5 time, from the output of the equipment interface 6 module to the loss of voltage at the output of the 7 pressurizer heater breaker.
8 MEMBER BROWN: Can you -- that's on the 9 previous slide?
10 MR. HARBUCK: That's on Slide --
11 MEMBER BROWN: You said SR --
12 MR. HARBUCK: That's on Slide 11, 3.3.3 --
13 MEMBER BROWN: 2 was on Slide 10.
14 MR. HARBUCK: Did I mix -- well, 3.3 --
15 CO-CHAIR REMPE: Well, if I can --
16 MR. HARBUCK: -- I believe it's 3.3.3.2, it 17 might be 3.3.3.3, I could have -- anyways, the 18 Surveillance Requirement number is not so important, 19 but the response time surveillance for the ESFAS 20 functions in the Instrumentation section only 21 explicitly addresses the pressurizer heater trip.
22 Now, the surveillances for the other ESFAS 23 functions, logic and actuation functions, that opens 24 or closes valves are included in each associated 25 system's LCO subsection.
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168 1 Now, the reason for the blue highlighted 2 text is because, unlike the Section 3.3 bases for 3 response time surveillances, the bases of the 4 surveillances to verify valve actuation time in LCO 5 subsections of affected systems do not include 6 descriptive language, which is consistent with the 7 omitted STS response time definitions. And in the 8 absence of those definitions, we are suggesting that 9 the applicant add such language to the bases for those 10 surveillances. Next slide.
11 CO-CHAIR REMPE: You okay, Charlie?
12 MEMBER BROWN: I'm still having trouble 13 with that. Somebody else understands it better than 14 I do. It seems to me, you've added -- the blue is not 15 part of the overall STS? I'm not real familiar --
16 MR. HARBUCK: The quotations are taken from 17 the bases, for that surveillance.
18 MEMBER BROWN: I'm trying to figure out 19 what the intent of the blue part is. I mean --
20 MR. HARBUCK: It would --
21 MEMBER BROWN: -- maximum response time --
22 MR. HARBUCK: That phrase, that sentence is 23 a phrase that's verbatim out of the definitions in the 24 Standard Tech Specs.
25 MEMBER BROWN: Okay. Go back to Slide 10, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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169 1 I might --
2 MR. HARBUCK: Okay.
3 MEMBER BROWN: -- be able to understand 4 that one better, 3.3.2. The Standard says, the 5 maximum digital response time is described in the 6 FSAR. Okay.
7 The SR encompasses the response time from 8 the output of the equipment interface until the trip 9 breakers open. Now, the next statement says, to me, 10 that you can do it piecemeal or overall. Is that --
11 MR. HARBUCK: That's true. And the total 12 response time there is referring not just --
13 MEMBER BROWN: Is that --
14 MR. HARBUCK: -- to what that surveillance 15 is --
16 MEMBER BROWN: -- in the Standard right 17 now?
18 MR. HARBUCK: Yes.
19 MEMBER BROWN: So, that's not an addition?
20 MR. HARBUCK: That's not an addition or 21 anything, I'm just point out that --
22 MEMBER BROWN: Okay. I thought --
23 MR. HARBUCK: -- in compensating --
24 MEMBER BROWN: I'm sorry, I thought they 25 were modifying the Standard.
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170 1 MR. HARBUCK: No.
2 MEMBER BROWN: I thought they were 3 modifying the Standard. I understand --
4 MR. HARBUCK: They were beefing up the 5 bases --
6 MEMBER BROWN: -- whether I agree with that 7 or not, that's another thing. But that's what's been 8 used in the past?
9 MR. HARBUCK: Right, yes.
10 MEMBER BROWN: Okay, thank you.
11 MR. HARBUCK: Okay.
12 MEMBER BROWN: Jose, you were on the same 13 page I was? Okay.
14 CO-CHAIR REMPE: We're a bit behind, so 15 let's go ahead, okay? We're good now.
16 MEMBER BROWN: Yes, I got it. I got it, 17 we're okay.
18 MR. HARBUCK: Slide 12. Okay. All right.
19 So, very briefly here. Some other defined 20 Surveillance Requirements.
21 The MPS, module protection system, 22 continuously conducts automatic self-testing to verify 23 channel and division operability between the channel 24 calibrations and actuation logic test.
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171 1 effective than performing the STS channel operational 2 test every 92 days and the STS actuation logic test 3 every 24 months, for verifying proper operation of the 4 MPS instrumentation and the RTS and ESFAS logic and 5 actuation divisions.
6 This is a reminder that in the Standard 7 Tech Specs, the circuits are typically analog and so, 8 these tests involve injecting a signal and seeing what 9 happens to the logic and doing the various 10 combinations.
11 MEMBER BROWN: Okay. If I understand --
12 MR. HARBUCK: And they don't --
13 MEMBER BROWN: -- this one --
14 MR. HARBUCK: And they don't do that one.
15 Okay.
16 MEMBER BROWN: And if I understand --
17 MR. HARBUCK: All right.
18 MEMBER BROWN: -- in the old days, with the 19 analog stuff, you had a periodic test that you had to 20 do, where you input stuff.
21 MR. HARBUCK: That's right.
22 MEMBER BROWN: Since we've got self-23 testing, that's performed and the only check of the 24 self-testing is on a two-year basis. That's the way 25 --
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172 1 MR. HARBUCK: The --
2 MEMBER BROWN: -- I'm reading it. They go 3 -- that's how I'm reading this is that they have to do 4 something to verify that the self-testing is actually 5 self-testing.
6 MR. HARBUCK: That particular point in the 7 Branch Technical Position that addresses the self-8 testing of digital systems would have to be addressed 9 by Instrumentation reviewers. My understanding is 10 that testing of the tester has been addressed in that 11 discussion of that chapter.
12 MEMBER BROWN: Yes, I don't remember.
13 MR. HARBUCK: But I'm not in a position to 14 --
15 MEMBER BROWN: It's been a while.
16 MR. HARBUCK: -- question that at this 17 point. Okay.
18 Now, the STS actuation logic test 19 definition has been revised for NuScale, since the 20 NuScale actuation logic test has a reduced scope. It 21 goes from the equipment --
22 MEMBER BROWN: Well, I'm reading the --
23 MR. HARBUCK: -- interface --
24 MEMBER BROWN: I'm reading the first 25 paragraph, which says, you depend on self-testing and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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173 1 division operability between channel calibrations and 2 actuation logic tests, which are on 24.
3 So, I think that reads, the way I'm 4 reading it, am I missing -- Jose, do you see that the 5 same way? They're relying two years on self-tests.
6 Every two years, there's going to be an operability of 7 the channel calibrations.
8 MEMBER MARCH-LEUBA: I understand what 9 you're saying, I don't understand what they're saying.
10 I think we're on the same page.
11 MEMBER BROWN: Yes, Page 12. Okay.
12 MR. HARBUCK: Okay. So, as I was saying, 13 the actuation logic test in NuScale verifies the 14 analog actuation priority logic in the equipment 15 interface module and the various manual switches which 16 have no automatic self-test features.
17 Now, the STS channel check definition is 18 also revised, as shown by the markup on this slide, to 19 account for the operation of the self-testing feature 20 of the digital platform. So, they've kept the 21 surveillance and they've kept essentially the same 22 frequency, but how it's performed is a bit different 23 than in typical plant. Okay.
24 Electrical Power Surveillance Requirements 25 is a bit of misnomer, but I wanted to say it to make NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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174 1 the point. NuScale protects the MPS from faults in 2 the non-safety-related electrical power system by use 3 of Class 1E isolation devices that automatically 4 separate the MPS from its electrical power upon 5 detecting an under voltage or overcurrent condition.
6 LCOs for the MPS reactor trip system and 7 ESFAS include, for the associated Class 1E isolation 8 devices, a Channel Calibration Surveillance 9 Requirement, which verifies operability of each 10 device.
11 They don't include such a surveillance for 12 the manual actuation and we've asked them to consider 13 whether that's appropriate.
14 Given the importance of maintaining water 15 inventory in the decay heat removal system following 16 its actuation, the main steam isolation and feed water 17 isolation valves must not leak more than the leakage 18 rate that will support decay heat removal system 19 operation for the time period needed following an 20 event that relies on the decay heat removal system for 21 core cooling.
22 NuScale has not provided values as part of 23 its application for these leakage limits, which are 24 the acceptance criteria for the leak rate verification 25 surveillances for the MSIVs and the feed water NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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175 1 isolation valves.
2 We are still evaluating NuScale's position 3 that these acceptance criteria will be established 4 when the in-service test program is put in place by 5 the COL applicant. NuScale has concluded, therefore, 6 that these values do not need to be provided as part 7 of the DCA.
8 CO-CHAIR REMPE: So, we're really running 9 behind, can you highlight key points on the remaining 10 slides, please?
11 MR. HARBUCK: Okay. In the next two 12 slides, I'll just briefly put it -- make it very 13 brief.
14 They've proposed, again, a provision 15 that's in an unapproved industry Change Traveler to 16 the STS that would allow Surveillance Requirements to 17 be suspended and not required to be met if the 18 isolation device or valve is put in its actuated 19 position. And we're not convinced that this is needed 20 or warranted. And so, we have open item related to 21 that. Okay.
22 Now, Slide 17. The next six slides 23 involve a listing of our open items, several of which 24 have been discussed, as indicated by the blue color on 25 their number.
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176 1 16-30D, at the bottom of the page, is 2 related to their adoption of a surveillance frequency 3 control program. We asked them to provide, in the 4 DCA, the initial frequencies and the bases for those 5 frequencies, which is necessary in order for a COL 6 applicant to actually implement that program. And 7 they put this in a table in FSAR Section 16.1. And 8 that table needs to be updated and so, this item is 9 simply tracking that.
10 MEMBER BLEY: On this long list of open 11 items, are there any on which you're really divided or 12 is it on most of them or are you getting pretty close 13 to agreement on how these are going to settle out?
14 MR. HARBUCK: I'll point that out, if I 15 mention one --
16 MEMBER BLEY: Okay.
17 MR. HARBUCK: -- if there's a -- I'm not 18 going to really talk about the ones that we've -- I 19 mean, I guess, one would be on Slide 18, the one about 20 providing the Surveillance Requirement acceptance 21 criteria for the valve leakage in the steam and feed 22 systems. That would be one where we'd perhaps are in 23 disagreement.
24 MEMBER CORRADINI: And again, a little bit 25 louder, that's the one that you're still in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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177 1 discussions with?
2 MR. HARBUCK: Right. On the -- Question 3 16-59 has to do with apparent disconnect on the 4 applicability of this instrument function, which 5 initiates the decay heat removal system, but there's 6 going to be a design change coming that we've been 7 made aware of that's going to alter their response to 8 that question. They haven't, as a result, they 9 haven't responded to that one yet and that's why it's 10 open.
11 And on the next slide, we have some issues 12 about the applicability of the LCO 3.1.9, we made a 13 suggestion.
14 And the Mode 3 applicability for LCO 15 3.3.1, what's interesting there is that there's an 16 interlock, which blocks it or bypasses it, and yet, 17 the applicability goes beyond, goes to a lower 18 temperature than at the point at which it's blocked, 19 so that's a question here.
20 And finally, there was some confusion in 21 an Action Requirement, which said, decrease power 22 below this interlock, the interlock refers to 15 23 percent rated power and the question was, which 24 interlock is the appropriate one to refer to?
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178 1 recently provided six new questions to NuScale in this 2 RAI 9642.
3 And I'll just mention briefly that we have 4 a concern about the design of their system to prevent 5 putting water into containment when the reactor 6 coolant system is at too high a temperature.
7 There's going to be in the PTLR a 8 temperature limit and the way they preclude it now is 9 that they have an interlock based on that temperature 10 on a control valve, a non-safety control valve in the 11 core flood and drain system. And we're asking, well, 12 why didn't they put an interlock on the containment 13 isolation valves, which are a higher level of safety 14 grade?
15 And then, lastly, we're asking them to add 16 some additional methodology references to the Core 17 Operating Limits Report Specification 5.6.3.
18 The last one refers back to the leak-19 before-break, where we're still reviewing the limit 20 they've proposed in LCO 3.7.3 of 1.5 gallons per hour.
21 And we're still reviewing their supplemental response.
22 And we're also, we think we're waiting for some 23 additional input, that may be incorrect, but that's 24 our understanding right now.
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179 1 number, but the COL item list or the sub-item list, as 2 it's been called, are all the places where information 3 is bracketed in the tech specs and bases, where we 4 want that list to be definitive as part of the design 5 cert, so that the COL applicant knows exactly what 6 they need to do to complete the tech specs.
7 CO-CHAIR REMPE: So, we have like one 8 minute to finish this.
9 MR. HARBUCK: Okay.
10 CO-CHAIR REMPE: We've got --
11 MR. HARBUCK: I'm --
12 CO-CHAIR REMPE: -- another meeting, so 13 maybe summarize.
14 MR. HARBUCK: Yes, I'm -- okay. The next 15 slide is just your -- okay. Slide 23 is just for your 16 information, to say we've got some more work to do, 17 because they made some design changes.
18 And then, there's some general issues.
19 One, I just mentioned, completing the COL item list.
20 Also, we have some issues related to their adopting 21 Travelers that aren't approved. And then, there's a 22 bunch of administrative items that need to be cleaned 23 up.
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180 1 that need to be resolved. I wouldn't say any of them 2 are show-stoppers or very significant, in and of 3 themselves.
4 And the Technical Branches listed will be 5 providing some support for us in completing those.
6 And when those are done, we'll be done with the tech 7 specs. That concludes the presentation.
8 CO-CHAIR REMPE: Okay, thank you. I 9 apologize for other people's sessions going longer.
10 MR. HARBUCK: I understand.
11 CO-CHAIR REMPE: Before we decide about 12 having public comment, I want to have the members say 13 whether they think we need to have a closed meeting to 14 address, for example, your buoyancy question, Dick.
15 Do you want to explore that further today?
16 So, any members see a need to have a closed meeting?
17 Mike, did you have a comment before I start asking for 18 public comments? Okay.
19 Do we have anyone on the phone line? Is 20 it open? Yes, it's open? Okay. Do we need to verify 21 that someone's out there or do we just --
22 MR. SNODDERLY: Is there anyone on the 23 public phone line?
24 MR. LEWIS: Marvin Lewis, public --
25 MS. FIELDS: Sarah Fields.
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181 1 CO-CHAIR REMPE: Okay. Do either of you 2 have a comment you'd like to make at this time?
3 MR. LEWIS: Yes, I do.
4 CO-CHAIR REMPE: Please go forward.
5 MR. LEWIS: Well, let me put it this way, 6 I've been listening to the ACRS since they went on the 7 telephone and I've been attending ACRS meetings as a 8 member of the public well before then, back to the 9 1970s. And I have to admit, today blew me away. I've 10 heard better questions from the ACRS membership than 11 I have ever heard before. And I thank you.
12 CO-CHAIR REMPE: Thank you for your 13 comment. Are there any other comments? So, with 14 that, I need to go around to each --
15 MS. FIELDS: Yes.
16 CO-CHAIR REMPE: Oh, sorry.
17 MS. FIELDS: This is Sarah Fields. I'd 18 like to make a comment. Yes, I also believe that 19 there were some very good questions, which I hope will 20 be answered down the line.
21 Going back to the Chapter 9 presentation, 22 9.1-2, the slide indicated that the COL applicant will 23 demonstrate that an NRC licensed cask can be lowered 24 into the dry dock and used to remove spent fuel 25 assemblies from the plant. And then, 9.1-3, they also NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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182 1 reference the fuel being transferred to a cask.
2 And I found that there seems to be 3 confusion between the term canister and cask. My 4 understanding is that fuel first goes into a metal 5 canister and is later transferred into a thicker 6 containment cask.
7 So, I think that that should be clarified, 8 whether fuel within the containment vessel is going to 9 be transferred to a metal canister and then, possibly, 10 later going into a cask.
11 That should be clarified, because a lot of 12 discussions in other dimensions related to spent fuel, 13 these fuels are confused and they should really be 14 specifically defined.
15 CO-CHAIR REMPE: Thank you for your 16 comment. Are there any other comments, individuals on 17 the line? And I'm not seeing anyone in the room that 18 has comments.
19 And so, we're going to go around and ask 20 the members for comments. In particular, if there's 21 some issues that rise to being mentioned in the 22 progress or interim letter we're going to be writing, 23 please emphasize those as we go around the room.
24 I heard several topics discussed today 25 that I think might warrant being mentioned in this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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183 1 interim letter, but it would be good for you to 2 highlight them if I've missed them. And I'm going to 3 start with Walt.
4 MEMBER KIRCHNER: For this section, where 5 I have questions, I think, on tech specs, I will take 6 note and bring them up in Chapter 4 and 5 and 6, which 7 we haven't yet reviewed. Issues like leak-before-8 break and breaks within containment.
9 And, here, we have things like leakage 10 rates into the containment, but there's the more 11 fundamental aspect of design, before you get to the 12 leak, that is something that we can take up with the 13 applicant and the staff when we get to Chapters 4, 5, 14 and 6. Thank you.
15 CO-CHAIR REMPE: Thank you. Dennis?
16 MEMBER BLEY: Nothing to add.
17 CO-CHAIR REMPE: Mike?
18 MEMBER CORRADINI: Nothing to add at this 19 time.
20 CO-CHAIR REMPE: Dick?
21 CO-CHAIR SKILLMAN: Nothing further, it's 22 already on the record and --
23 CO-CHAIR REMPE: Okay. So, you've 24 mentioned several points and you don't have any 25 requests that you'd like to have emphasized in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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184 1 full committee meeting? Because if we're going to 2 have anything documented in the letter, this would be 3 a good time to summarize some of those points.
4 CO-CHAIR SKILLMAN: Well, I'm going to pull 5 those points out of the record when I see it.
6 CO-CHAIR REMPE: Okay.
7 CO-CHAIR SKILLMAN: I've identified them 8 contemporaneously, my concern with homogeneity, the 9 crane, the floor, the structure, not the strength of 10 the floor, but the membrane vulnerability to load 11 drop, and several others such as those. And those are 12 already in the record --
13 CO-CHAIR REMPE: Okay.
14 CO-CHAIR SKILLMAN: -- so I'll pluck those 15 from the record.
16 CO-CHAIR REMPE: Okay, thank you.
17 CO-CHAIR SKILLMAN: Thank you.
18 CO-CHAIR REMPE: Harold?
19 MEMBER RAY: Our last staff presentation 20 included the excellent comment that the COL applicant 21 needs to know exactly what they need to do. And I 22 just have some concern on the larger areas covered and 23 I think with regard to tech specs, that's being 24 addressed, it seems, from what we've heard.
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185 1 where a design certification credits an assumption, 2 that a COL applicant will comply with requirements 3 which are yet to be identified and will be provided 4 later by the certificate holder, then that needs to be 5 explicit.
6 We can't answer all their questions at the 7 time of certification, but the fact that things are 8 going to be provided later, requirements are going to 9 be imposed on the COL applicant later, or provided to 10 the COL applicant later, needs to be identified. And 11 that's all I want to say at this point.
12 CO-CHAIR REMPE: Thank you. Jose?
13 MEMBER MARCH-LEUBA: Same as Dick, my 14 comments are on the record and I will provide you with 15 some comments for the letter later, when we can review 16 them.
17 CO-CHAIR REMPE: Okay. Ron?
18 MEMBER BALLINGER: No further comment.
19 CO-CHAIR REMPE: Charlie?
20 MEMBER BROWN: No further comment.
21 CO-CHAIR REMPE: Vesna?
22 MEMBER DIMITRIJEVIC: No further comment.
23 CO-CHAIR REMPE: Before I get to Steve, 24 what about Matt?
25 MEMBER SUNSERI: I don't have any comments, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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186 1 thanks, Joy.
2 CO-CHAIR REMPE: Pete?
3 MEMBER RICCARDELLA: I don't have any 4 comments at this time, but I will bring out some 5 thoughts and concerns I have at the letter writing, 6 during the letter writing.
7 CO-CHAIR REMPE: Okay. So, let me hear 8 from Steve, but hold that thought, I want to make a 9 comment, but go ahead, Steve?
10 DR. SCHULTZ: No further comments that 11 aren't on the record.
12 CO-CHAIR REMPE: Okay. So, we will be 13 having an interim letter and waiting until that letter 14 is always a little difficult. So, I know, Jose and 15 Dick and Harold, if you'll send, and you also, Pete, 16 some draft comments, it'll make it easier.
17 And I think Matt's going to combine this 18 with the letter for 10, 11, and 12. So, copy me, but 19 also send them to Matt, okay? Is that your 20 understanding, the process, Mike, that you'd 21 envisioned, that he would be leading --
22 MEMBER CORRADINI: I would appreciate it if 23 I got to know what the members want.
24 CO-CHAIR REMPE: That would be true, too.
25 Yes, copy Mike, Matt, and me.
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187 1 MEMBER CORRADINI: But from the standpoint 2 that we have, I think, on the full committee, we have 3 the morning Thursday for the staff and the applicant 4 to present.
5 So, to the extent that there are certain 6 things we want to put in the letter, because again, my 7 personal view is, if it's something that's 8 significant, we want to issue it. If it's something 9 that is a matter of discussion, we can enter it, but 10 I'm looking for things significant that deviate from 11 the open items, that we want or the staff wants.
12 CO-CHAIR REMPE: So, I think, Dick and 13 Harold and Jose have emphasized those, about how you 14 check about the boron mixing and different things like 15 that, they've emphasized today, and that's why I 16 wanted to have them bring it up again as we went 17 around the circle. Is there anything else I've 18 forgotten, Mike Snodderly?
19 MR. SNODDERLY: No, I'll just remind the 20 Committee of the bylaws on letter writing and that 21 when these comments are circulated, it should not 22 involve more than the majority, no more than -- it has 23 to be a less than a minority. Or then, you should 24 wait until the open letter writing session for further 25 discussions.
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188 1 CO-CHAIR REMPE: Well, yes, we are just 2 supposed to have it go to the Subcommittee, but the 3 draft letter should be a product of the Subcommittee, 4 so it is fine to have it sent to us that way. Okay?
5 So, thank you and with that, we're going to adjourn.
6 (Whereupon, the above-entitled matter went 7 off the record at 12:29 p.m.)
8 9
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LO-0319-64871 March 14, 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 Presentation: NuScale Chapter 9, Auxiliary Systems, PM-0319-64807, 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 March 20, 2019.
The materials support NuScales presentation of Chapter 9, Auxiliary Systems, of the NuScale Design Certification Application. is the presentation entitled ACRS Presentation: NuScale Chapter 9, Auxiliary Systems, PM-0319-64807, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@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 Getachew Tesfaye, NRC, OWFN-8H12 : ACRS Presentation: NuScale Chapter 9, Auxiliary Systems," PM-0319-64807, 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-0319-64871 :
ACRS Presentation: NuScale Chapter 9, Auxiliary Systems," PM-0319-64807, 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 Presentation:
NuScale Chapter 9, Auxiliary Systems Scott Harris, PE Supervisor, Mechanical Systems Corrie Nichol, Ph.D.
Refueling and Remote Handling Carrie Fosaaen Supervisor, Licensing John Fields Licensing Engineer March 20, 2019 1
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Chapter 9: Auxiliary Systems Section Title 9.1 Fuel Storage and Handling 9.2 Water Systems 9.3 Process Auxiliaries 9.4 Air Conditioning, Heating, Cooling, and Ventilation Systems 9.5 Other Auxiliary Systems 9A Fire Hazard Analysis 2
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9.1 - Fuel Storage and Handling Section Title 9.1.1 Criticality Safety of Fresh and Spent Fuel Storage and Handling 9.1.2 New and Spent Fuel Storage 9.1.3 Spent Fuel Pool Cooling and Cleanup System 9.1.4 Fuel Handling Equipment 9.1.5 Overhead Heavy Load Handling Systems 3
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9.1.1 - Criticality Safety of Fresh and Spent Fuel Storage and Handling
- Storage rack design does not require loading patterns or zones
- Spent fuel pool connected to refueling and reactor pools which forms ultimate heat sink - large volume prevents undetected addition of unborated water
- Max keff remains below applicable limits and no abnormal condition causes inadvertent criticality (TR-0816-49833, Rev. 1.)
- Fuel handling procedures place controls on movement of each fuel assembly and the designated storage location in a rack
- TS 4.3 - Fuel Storage
- TS 5.5.12 - Spent Fuel Storage Rack Neutron Absorber Monitoring Program
- COL Item 9.1-1: A COL applicant will develop plant programs and procedures for safe operations during handling and storage of new and spent fuel assemblies, including criticality control 4
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9.1.2 - New and Spent Fuel Storage
- Storage and handling facility for new and spent fuel assemblies located in the RXB - racks can store new or spent fuel assemblies
- The fuel storage racks, spent fuel pool, liner, and RXB meet Seismic Category I requirements and protected from non-Seismic Category I SSCs
- 9.1-2: A COL applicant will demonstrate that an NRC-licensed cask can be lowered into the dry dock and used to remove spent fuel assemblies from the plant
- 9.1-8: A COL applicant will submit an evaluation of the spent fuel storage racks that addresses structural, thermal-hydraulic, criticality, and material analysis aspects of the design. This evaluation is dependent on a vendor-specific design and the as-built configuration of spent fuel storage racks. The design of the spent fuel storage racks is considered acceptable when it meets the criteria of Appendix D to Design Specific Review Standard 3.8.4.
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Ultimate Heat Sink Configuration 6
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9.1.3 - Spent Fuel Pool Cooling and Cleanup System Spent Fuel Pool Cooling System and Reactor Pool Cooling System
- Performs the following nonsafety-related functions:
- Maintains reactor, refueling pool and spent fuel pool temperature during normal operations and refueling
- Maintains ultimate heat sink water level by providing makeup from the demineralized water system
- Provides chemistry control of pool including boron concentration
- Provides reactor pool temperature information signals for post-accident monitoring 7
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9.1.3 - Spent Fuel Pool Cooling and Cleanup System Pool Cleanup System
- Performs the following nonsafety-related functions:
- removes impurities to reduce radiation dose rates and maintain water chemistry/clarity in pools and dry dock Pool Surge Control System
- Performs the following nonsafety-related functions:
- drains the dry dock using the evacuation pumps to support maintenance and refueling activities
- transfers and stores excess water volume from the UHS to maintain pool level during surge events such as when an NPM is added to the pool 8
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9.1.3 - Spent Fuel Pool Cooling and Cleanup System Pool Leakage Detection System
- Performs the following nonsafety-related functions:
- provides for collection of water leaking from the pool liner
- directs the flow to sumps for detection of collected leakage for operator evaluation
- Valves used to isolate each channel drainage line and leakage rate measuring line
- Channels collect leakage from pool wall and floor liner plates and direct it to a sump in radioactive waste drain system and routed to liquid radioactive waste system for further processing 9
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9.1.4 - Fuel Handling Equipment
- Functions
- Receipt of new fuel
- Refueling operations
- Loading of spent fuel into a dry storage cask
- Design considerations - ANSI/ASME design requirements
- Dropping fuel assembly following a safe shutdown earthquake minimized
- Radiological shielding for personnel who operate the equipment
- Protection from a criticality event
- Capability to permit periodic inspection and testing of components
- Speed of the fuel handling machine, trolley, and hoist motions are limited 10 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.4 - Fuel Handling Equipment 11 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.4 - Fuel Handling Equipment COL Items:
- 9.1-3: A COL applicant will develop procedures related to the transfer of spent fuel to a transfer cask
- 9.1-4: A COL applicant will provide the periodic testing plan for fuel handling equipment 12 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.5 - Overhead Heavy Load Handling Systems
- Major components:
- Reactor Building Crane (RBC)
- NPM lifting fixture - comprised of the NPM lifting lugs and top support structure diagonal lifting braces - permanently installed
- module lifting adapter (MLA) - connection for crane to lift and carry the NPM
- wet hoist - main below-the-hook structure for lifting equipment underwater
- RBC:
- Designed as a single-failure-proof crane - highest qualification
- Movement uses position control system interlocks
- Path and the maximum height limited - shielding maintained
- Communication system provided between the MCR and RBC 13 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.5 - Overhead Heavy Load Handling Systems 14 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.5 - Overhead Heavy Load Handling Systems Equipment Rated Design Code Single Seismic Max Max Max Capacity Failure Cat Traverse Hoist Lift (Tons) Proof Speed Speed Height (fpm) 1 (fpm) 1 (ft)
RBC Main 850 ASME NOG-1, Y I 30 2 50 Hoist Type I RBC Aux 15 ASME NOG-1, Y I N/A 252 / 373 65 Hoist Type I MLA 790 ANSI N14.6 Y N/A N/A N/A N/A NPM Lift 790 ANSI N14.6 Y I N/A N/A N/A Fixture Wet Hoist 250 ASME NOG-1, Y N/A N/A 4 75 Type I Jib Crane 1 ASME NUM-1, N II 30 21 40 Hoist Type II Notes
- 1. Bridge, trolley and hoist speeds are within recommended ranges of ASME NOG-1
- 2. With load on hoist
- 3. Without load on hoist 15 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.1.5 - Overhead Heavy Load Handling Systems
- COL Items:
- 9.1-5: The COL applicant will describe the process for handling and receipt of critical loads including NPMs
- 9.1-6: The COL applicant will provide a design for a spent fuel cask and handling equipment including procedures and programs for safe handling.
- 9.1-7: The COL applicant will provide a description of the program governing heavy loads handling. The program should address:
- operating and maintenance procedures
- inspection and test plans
- personnel qualifications and operator training
- detailed description of the safe load paths for movement of heavy loads 16 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.2 - Water Systems Section Title 9.2.1 Station Service Water System - not in NuScale Design 9.2.2 Reactor Component Cooling Water System 9.2.3 Demineralized Water System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Condensate Storage Facilities 9.2.7 Site Cooling Water System 9.2.8 Chilled Water System 9.2.9 Utility Water Systems 17 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.2 - Water Systems
- Common Elements of NuScale Water System design (except for 9.2.5 - UHS)
- Does not perform safety-related or risk-significant functions
- Classified as nonsafety-related
- Not an essential source of water to prevent or mitigate the consequences of accidents or to shut down the reactor and maintain it in a safe-shutdown condition
- Not required during or after a natural phenomenon event
- Includes provisions that ensure system failure will not adversely affect the functional performance of safety-related systems or components
- Components in proximity to Seismic Category I SSCs designed to Seismic Category II 18 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.2.5 - Ultimate Heat Sink
- Set of stainless steel-lined, reinforced concrete pools of borated water located below grade in the RXB (reactor, refueling, and SFP)
- Sized such that active cooling systems are not required for accident conditions; the combined volume of water in the UHS provides sufficient cooling for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without additional makeup water
- Provided with a seismically qualified makeup line that can provide additional water to the UHS and with redundant water level instrumentation
- During normal plant operations, heat is removed from the UHS through the reactor pool cooling system and rejected into the atmosphere through a cooling tower or other external heat sink
- In a DBA involving a sustained loss of all AC power, decay heat is removed from the NPMs through passive heat transfer to the pool
- TS 3.5.3 - Ultimate Heat Sink 19 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.2 - Water Systems
- COL Items:
- 9.2-1: A COL applicant will select the appropriate chemicals for the RCCWS based on site-specific water quality and materials requirements
- 9.2-2: A COL applicant will describe the source and pre-treatment methods of potable water for the site, including the use of associated pumps and storage tanks
- 9.2-3: A COL applicant will describe the method for sanitary waste storage and disposal, including associated treatment facilities
- 9.2-4: A COL applicant will provide details on the prevention of long-term corrosion and organic fouling in the SCWS
- 9.2-5: A COL applicant will identify the site-specific water source and provide a water treatment system that is capable of producing water that meets the plant water chemistry requirements 20 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3 - Process Auxiliaries Section Title 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Equipment and Floor Drain Systems 9.3.4 Chemical and Volume Control System 9.3.5 Standby Liquid Control System (Applicable only to BWRs) 9.3.6 Containment Evacuation System and Containment Flooding and Drain System 21 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3 - Process Auxiliaries
- Common Elements of NuScale Process Auxiliaries Design
- No safety-related or risk significant functions, not credited for mitigation of design basis accidents, and has no safe-shutdown functions (except 9.3.4 - CVCS)
- SSCs whose failure could adversely affect Seismic Category I SSCs during or following an SSE are designed as Seismic Category II
- COL Item 9.3-1: COL applicant submits a leakage control program
- COL Item 9.3-2: COL applicant develops PASS contingency plans for using the PSS and the CES off-line radiation monitor to obtain reactor coolant and containment atmosphere samples 22 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3.4 - Chemical and Volume Control System
- CVCS purifies reactor coolant, manages chemistry of the coolant (including boron concentration), provides reactor coolant inventory makeup and letdown, and supplies spray flow to the pressurizer to reduce RCS pressure
- CVCS is provided for each NPM
- CVCS is used in combination with the MHS during startup to raise reactor coolant temperature and to generate natural circulation flow in the RCS before nuclear heat addition
- BAS is a shared system for all 12 NPMs
- BAS prepares, stores, and transfers borated water for use by the CVCS or by the SFPCS for adding boron to the SFP as needed 23 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3.4 - Chemical and Volume Control System
- CVCS is only system with connections to the RCS and piping that runs outside containment
- CIV function part of CNTS - Section 6.2
- CVCS is equipped with two automatic, safety-related, fail-closed, demineralized water isolation valves to ensure CVCS operation does not inadvertently cause a dilution of the RCS boron concentration
- CVCS required for NPM operation per TS 3.4.6 24 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3.6 - Containment Evacuation System and Containment Flooding and Drain System
- CES and CFDS are used to transfer liquids and gases between the CNV free volume and other plant systems.
- CES establishes and maintains a vacuum in the CNV by removing water vapor and non-condensable gases from the CNV
- CES Performs Leakage detection function
- CFDS is used to flood a CNV with borated reactor pool water after shutdown in preparation for NPM refueling and to drain water back to the reactor pool in preparation for NPM startup
- CFDS can provide borated coolant inventory to containment during BDBE for decay heat removal
- CIV function part of CNTS - Section 6.2
- CES required for NPM operation per TS 3.4.7, 3.4.9 & 3.7.3 25 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3.6 - CES Flow Diagram 26 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.3.6 - CFDS Flow Diagram 27 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.4 - Air Conditioning, Heating, Cooling, and Ventilation Systems Section Title 9.4.1 Control Room Area Ventilation System 9.4.2 Reactor Building and Spent Fuel Pool Area Ventilation System 9.4.3 Radioactive Waste Building Ventilation 9.4.4 Turbine Building Ventilation System 9.4.5 Engineered Safety Feature Ventilation System - not in NuScale Design 28 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.4 - Air Conditioning, Heating, Cooling, and Ventilation Systems
- Common Elements of NuScale Ventilation System Design
- No safety-related or risk significant functions, not credited for public dose mitigation of design basis accidents, and has no safe-shutdown functions
- Components whose failure could adversely affect Seismic Category I SSC or could result in incapacitating injury to occupants of the control room during or following an SSE are designed as Seismic Category II
- A COL item for each system to specify periodic testing and inspection program requirements (COL Items 9.4-1, 9.4-2, 9.4-3 and 9.4-4) 29 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.5 - Other Auxiliary Systems Section Title 9.5.2 Communication System 9.5.3 Lighting Systems 9.5.1 Fire Protection Program 9.A Fire Hazards Analysis 30 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.5.2 - Communication System
- COMS includes components for intra-plant and plant-to offsite communications. Communication methods include private branch exchange, a public address and general alarm system, sound-powered telephones, distributed antenna, and radios for point-to-point communication between plant personnel in vital areas of plant under expected conditions.
- COMS serves no safety-related or risk-significant functions
- COMS not credited for mitigation of design basis accidents and has no safe-shutdown functions
- COL Item 9.5-1: COL applicant provides a description of the offsite communication system
- COL Item 9.5-2: COL applicant determines the location for the security power equipment within a vital area 31 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.5.3 - Lighting Systems
- PLS provides artificial illumination for buildings, rooms, spaces, and outdoor areas of the plant
- PLS functions include normal plant lighting, emergency plant lighting, and normal and emergency MCR lighting
- PLS provides illumination for each area of the plant in accordance with the guidelines of NUREG-0700
- MCR emergency lighting is fed from the highly reliable DC power system for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- PLS provides emergency lighting with battery pack for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for egress or exiting the area in accordance with NFPA 804
- PLS provides emergency lighting with battery pack for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the post-fire safe shutdown activities in areas outside the MCR and RSS (if required)
- No COL Items 32 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9.5.1 - Fire Protection Program
- Objective:
- Use the concept of defense-in-depth to achieve the required degree of reactor safety by using administrative controls, fire protection systems and features, and redundant safe shutdown capability
- Safe shutdown relies on passive fire protection - redundant safe shutdown equipment is separated by 3-hour fire barriers
- Defense-in-depth principles achieve the following:
- To prevent fires from starting;
- To rapidly detect, control, and extinguish promptly those fires that do occur; and
- To provide protection for SSCs important to safety so that a fire that is not promptly extinguished by the fire suppression activities does not prevent the safe shutdown of the plant 33 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9A - Fire Hazards Analysis
- FHA demonstrates that the plant maintains the ability to perform safe shutdown functions and minimize radioactive material releases to the environment in the event of a fire
- FHA has the following objectives:
- consider in situ and transient fire hazards
- determine the effects of a fire in any location in the plant on the ability to safely shut down the reactor or to minimize and control the release of radioactivity to the environment
- specify measures for fire prevention, detection, suppression, and containment for each fire area containing safety-related and risk-significant SSCs, in accordance with NRC guidance and regulations 34 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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9A - Fire Hazards Analysis
- FHA Special Cases
- Main Control Room Fire Area
- Manual switches that provide backup control of systems automatically controlled by the MPS are located in the MCR
- MCR safety-related manual switches associated with MPS actuations (e.g., reactor trip and containment isolation) are isolated using switches located outside the MCR
- Containment Fire Area
- Fire in containment is highly unlikely - held at a vacuum, electrical conductors are of noncombustible construction or routed in conduit, inaccessible during operation
- Flooding of containment for heat removal during shutdown prevents ignition or spread of fire 35 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
9A - Fire Hazards Analysis
- FHA Special Cases
- Top-of-Module Fire Area
- Fire on the top of the NPM is highly unlikely - enclosed in bioshield, all cabling under bioshield is in conduit or 3-hr rated cable
- Top of NPM is inaccessible during power operations
- Hydraulic fluid for valve controls is noncombustible
- CIV/DHRS valve actuation based on stored pressurized nitrogen accumulators on each valve - places valve in its safe position when hydraulic operating pressure is relieved by the remotely located hydraulic unit
- Module has 3-hr fire barrier on each side with exception of bioshield wall facing the pool - Greater than 20 feet between intervening combustibles on pool side
- During shutdown bioshield is removed and administrative controls limit transient combustibles on the top of the NPM 36 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
9.5.1 and 9A
- Generic COL items related to FP program development, procedure development, implementation schedule, QA program and FP training and qualifications
- Provide Final Safe Shutdown Analysis - based on as-built design
- cable routing
- COL applicant is required to provide a site specific, as-built FHA per RG 1.189
- COL applicant is required to provide a site specific, as-built SSA per RG 1.189 37 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
- ANB - Annex Building
- CNTS - Containment System
- ANSI - American National Standards
- CNV - Containment Vessel Institute
- COL- Combined License
- ASME - American Society of Mechanical Engineers
- COMS - Communication System
- BAS - Boron Addition System
- CRB - Control Building
- BDBE - Beyond Design Basis Event
- CRDM - Control Rod Drive Mechanism
- BPDS - Balance-of-Plant Drain System
- CAS - Compressed Air Systems
- CRHS - Control Room Habitability System
- CES - Containment Evacuation System
- CST - Condensate Storage Tank
- CFDS - Containment Flooding and Drain
- CUB - Central Utility Building System
- CVCS - Chemical and Volume Control
- CFR - Code of Federal Regulations System
- CFWS - Condensate and Feedwater
- DBA - Design Basis Accident System
- DHRS - Decay Heat Removal System
- CHWS - Chilled Water System
- DWS - Demineralized Water System
- CIVs - Containment Isolation Valves 38 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
- EFDS - Equipment and Floor Drain
- LRWS - Liquid Radioactive Waste System Systems
- MCR - Main Control Room
- ESF - Engineered Safety Features
- MHA - Maximum Hypothetical Accident
- FDS - Fire Detection System
- MHS - Module Heatup System
- FHA - Fire Hazards Analysis
- MLA - Module Lifting Adapter
- MPS - Module Protection System
- FPS - Fire Proteciton System
- NDS - Nitrogen Distribution System
- FSS - Fire Suppression System
- NFPA - National Fire Protection
- GRWS - Gaseous Radioactive Waste Association System
- NPM - NuScale Power Module
- HEPA - High Energy Particulate Air
- OHLHS - Overhead Heavy Load Handling
- HVAC - Heating, Ventilating and Air Systems Conditioning
- PASS - Post-Accident Sampling System
- IAS - Instrument Air System
- PLS - Plant Lighting System
- IEEE - Institute of Electrical and Electronic Engineers
- LOCA - Loss of Coolant Accident
- PSS - Process Sampling System 39 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
- PWS - Potable Water System
- SSC - Structure System or Component
- RBC - Reactor Building Crane
- TGB - Turbine Generator Building System
- TS - Technical Specifications
- RCCWS - Reactor Component Cooling Water System
- RG - Regulatory Guide
- UWS - Utility Water System
- RWB - Radioactive Waste Building
- RWDS - Radioactive Waste Drain System
- RXB - Reactor Building
- SAS - Service Air System
- SCWS - Site Cooling Water System
- SFP - Spent Fuel Pool
- SFPCS - Spent Fuel Pool Cooling System
- SSA - Safe Shutdown Analysis 40 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 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 41 PM-0319-64807 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presentation to the ACRS Subcommittee NuScale Power, LLC (NuScale)
Design Certification Application Review Safety Evaluation with Open Items: Chapter 9 AUXILIARY SYSTEMS March 20, 2019
Technical Reviewers:
Alexandra Siwy - NRO/DSRA/SRSB Alissa Neuhausen - NRO/DEI/SEB Ryan Nolan - NRO/DSRA/SRSB Raul Hernandez - NRO/DSRA/SCPB Hanry Wagage - NRO/DSRA/SCPB Chang Li - NRO/DSRA/SCPB Angelo Stubbs - NRO/DSRA/SCPB Bob Vettori - NRO/DSRA/SPRA Edward Stutzcage - NRO/DLSE/RGRB Tony Gardner - NRR/DMLR/MCCB Alexander Chereskin - NRR/DMLR/MCCB Fanta Sacko - NRR/DE/EENB Andrew Yeshnik - NRO/DEI/MCB Nan Chien - NRO/DSRA/SCPB Dawnmathews Kalathiveettil - NRR/DE/EICA Project Managers:
Greg Cranston - Lead Project Manager Getachew Tesfaye - Chapter 9 Project Manager March 20, 2019 Chapter 9, Auxiliary Systems 2
Chapter 9: Auxiliary Systems Technical Reviewers:
Alexandra Siwy - SRSB Raul Hernandez - SPSB Section 9.1 - Fuel Storage and Handling Alissa Neuhausen - SEB Andrew Yeshnik - MCB Chang Li - SCPB Angelo Stubbs - SCPB Section 9.2 - Water Systems Nan Chien - SCPB Bob Vettori - SPRA Raul Hernandez - SCPB Tony Gardner - MCCB Edward Stutzcage - RGRB Section 9.3 - Process Auxiliaries Bob Vettori - SPRA Hanry Wagage - SCPB Ryan Nolan - SRSB Section 9.4 - Air Conditioning, Heating, Cooling, Nan Chien - SCPB and Ventilation Systems Robert Vettori - SPRA Dawnmathews Kalathiveettil - EICA Section 9.5 - Other Auxiliary Systems Fanta Sacko - EENB Nan Chien - SCPB March 20, 2019 Chapter 9, Auxiliary Systems 3
Section 9.1: Fuel Storage and Handling Subsections:
9.1.1 - Criticality Safety of New and Spent Fuel Storage 9.1.2 - Fuel Storage System 9.1.3 - Spent Fuel Pool Cooling and Cleanup System 9.1.4 - Light and Heavy Systems and Heavy Load Handling System March 20, 2019 Chapter 9, Auxiliary Systems 4
9.1.1 - Criticality Safety of New and Spent Fuel Storage (1/2)
Review objective: verify that fuel will remain subcritical during storage and handling in accordance with 10 CFR 50.68 and GDC 62
- Review guidance: SRP Section 9.1.1, Revision 3
- Review areas: computational methods and validation, scope of conditions analyzed, analysis models and assumptions, handling of bias and uncertainty, analysis results, storage rack materials Staff Review and Findings:
- Criticality calculation methodology is acceptable and adequately benchmarked
- Criticality models correctly incorporate design information and use appropriate assumptions
- Open Item 9.1.1-2: Applicant is making the structural analysis of the racks a COL item, which the staff will evaluate in Phase 4 March 20, 2019 Chapter 9, Auxiliary Systems 5
9.1.1 - Criticality Safety of New and Spent Fuel Storage (2/2)
Staff Review and Findings, continued:
- Except for analyses related to Open Item 9.1.1-2, applicants criticality analyses comply with 10 CFR 50.68 and GDC 62
- Staff confirmatory analyses support the applicants results
- Tier 1 information complete and consistent with Tier 2 information
- Technical Specifications adequately protect the assumptions in the criticality analysis
- NuScale has not specified a Neutron Absorbing material.
Open item 9.1.1-1: Request additional details on the impacts of manufacturing on the neutron attenuation, materials qualification for the SFP environment, and manufacturing process controls.
- The applicant is implementing a neutron absorber monitoring program that is consistent with TSTF-577, Rev. 1, and utilizes NEI 16-03, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools.
March 20, 2019 Chapter 9, Auxiliary Systems 6
9.1.2 - New and Spent Fuel Storage (1/1)
Review Objective
- The SFP has the necessary design features unique to fuel storage during initial receipt, refueling operations, and accident conditions, including maintaining cooling and limiting offsite exposure in the event of a fuel handling accident.
Items of major interest
- Shared system between all 12 modules Open Items
- 9.1.2 Related to the pool leakage detection system capability to monitor the UHS liner integrity - Based on latest RAI response and proposed DCA changes, the staff finds this Open Item Closed.
- Staff finds that the SFP conforms with the guidance of SRP 9.1.2 March 20, 2019 Chapter 9, Auxiliary Systems 7
9.1.3 - Spent Fuel Pool Cooling and Cleanup System (1/1)
Review Objective
- Ensure the safety-related Spent Fuel Pool Cooling System is capable of maintaining the spent fuel assemblies cooled and covered with water during all storage conditions.
Items of major interest
- Staff reviewed SFPCCS in accordance with the guidance in SRP 9.1.3
- Ensure system configuration ensures adequate water inventory in the SFP
- Evaluate all key assumptions of the thermal analyses
- Cooling capability during accident scenarios addressed in Section 9.2.5 UHS
- Staff finds that the SFPCS conforms with the guidance of SRP 9.1.3 March 20, 2019 Chapter 9, Auxiliary Systems 8
9.1.4 and 9.1.5 - Light Load Handling Systems (Related to Refueling) (1/1)
Review Objective To ensure the design and operation of the load handling systems can prevent or minimize the likelihood of an event that could cause a release of radioactivity, a criticality accident, or an inability to cool fuel within the reactor vessel or spent fuel pool; or could prevent the safe shutdown of the reactor.
Items of major interest
- Responses of load handling equipment to a safe shutdown earthquake (SSE) event consistent with guidance in SRPs 9.1.4 and 9.1.5.
- Design criteria for single-failure-proof load handling components consistent with guidelines of NUREG-0554
- Procedures for safe load handling operations consistent with guidelines of NUREG-0612
Staff finds that the light and heavy load handling conforms with the guidance of SRPs 9.1.4 and 9.1.5.
March 20, 2019 Chapter 9, Auxiliary Systems 9
Section 9.2 - Water Systems Subsections:
9.2.1 - Station Service Water System 9.2.2 - Component Cooling Water System 9.2.3 - Demineralized Water System 9.2.4 - Potable and Sanitary Water Systems 9.2.5 - Ultimate Heat Sink 9.2.6 - Condensate Storage Facilities 9.2.7 - Site Cooling Water System 9.2.8 - Chilled Water Systems 9.2.9 - Utility Water Systems March 20, 2019 Chapter 9, Auxiliary Systems 10
9.2.1 - Station Service Water System (1/1)
Key Design Features and Considerations
- The NuScale passive design does not rely on active systems such as a service water system to provide cooling to essential equipment.
- The NuScale Power Modules are partially submerged in the reactor pool portion of the plant UHS.
- This design configuration ensures passive heat transfer from essential systems and components directly to the UHS, with no intermediate heat transfer loop such as that provided by a typical LWR essential service water system.
Staffs Review and Evaluation Results
- The staff reviewed the NuScale system design and confirmed the above statements.
- The UHS, reviewed in Section 9.2.5 of this SER, performs the safety function of decay heat removal, which the service water system usually performs in LWR active designs.
- The site cooling water system (SCWS), reviewed in Section 9.2.7 of this SER, performs the heat removal function for the nonsafety-related systems, which the service water system usually performs in LWR designs.
March 20, 2019 Chapter 9, Auxiliary Systems 11
9.2.2 - Component Cooling Water System (1/1)
Key Design Features and Considerations
- RCCWS is a closed loop cooling system comprised of two identical subsystems each supporting up to six reactor modules.
- RCCWS provides cooling to the CVCS non-regenerative heat exchanger, CES condenser and pumps, the PSS coolers and analyzer control units and the CRDMs electromagnetic coils housing .
- RCCWS is not safety-related, however the CVCS, CES, and PSS components cooled by the RCCWS are located in the RXB, the CRDM electromagnetic coils, are located inside containment and outside of the reactor vessel.
Staffs Review and Evaluation Results
- Staff reviewed CCWS in accordance with SRP 9.2.2. Since the system is not credited for performing any safety related function the review focused on system design that ensures system failure would not impact important to safety SSCs.
- The system was found to be in compliance with GDCs 2 and 4 based on the RCCWS SSCs inside the reactor building and containment having appropriate seismic classifications and having appropriate environmental qualifications . The staff also found the system to be in compliance with GDCs 60 , 64.
March 20, 2019 Chapter 9, Auxiliary Systems 12
9.2.3 - Demineralized Water System (1/1)
Key Design Features and Considerations
- The DWS is designed to treat the water from the utility water system (UWS) and provide and distribute high-quality demineralized water to the plant.
- The DWS provides plant support during abnormal conditions by providing additional makeup water to the SFPCS to compensate for inventory loss, and to the condenser for emergency fill.
- The DWS is not safety-related and is not required for any DBE.
Staffs Review and Evaluation Results
- Staff reviewed the systems in accordance with SRP 9.2.3
- The staff found the design to be in accordance with the guidance in SRP 9.2.3, and in compliance with GDCs 2, 5, 60 and 10CFR 20.1406.
March 20, 2019 Chapter 9, Auxiliary Systems 13
9.2.4 - Potable and Sanitary Water Systems (1/1)
Key Design Features and Considerations
- Systems provide potable water for domestic use and human consumption and collects and transfers site sanitary waste for treatment and discharge.
- The PWS piping in the control building penetrating the control room envelope and habitability boundary.
- Systems are site specific therefore only conceptual design information, interface requirements and design features to prevent radioactive contamination of the system is included in the application.
Staffs Review and Evaluation Results
- Staff reviewed the systems in accordance with SRP 9.2.4
- The staff found the conceptual design information and the interface requirements in the DCD to be in accordance with 10 CFR 52.47(a)
- Design provision made to prevent inadvertent contamination of the domestic water and sanitary systems with radioactive material was found to be acceptable and the design is in compliance with GDC 60.
- PWS uses passive design feature (loop seal) instead of isolation valves to ensure integrity of CR envelope.
March 20, 2019 Chapter 9, Auxiliary Systems 14
9.2.5 - Ultimate Heat Sink (1/1)
Key Design Features and Considerations
- The NuScale UHS is a set of safety-related pools of borated water that consists of the combined water volume of the reactor pool, RFP, and SFP.
- Up to 12 NPMs are located in the reactor pool and share the combined volume of water.
- During accident scenarios, the NuScale design credits the safety-related water inventory stored in the UHS to passively remove the decay heat.
- The applicant considered GDCs 2, 4, 5, 45, 46, 61, and PDC 44 in the design.
Staffs Review and Evaluation Results
- The staffs review confirmed that the UHS has sufficient water inventory for 30 days to remove the decay heat from the NPM and the stored fuel from the pool through boiling and evaporation, removing enough heat to maintain the spent fuel and fuel in the NPMs sufficiently cool to prevent fuel damage.
March 20, 2019 Chapter 9, Auxiliary Systems 15
9.2.6 - Condensate Storage Facilities (1/1)
Key Design Features and Considerations
- The CST provides a volume for makeup and rejection of condensate to and from the condenser hotwell based on hotwell level.
- The CST does not serve a safety function, and it does not interface with other systems that could adversely affect safety-related or augmented equipment.
- In the event of failure of non-seismic CSF storage tanks, site grading and drainage will ensure SSCs important to safety will not be adversely affected Staffs Review and Evaluation Results
- The system was found to be in compliance with GDCs 2 and 60 and 10 CFR 20.1406.
- GDCs 5, 44, 45, and 46 were found not applicable because system is not safety-related and not credited for supporting safety-related systems.
March 20, 2019 Chapter 9, Auxiliary Systems 16
9.2.7 - Site Cooling Water System (1/1)
Key Design Features and Considerations
- Transfer heat from plant auxiliary systems to the SCWS cooling towers.
- Service loads for the SCWS include equipment in the RXB, Central Utility Building, North and South TGB, and Auxiliary Boiler Building.
- SCWS does not provide cooling to safety-related or risk-significant SSCs, and is not credited for performing any safety-related functions.
Staffs Review and Evaluation Results
- The staff reviewed the SCWS to determine if system failures could potentially impact import to safety SSCs (GDCs 2, 4 & 5) or result in the release of radiation to environment (GDC 60).
- Staff concluded that the SCWS complies with GDC 2, 4, 5 and 60 as well as 10 CFR 20.1406.
March 20, 2019 Chapter 9, Auxiliary Systems 17
9.2.8 - Chilled Water System (1/1)
Key Design Features and Considerations
- Primary CHWS provides cooling for the normal CRVS, RWBVS, RBVS, and other equipment in the radioactive waste processing systems.
- Standby CHWS is dedicated only to the CRVS in the event of a loss of normal ac power.
Staffs Review and Evaluation Results
- Staff reviewed CHWS in accordance with SRP 9.2.7.
- The CHWS has the heat removal capacity to provide, during plant normal operation, a heat sink for various air handling units and cooling loads in the radioactive waste processing systems
- The CHWS design complies with GDCs 1, 2, 4, 5, 44, 10 CFR 20.1406, and 10CFR52.47(b)(1)
March 20, 2019 Chapter 9, Auxiliary Systems 18
9.2.9 - Utility Water Systems (1/1)
Area of Review
- Staff reviewed the Utility Water System in accordance with the applicable guidance in SRP 9.2.4.
- Staff reviewed to ensure compliance with GDC 5, 60, 64, and 10CFR 20.1406.
- RAIs issued, all closed/resolved. No open items.
Staffs Findings
- The Utility Water System is in accordance with the applicable guidance in SRP 9.2.4 and is in compliance with GDC 5, 60, 64 and 10CFR 20.1406.
March 20, 2019 Chapter 9, Auxiliary Systems 19
Section 9.3 : Process Auxiliaries Subsections:
- 9.3.1 - Compressed Air Systems
- 9.3.2 - Process and Post-Accident Sampling Systems
- 9.3.3 - Equipment and Floor Drain Systems
- 9.3.4 - Chemical and Volume Control System
- 9.3.5 - Not Applicable
- 9.3.6 - Containment Evacuation System and Containment Flooding and Drain System March 20, 2019 Chapter 9, Auxiliary Systems 20
9.3.1 - Compressed Air Systems (1/1)
Review Objective
- Ensure the quality of air does not prevent the safety-related air operated valves from performing their safety function Items of major interest
- Staff reviewed compressed air and gas systems in accordance with the guidance in SRP 9.3.1.
- CAS has no safety function, but the system follows the guidance that ensures high quality air is available.
- The CAS is shared between 12 modules simultaneously.
- Staff finds that the compress air and gas systems conforms with the guidance of SRP 9.3.1 March 20, 2019 Chapter 9, Auxiliary Systems 21
9.3.2 - Process and Post-Accident Sampling Systems (1/3)
System Description
The function of the process sampling system (PSS) is to provide the means to obtain representative liquid and gaseous samples from various primary and secondary process streams and components for monitoring and analyzing the chemical and radiochemical conditions.
Scope of Review
- Sampling locations and capabilities
- Detection and control of leakage outside containment
- Containment isolation March 20, 2019 Chapter 9, Auxiliary Systems 22
9.3.2 - Process and Post-Accident Sampling Systems (2/3)
Conclusions
- No dedicated post-accident PSS used. Normal PSS is used during post-accident conditions.
- There are sample points for collecting and analyzing liquid and gaseous samples from the primary coolant, secondary coolant, containment air, spent fuel pool, and other sources
- Includes the sampling locations in the SRP acceptance criteria, and other locations
- COL applicant will .
- Develop operational procedures and maintenance program for leak detection and contamination control
- Develop post-accident sampling contingency plans for using the process sampling system and the containment evacuation system off-line radiation monitor to obtain reactor coolant and containment atmosphere samples.
March 20, 2019 Chapter 9, Auxiliary Systems 23
9.3.2 - Process and Post-Accident Sampling Systems (3/3)
Conclusions (continued)
- The process of collecting liquid post-accident samples involves opening containment isolation valves and routing liquid to the CVCS system to the sampling lines for liquid samples.
- The process of collecting gaseous post-accident samples involves opening containment isolation valves and routing gases through the containment evacuation system to the sample lines and back to containment through the containment flood and drain system.
Staff has several open questions related to the post-accident sampling process, including radiological concerns related to post-accident sampling and the dose to operators when taking samples, in accordance with 10 CFR 50.34(f)(2)(viii).
However, NuScale has recently proposed an exemption from 10 CFR 50.34(f)(2)(viii), which is currently being evaluated by staff. The staff will re-assess the status of the open questions following the review of the exemption.
March 20, 2019 Chapter 9, Auxiliary Systems 24
9.3.3 - Equipment and Floor Drain Systems (1/1)
Area of Review
- The EFDS is comprised of two separate, unconnected systems, the radioactive waste drain system (RWDS,) and the balance-of-plant drain system (BPDS).
- Staff reviewed the EFDS is accordance with the guidance in SRP 9.3.3.
- Staff reviewed to ensure compliance with GDC 2, 4, 5, 60, 64, and 10CFR 20.1406.
- RAIs issued, all closed/resolved. No open items.
Staff Findings
- The EFDS conforms with the guidance of SRP 9.3.3 and is compliance with GDC 2, 4, 5, 60, 64, and 10CFR 20.1406.
March 20, 2019 Chapter 9, Auxiliary Systems 25
9.3.4 - Chemical and Volume Control System (1/3)
Findings
- Design Basis Only safety-related functions of the CVCS are containment and demineralized water isolation.
- Functionality
- System functional design to include all components necessary for RCS chemical and volume control
- Components, instrumentation and controls appropriate
- Adequate chemistry/purity control
- Utilized to establish natural circulation flow to support startup operations
- Protection
- Protection against boric acid precipitation adequate
- Appropriate seismic classification completed for CVCS
- Purifying equipment appropriately protected
- CVCS leakage control adequate to minimize contamination March 20, 2019 Chapter 9, Auxiliary Systems 26
9.3.4 - Chemical and Volume Control System (2/3)
- Underlying purpose of GDC 33 is met by:
CVCS isolation capability CNV and ECCS capability to maintain RCS inventory and core coolability to meet the SAFDLs March 20, 2019 Chapter 9, Auxiliary Systems 27
9.3.4 - Chemical and Volume Control System (3/3)
- CVCS injects into the lower riser 8 feet above the core.
- Dilution mixing methodology during startup is reviewed in Section 15.4.6.
- T-H phenomena for boron dilution in the long term is reviewed in Section 15.0.6.
March 20, 2019 Chapter 9, Auxiliary Systems 28
9.3.6 - Containment Evacuation System and Containment Flooding and Drain System (1/1)
Area of Review
- Staff reviewed the CES and CFDS design function of transferring liquids and gases between the containment vessel free volume and other plant systems Staff Findings
CFDS and CVCS systems provide single failure proof capability to cool the nuclear power module from normal operating conditions to conditions equivalent to cold shutdown in a conventional plant With the discharge line isolated, the radioactive gases are left in the CFDS to decay until the level is below the limit for release through the RBVS plant exhaust stack Exemption from 10 CFR 50.34(f)(2)(xiv)(E) as applied to the CES is described in Part 7, Chapter 13
- The CES and CFDS conforms with the guidance of DSRS 9.3.6 March 20, 2019 Chapter 9, Auxiliary Systems 29
Section 9.4: Heating, Ventilation and Air Conditioning Systems Subsections:
9.4.1 - Control Room Area Ventilation System 9.4.2 - Reactor Building and Spent Fuel Pool Area Ventilation System 9.4.3 - Radioactive Waste Building Ventilation System 9.4.4 - Turbine Building Ventilation System March 20, 2019 Chapter 9, Auxiliary Systems 30
9.4.1- Control Room Area Ventilation System (1/1)
Area of Review
- Staff reviewed the CRVS design to determine whether it is capable of maintaining suitable environmental conditions for personnel under normal conditions.
Staff Findings
- NuScale does not credit operator actions to mitigate DBEs.
- CRVS is not credited to be operational during design-basis accidents.
Staff Conclusion
- Staff concluded that the CRVS complies with GDC 2, 4, 5, 19 and 60 as well as with 10 CFR 20.1406, 10 CFR 50.63 and 10 CFR 52.47(b)(1).
March 20, 2019 Chapter 9, Auxiliary Systems 31
9.4.2 - Reactor Building and Spent Fuel Pool Area Ventilation System (1/1)
Area of Review
- The capability of maintaining acceptable ambient conditions in the Reactor Building to support personnel and equipment.
- The ability of controlling airborne radioactivity in the area during normal operation and fuel handling accident.
Staff Findings
- RBVS spent fuel pool exhaust ductwork provides a protective function for the safety-related, Seismic Category I Reactor Building
- NuScale did not take credit to meet offsite dose limit due to leakage from the Reactor Building in case of a HELB event.
Staff Conclusion Staff concluded that the RBVS complies with GDC 2, 5, 60, 61 and 64 as well as with 10 CFR 20.1406 and 10 CFR 52.47(b)(1).
March 20, 2019 Chapter 9, Auxiliary Systems 32
9.4.3 - Radioactive Waste Building Ventilation System (1/1)
Area of Review
- The capability to support personnel access and equipment functions by maintaining a suitable operating environment in the RWB.
- The ability to support the control of radioactive contamination by maintaining airflow from areas of lesser potential contamination to areas of greater potential contamination, maintaining the RWB at a negative pressure with respect to the outside atmosphere, and collecting potentially contaminated discharges vented from equipment in the RWB.
Staff Findings
- Staff finds that the RWBVS conforms with the guidance of SRP 9.4.3.
Staff Conclusion Staff concluded that the RBVS complies with GDC 2, 5, and 60 as well as with 10 CFR 20.1406 and10 CFR 52.47(b)(1).
March 20, 2019 Chapter 9, Auxiliary Systems 33
9.4.4 - Turbine Building Ventilation System (TBVS) (1/1)
Area of Review
- Capability to maintain a suitable environment for all equipment and personnel in the TGB during startup, shutdown, and normal plant operation.
Staff Findings
- Staff finds that the TBVS conforms with the guidance of SRP 9.4.4.
Staff Conclusion
- Staff concluded that the TBVS complies with GDC 2, 5, and 60 as well as with 10 CFR 52.47(b)(1).
March 20, 2019 Chapter 9, Auxiliary Systems 34
Section 9.5: Other Auxiliary Systems Subsections:
9.5.1 - Fire Protection Program 9.5.2 - Communication Systems 9.5.3 - Lighting Systems 9.5.4 - Emergency Diesel Engine (Backup Diesel Generators) Support Systems March 20, 2019 Chapter 9, Auxiliary Systems 35
9.5.1 - Fire Protection Program (1/1)
Review Objectives
- Staff reviewed the Fire Protection Program and Fire Hazard Analysis in accordance with the applicable guidance in SRP 9.5.1.1, SECY 90-016, Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements. and Regulatory Guide 1.189, Fire Protection for Nuclear Power Plants.
- Staff reviewed to ensure compliance with 10CFR 50.48, Fire Protection, GDC 3, 5, 19, and 23.
- RAIs issued, all closed/resolved. No open items.
Staffs Findings
- The Fire Protection Program and Fire Hazard Analysis are in accordance with the applicable guidance in SRP 9.5.1.1, SECY 90-016, Regulatory Guide 1.189, and are in compliance with 10CFR 50.48, GDC 3, 5, 19, and 23.
March 20, 2019 Chapter 9, Auxiliary Systems 36
9.5.2 - Communication Systems (1/1)
- The staff finds that the COMS designs are acceptable and meet the applicable requirements of 10 CFR Part 50, and 10 CFR Part 52.
- No open items.
- The NRC staff is present to address any questions or comments from the Committee members.
March 20, 2019 Chapter 9, Auxiliary Systems 37
9.5.3 - Lighting Systems (1/2)
NuScale Design:
The plant lighting system includes normal plant lighting, emergency plant lighting, and normal and emergency main control room lighting.
Normal plant lighting provides illumination on the plant site and for plant buildings Emergency plant lighting provides illumination outside the control room upon loss of normal lighting Normal and Emergency Main Control Room lighting provides illumination under all operating, maintenance, testing, and emergency conditions Staff Review:
Staff reviewed the capability of the plant lighting levels to provide adequate illumination in all plant areas during all plant operating conditions, and to operate without adversely impacting the operation, control, and maintenance of structure, system and components.
March 20, 2019 Chapter 9, Auxiliary Systems 38
9.5.3 - Lighting Systems (2/2)
Open Item:
The open item in this section is related to RG 1.75 and the physical separation between nonsafety-related lighting circuits and safety-related circuits, and GDC 17 and 18 exemptions.
The completion of the staffs review of plant lighting system is awaiting the completion of Chapter 8 open item, regarding the GDC 17 and 18 exemptions.
Confirmatory Item:
The confirmatory item will be addressed in our Phase 4 SER.
March 20, 2019 Chapter 9, Auxiliary Systems 39
9.5.4 - Emergency Diesel Engine (Backup Diesel Generators) Support Systems (1/1)
- The applicant did not provide any physical or functional description for these systems in FSAR Section 9.5.
- In response to the staffs RAI, the applicant stated that the backup diesel generators (BDGs) are described in FSAR Section 8.3.1.1.2 as stand-alone, skid-based installations including support features such as lubrication and cooling. These support features are not assigned to discrete systems similar to those used for traditional large LWR safety-related diesel applications as presupposed by the guidance in SRP 9.5.4 through 9.5.8. Hence no need to include these SRP sections in the FSAR.
- The staff agrees with the applicant determination that the EDESS are nonsafety, non-risk-significant.
- No open items.
- The NRC staff is present to address any questions or comments from the Committee members.
March 20, 2019 Chapter 9, Auxiliary Systems 40
ACRONYMS BDG - backup diesel generator SRP - standard review plan RCCWS - reactor component cooling water SSC - structure, system and component system SSE - safe shutdown earthquake CDI - conceptual design information ITAAC - inspections, tests, analyses, and COL - combined license acceptance criteria CSF - condensate storage facility TBVS - turbine building ventilation system CST - condensate storage tank UHS - ultimate heat sink CVCS - chemical and volume control system CWS - chilled water system DBA - design basis accident DCD - design control document EDG - emergency diesel generator EFDS - equipment and floor drainage systems GDC - general design criterion RAI - request for additional information RCS - reactor coolant system SAFDL - specified acceptable fuel design limit SFP - spent fuel pool SFPCS - spent fuel pool cooling system SFPCCS - spent fuel pool cooling and cleanup system March 20, 2019 Chapter 9, Auxiliary Systems 41
LO-0319-64764 March 7, 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 Presentation: NuScale Technical Specifications FSAR Chapter 16 and DCA Part 4, PM-0319-64414, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safegurards (ACRS) NuScale Subcommittee meeting on March 20, 2019. The materials support NuScales presentation of Chapter 16 and Design Certification Application (DCA) Part 4, Technical Specifications. is the nonproprietary version of the presentation entitled ACRS Presentation: NuScale Technical Specifications FSAR Chapter 16 and DCA Part 4, PM-0319-64414, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN, 8H12 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston. NRC, 8H12 Samuel Lee, NRC, 8H12 Getachew Tesfaye, NRC, 8H12 : ACRS Presentation: NuScale Technical Specifications FSAR Chapter 16 and DCA Part 4, PM-0319-64414, 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-0319-64764 :
ACRS Presentation: NuScale Technical Specifications FSAR Chapter 16 and DCA Part 4, PM-0319-64414, 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 Presentation:
NuScale Technical Specifications FSAR Chapter 16 and DCA Part 4 Ross Snuggerud Plant Operations Supervisor Carrie Fosaaen Licensing Supervisor Karl Gross Licensing Lead March 20, 2019 1
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Chapter 16: Technical Specification
- Different TS
- Simpler natural circulation and single vessel design
- Passive ECCS with passive support systems
- Digital module protection system
- No safety-related electrical or offsite power needed
- No credited HVAC, no active support systems
- Reactor relocated for refueling activities
- Smaller than existing TS
- Approx. 150 pages of Technical Specifications
- only about 90 pages of LCOs
- (AP1000 approx. 350 pages)
- Approx. 275 pages of Bases
- (AP1000 approx. 670 pages) 2 PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Content Selection
- Applied 10 CFR 50.36 criteria to NuScale design, operations, and safety analyses
- Technical report describing process and results
- Includes LCO Criterion 4 content
- Manual Actuation Function
- Remote Shutdown Station
- Developed led by the Operating organization
- Incorporate industry and staff operating experience
- Adoption of Industry Practices as Appropriate
- Used STS Writers Guide format and guidance
- Monitored NRC/Industry working group on Tech Spec content
- Considered, adapted, and incorporated concepts as appropriate 3
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
NuScale MODE Definitions Reactivity Indicated Reactor Coolant MODE Title Condition (keff) Temperatures (°F) 1 Operations 0.99 All 420 2 Hot Shutdown < 0.99 Any 420 3 Safe Shutdown(a) < 0.99 All < 420 4 Transition(b)(c) < 0.95 N/A 5 Refueling(d) N/A N/A (a) Any CRA capable of withdrawal, any CVCS or CFDS connection to the module not isolated.
(b) All CRAs incapable of withdrawal, CVCS and CFDS connections to the module isolated, and all reactor vent valves electrically isolated.
(c) All reactor vessel flange bolts fully tensioned.
(d) One or more reactor vessel flange bolts less than fully tensioned.
4 PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Limiting Conditions for Operation 3.3 Instrumentation 3.3.1 Module Protection System Instrumentation 3.3.2 Reactor Trip System Logic and Actuation 3.3.3 Engineered Safety Features Actuation System Logic and Actuation 3.3.4 Manual Actuation Functions 3.3.5 Remote Shutdown Station 5
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Limiting Conditions for Operation 3.5 Passive Core Cooling Systems 3.5.1 Emergency Core Cooling System 3.5.2 Decay Heat Removal System 3.5.3 Ultimate Heat Sink 3.6 Containment Systems 3.6.1 Containment 3.6.2 Containment Isolation Valves 3.7 Plant Systems 3.7.1 Main Steam Isolation Valves 3.7.2 Feedwater Isolation 3.7.3 In-Containment Secondary Piping Leakage 3.8 Refueling Operations 3.8.1 Nuclear Instrumentation 3.8.2 Decay Time 6
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Technical Specification RAI Received and Addressed
- 60 Chapter 16 RAI Received
- Many included multiple subparts,
- e.g., 16-60 handled as 81 subparts.
- 59 have been responded to
- 16-59, and 2 of 81 subparts of RAI 16-60 pending
- 13 supplemental responses submitted 7
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Key Technical Specification Questions
- STS Content Evolution Based on Operating Plants
- Instrumentation Surveillance Testing
- Response Time Testing
- RCS Leak Rate Monitoring
- RCS Specific Activity Limit
- In-Containment Secondary Piping Leakage 8
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Conclusion NuScale has developed design-specific Technical Specifications and Bases that
- Reflect NuScale design, analyses, and operations
- Ensure safety and licensing bases will be implemented during operations
- Align with industry standard technical specifications format and content
- Includes consideration of industry experience
PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Acronyms CFDS Core Flood and Drain System COLA Combined Operating License Applicant CRA Control Rod Assembly CVCS Chemical and Volume Control System ECCS Emergency Core Cooling System ESFAS Engineered Safety Features Actuation System FW Feedwater (as in FW Isolation Valve)
HVAC Heating, Ventilation, and Air Conditioning ICIS In-Core Instrument System LCO Limiting Condition for Operations MCS Module Control System MPS Module Protection System MSIV Main Steam Isolation Valve PWR Pressurized Water Reactor RCS Reactor Coolant System RTS Reactor Trip System SFCP Surveillance Frequency Control Program SR Surveillance Requirement SSC Structures, Systems, and Components STS Standard Technical Specifications (NUREGs as published by NRC)
TS Technical Specifications 10 PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 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 11 PM-0319-64414 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presentation to the ACRS Subcommittee NuScale Power, LLC NuScale Design Certification Application Review Safety Evaluation with Open Items: Chapter 16 GENERIC TECHNICAL SPECIFICATIONS AND BASES MARCH 20, 2019
- Technical Staff Presenter Craig Harbuck
- Project Managers Gregory Cranston - Lead Project Manager Getachew Tesfaye - Chapter 16 Project Manager March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 2
Staff Review Team Primary Reviewers Technical Specifications Branch*........................ Craig Harbuck, Bob Tjader
- Secondary Reviewers Containment and Ventilation Branch.............................................................
.Nan Chien, Syed Haider, Boyce Travis Mechanical Engineering Branch .......................................................Tuan Le Radiation Protection and Accident Consequences Branch
...................................................... Lavera, Ronald, Michelle Hart Reactor Systems, Nuclear Performance, and Code Review Branch ..........
......................................... Matt Thomas, Timothy Drzewiecki, Alexandra Burja Instrumentation, Controls, and Electrical Engineering Branch (ICE) ..........
............................... Dawnmathews Kalathiveettil, Dinesh Taneja, Joe Ashcraft Plant Systems Branch ............................................ Angelo Stubbs, Chang Li Probabilistic Risk Assessment Branch .... Anne-Marie Grady, Marie Pohida Electrical Engineering Branch**............................................... Fanta Sacko Materials & Chemical Engineering Branch ........... Leslie Terry, Greg Makar
- Center of expertise NRR/DSS/STSB ** Center of expertise NRR/DE/EENB March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 3
- Outline Overview of Chapter 16 Technical Topics
- Defined Terms
- Electrical Power Surveillance Requirements
- Allowed Leakage for Main Steam and Feedwater Isolation Valves
- Exceptions to Meeting Surveillances
- Open Item Listing
- Pending MPS design changes
- General Issues Review Status Summary March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 4
Overview of Design Certification Application (DCA)
Part 2, Chapter 16; and Part 4 1.0 USE AND APPLICATION 2.0 SAFETY LIMITS (SLs) 1.1 Definitions 2.1 SLs 1.2 Logical Connectors 2.1.1 Reactor Core SLs 1.3 Completion Times 2.1.2 RCS Pressure SL 1.4 Frequency 2.2 SL Violations 3.0 LCO APPLICABILITY 3.0 LCO AND SR (#) - number of LCO subsections LCO 3.0.1 3.1 REACTIVITY CONTROL SYSTEMS (9)
LCO 3.0.2 3.2 CORE OPERATING LIMITS (2)
LCO 3.0.3 3.3 INSTRUMENTATION (5)
LCO 3.0.4 3.4 REACTOR COOLANT SYSTEM (RCS) (10)
LCO 3.0.5 3.5 PASSIVE CORE COOLING SYSTEMS (PCCS) (3)
LCO 3.0.6 3.6 CONTAINMENT SYSTEMS (2)
LCO 3.0.7 3.7 PLANT SYSTEMS (3)
LCO 3.0.8 3.8 REFUELING OPERATIONS (2) 3.0 SR APPLICABILITY 4.0 DESIGN FEATURES (3 sections)
SR 3.0.1 SR 3.0.2 5.0 ADMINISTRATIVE CONTROLS (7 sections)
SR 3.0.3 SR 3.0.4 March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 5
Overview of DCA Part 4 LCO selection criteria results Criterion 1 Criterion 3 3.4.7, RCS Leakage Detection Instrumentation 3.1.9 Boron Dilution Control (Chemical and Volume Control System (CVCS)) demineralized water isolation valves)
Criterion 2 3.3.1 Module Protection System Instrumentation 3.3.2 Reactor Trip System (RTS) Logic and Actuation 3.1.1 SHUTDOWN MARGIN (SDM) 3.3.3 Engineered Safety Features Actuation System (ESFAS) Logic 3.1.2 Core Reactivity and Actuation 3.1.3 Moderator Temperature Coefficient (MTC) 3.4.4 Reactor Safety Valves (RSVs) 3.1.4 Rod Group Alignment Limits 3.4.6 CVCS Isolation Valves 3.1.5 Shutdown Bank Insertion Limits 3.4.10 Low Temperature Overpressure Protection (LTOP) Valves 3.1.6 Regulating Bank Insertion Limits 3.5.1 Emergency Core Cooling System (ECCS) 3.1.7 Rod Position Indication 3.5.2 Decay Heat Removal System (DHRS) 3.1.9 Boron Dilution Control (Boric Acid supply 3.5.3 Ultimate Heat Sink boron concentration; demineralized water 3.6.1 Containment makeup flow rate) 3.6.2 Containment Isolation Valves 3.2.1 Enthalpy Rise Hot Channel Factor (FH) 3.7.1 Main Steam Isolation Valves (MSIVs) 3.2.2 AXIAL OFFSET (AO) 3.7.2 Feedwater Isolation 3.4.1 RCS Pressure, Temperature, and Flow 3.8.1 Nuclear Instrumentation (refueling neutron flux channels)
Resistance Critical Heat Flux Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits Criterion 4 3.4.5 RCS Operational LEAKAGE 3.3.4 Manual Actuation Functions 3.4.8 RCS Specific Activity 3.3.5 Remote Shutdown Station (RSS) 3.4.9 Steam Generator (SG) Tube Integrity 3.5.3 Ultimate Heat Sink No Criterion 3.7.3 In-Containment Secondary Piping Leakage 3.8.2 Decay Time 3.1.8 PHYSICS TESTS Exceptions March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 6
Overview of DCA Part 4 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 4.1.1 Site and Exclusion Boundaries 4.2.1 Fuel Assemblies 4.3.1 Criticality 4.1.2 Low Population Zone (LPZ) 4.2.2 Control Rod Assemblies 4.3.2 Drainage 4.3.3 Capacity 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.5 Programs and Manuals 5.1.1 [plant manager] 5.5.1 Offsite Dose Calculation Manual 5.1.2 [shift manager] 5.5.2 Radioactive Effluent Control Program 5.2 Organization 5.5.3 Component Cyclic or Transient Limit 5.2.1 Onsite and Offsite Organizations 5.5.4 Steam Generator Program 5.2.2 Facility Staff 5.5.5 Secondary Water Chemistry Program 5.3 Facility Staff Qualifications 5.5.6 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.3.1 facility staff 5.5.7 TS Bases Control Program 5.3.2 licensed SRO and licensed RO 5.5.8 Safety Function Determination Program 5.4 Procedures 5.5.9 Containment Leakage Rate Testing Program 5.4.1 Written procedures shall be 5.5.10 Setpoint Program (SP)
- a. Regulatory Guide 1.33 5.5.11 Surveillance Frequency Control Program (FSAR Table 16.1-1)
- b. Emergency Operating Procedures 5.5.12 Spent Fuel Storage Rack Neutron Absorber Monitoring Program
- c. QA and Environmental Monitoring 5.6 Reports
- e. Section 5.5 programs 5.6.2 Radiological Effluent Release Report
- f. availability and reliability 5.6.3 Core Operating Limits Report (COLR) controls for SSCs in 5.6.4 RCS Pressure and Temperature Limits Report (PTLR) owner-controlled requirements 5.6.5 Steam Generator Tube Inspection Report manual 5.7 High Radiation Area 5.7.1 1 rem per hour at 30 cm 5.7.2 > 1 rem per hour at 30 cm, but 500 rads per hour at 100 cm March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 7
Technical Topics Defined Terms
- LEAKAGE
- Mark up of STS definition of Pressure Boundary LEAKAGE:
LEAKAGE (except primary to secondary SG LEAKAGE) through a nonisolable fault in a RCS component body, pipe wall, or vessel wall. A fault in an RCS component body, pipe wall, or vessel wall is isolated if LEAKAGE through the isolation device is 0.5 gpm per nominal inch of valve size up to a maximum limit of 5 gpm.
- Based on traveler TSTF-534, "Clarify Application of Pressure Boundary Leakage Definition" Revision 0, rejected by the staff in 2012
- Unclear where in NuScale reactor coolant pressure boundary this provision could be used, much less would be beneficial to unit operation or a safety enhancement
- Open Item under RAI 9031, Question 16.2, Subquestion f (Supplemental response pending)
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 8
Technical Topics Defined Terms (continued)
- OPERABLE-OPERABILITY
- Mark up of STS definition plus staff-suggested changes for consistency with NuScale instrumentation design:
A system, subsystem, separation group, channel, division, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, separation group, channel, division, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
- Open Item under RAI 9031, Question 16.2, Subquestion f (Supplemental response pending)
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 9
Technical Topics Defined Terms (continued)
- The applicant has proposed to omit these defined terms, and just use response time in the corresponding surveillances. Digital response time assumes an allocated value.
- Bases for each surveillance explains the response time:
3.3.1 MPS instrumentation Functions - channel actuation response time; B 3.3.1:
the time from when the process variable exceeds its setpoint until the output from the channel analog logic reaches the input of the MPS digital logic.
Response time may be verified by any series of sequential, overlapping or total channel measurements, including allocated sensor response time, such that the response time is verified.
3.3.2 RTS Logic and Actuation - RTS division response time; B 3.3.2:
The maximum digital time response is described in the FSAR. This SR encompasses the response time of the RTS division from the output of the equipment interface modules (EIM) until the RTBs are open.
Total response time may be verified by any series of sequential, overlapping, or total channel measurements.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 10
Technical Topics Defined Terms (continued)
- RTS RESPONSE TIME and ESF RESPONSE TIME (continued) 3.3.3 ESFAS Logic and Actuation - only addresses response time for opening of pressurizer heater breakers; B 3.3.3:
The maximum digital time response is described in the FSAR. This SR encompasses the response time of the ESFAS from the output of the EIM to the loss of voltage at the output of the pressurizer heater breaker.
Total response time may be verified by any series of sequential, overlapping, or total channel measurements.
The response time of valves actuated by the ESFAS are verified in accordance with the IST program (and associated SRs in Subsections 3.4.6, 3.4.10, 3.5.1, 3.5.2, 3.6.2, 3.7.1, 3.7.2; Bases do not include language about response time, as provided in Section 3.3 Bases - subject of RAI 9642, Question 16-65)
Example: SR 3.6.2.3 - verify isolation time of each automatic containment isolation valve in accordance with the INSERVICE TESTING PROGRAM; B 3.6.2:
Isolation time is measured from output of the MPS EIM until the valves are isolated.
- Open Item under RAI 9614, Question 16-50 (ICE evaluating response)
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 11
Technical Topics RTS and ESFAS Surveillance Requirements
- MPS digital logic channels, RTS digital actuation logic divisions, and ESFAS digital actuation logic divisions rely on self-testing features to verify channel and division OPERABILITY between CHANNEL CALIBRATIONS and ACTUATION LOGIC TESTS (24 month frequencies)
CHANNEL OPERATIONAL TEST (92 day frequency) retained only for RCS leak detection monitors implemented in Module Control System (containment evacuation system (CES) gaseous radioactivity monitor channel, and CES condensate channels.)
ACTUATION LOGIC TEST verifies (B 3.3.2/B 3.3.3) actuation and priority logic (APL) on all RTS/ESFAS EIMs, the enable nonsafety control switches, the main control room isolation switches, the [ESFAS] override switches, and the operating bypass switches.
The ACTUATION LOGIC TEST includes a review of any alarms or failures reported by the self-testing features.
CHANNEL CHECK definition is revised as shown in markup of STS definition:
A CHANNEL CHECK shall be the qualitative assessment, by observation, of verification through the absence of alarms from the automatic analog and binary process signal monitoring features used to monitor channel behavior during operation. Deviation beyond the established acceptance criteria is alarmed to allow appropriate action to be taken. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. This determination can be made using computer software or be performed manually.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 12
Technical Topics Electrical Power Surveillance Requirements
- No LCOs included for electrical power systems, which are all nonsafety-related
- Module Protection System is protected from faults in the nonsafety-related electrical power system by use of Class 1E isolation devices
- Open on undervoltage or overcurrent
- Details of design of Class 1E isolation devices not included in DCA Part 2, Chapters 7 and 8
- Isolation settings are determined in accordance with the Setpoint Program
- OPERABILITY verified by CHANNEL CALIBRATION surveillance requirements in Subsections 3.3.1, MPS Instrumentation; 3.3.2, RTS Logic and Actuation; 3.3.3, ESFAS Logic and Actuation support OPERABILITY of MPS Instrumentation Functions, RTS Logic and Actuation Functions, and ESFAS Logic and Actuation Functions Open Item under RAI 9051, Question 16-28, regarding whether Class 1E isolation device CHANNEL CALIBRATION surveillance is needed for Manual RTS and ESFAS Functions of Subsection 3.3.4 March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 13
Technical Topics Allowed Leakage for Main Steam and Feedwater Isolation Valves
- Verifying secondary system isolation valve leakage is within limits is required by SR 3.7.1.3 (main steam isolation valves (MSIVs)) and SR 3.7.2.3 (feedwater isolation valves (FWIVs), feedwater regulating valves (FWRVs), and feedwater check valves (FWCVs) in accordance with the Inservice Testing Program.
Leakage limit ensures decay heat removal system (DHRS) maintains sufficient inventory to remain OPERABLE following events that credit DHRS actuation.
Applicants response to RAI 9614, Question 16-58, says that Surveillance acceptance criteria values for main steam and feedwater isolation valve leakage would be provided by a COL applicant to complete COL Item 3.9-5, by establishing an Inservice Testing program in accordance with ASME OM Code and 10 CFR 50.55a.
- Open Item under RAI 9614, Question 16-58, to track completion of the staffs ongoing evaluation of the response.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 14
Technical Topics Exceptions to Meeting Surveillances
- While the unit is operating under an exception, operational impact may be reduced by allowing repositioning of the valve, breaker, or device under administrative control.
RTBs:
Not required to be met for RTBs that are open.
Class 1E Electrical Isolation Devices:
Not required to be met for Class 1E isolation devices that have isolated 1E circuits from non-1E power.
Pressurizer Heater Breakers:
Not required to be met for pressurizer heater breakers that are open or closed under administrative control.
CVCS Isolation Valves Not required to be met for valves that are closed or open under administrative controls.
Verify each automatic CVCS valve actuates to the isolation position on an actuation signal except for valves that are open under administrative controls.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 15
Technical Topics Exceptions to Meeting Surveillances (continued)
Reactor Vent Valves and Recirculation Valves:
Not required to be met for valves that are open.
Containment Isolation Valves, Verify the isolation time is within limits except for valves that are open under administrative controls.
Verify each automatic containment isolation valve actuates to the isolation position on an actuation signal except for valves that are open under administrative controls.
Revision 2 of DCA Part 4 does not include such exceptions for DHRS Actuation Valves MSIVs, MSIV Bypass Valves FWIVs, FWRVs
- Open Item under RAI 9051, Question 16-28, because there are no apparent scenarios in which administrative control of an automatic valve, breaker, or device provides a safety benefit over entering the LCO action requirements. SR 3.0.1 already states that surveillances do not have to be performed on inoperable equipment.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 16
Technical Topics Open Item Listing (1 of 6)
RAI # Action Issue 9033 16-2.f NuScale Definition of RCS Pressure Boundary LEAKAGE Additional information about locations in FSAR that describe contents of Owner Controlled 9051 16-27 NuScale Requirements Manual (Note that creation of this document is the subject of COL Item 16.1-2)
Append the phrase in accordance with the 9051 16-28 NuScale Setpoint Program to Channel Calibration Surveillance statement Include a Channel Calibration Surveillance for 9051 16-28 NuScale Class 1E electrical isolation devices in LCO 3.3.4, Manual RTS and ESF Functions 9051 16-28 Exceptions to meeting Surveillances and use of NuScale 9642 16-64 administrative controls while using an exception Update FSAR Table 16.1-1, Surveillance 9034 16-30.d NuScale Frequency Control Program Base Frequencies March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 17
Technical Topics Open Item Listing (2 of 6)
RAI # Action Issue Omission of the Response Time definitions and 9614 16-50 Staff defined terms, and the adequacy of the proposed 9642 16-65 NuScale response time verification Surveillances (editorial) SR 3.7.2.1 should not contain required because both FWIVs are required OPERABLE by 9614 16-54 NuScale LCO 3.7.2, and the SR only applies to valves with pressurized gas accumulators (editorial) Phrasing of surveillance statements to 9614 16-54 NuScale verify accumulator pressure for valves with the NuScale-specific actuator design Defer providing values for INSERVICE TESTING PROGRAM Surveillance acceptance criteria for 9614 16-58 Staff main steam and feedwater isolation valve leakage limits March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 18
Technical Topics Open Item Listing (3 of 6)
RAI # Action Issue Applicability of LCO 3.3.1, MPS Function 22b, 9614 16-59 NuScale DHRS actuation on High Narrow Range Containment Pressure in MODE 3 Editorial comments on SR 3.1.7.1, SR 3.1.9.1, 9634 16-60 NuScale SR 3.2.1.1, SR 3.3.2.3, SR 3.3.3.3, SR 3.4.10.4, and SR 3.5.1.4.
Clarify whether Low Pressurizer Level or Low 9634 16-60.2 NuScale Low Pressurizer Level initiates secondary system isolation (SSI) actuation.
(editorial) Conformance to Improved TS writers guide section 4.1.3.b. Remove unnecessary 9614 16-52 NuScale instances of the words inoperable and required 9634 16-60.34
& Staff in Condition, Required Action, and Surveillance 9634 16-60.63 statements; and also Action Notes on separate Condition entry.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 19
Technical Topics Open Item Listing (4 of 6)
RAI # Action Issue LCO 3.1.9, Boron Dilution Control, Action B, requires isolating the dilution source, which completes the safety function of the DWSI makeup isolation valves.
9634 16-60.37.1 NuScale It appears that the Applicability of LCO 3.1.9 for these valves would be more accurate by stating: MODES 1, 2, and 3 with any CVCS demineralized water isolation valve open.
Correct reference to interlock (either N-2H or N-2L) in LCO 3.3.1, Required Action E.1 (Reduce THERMAL POWER to below the N-2L 9634 16-60.37.3 NuScale interlock. l 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). The affected MPS Functions (2a, 18a, 18b, 18c) are applicable in MODE 1 above the N-2H interlock (15% RTP)
Clarify rationale for Applicability not including MODE 3 for MPS Low Low Pressurizer Pressure Functions 9.b, 9634 16-60.70 NuScale DHRS, 9.c, CVCSI, and 9.d, PHT, of LCO 3.3.1, between 350°F and 420°F.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 20
Technical Topics Open Item Listing (5 of 6)
RAI # Action Issue What are the appropriate means of addressing prevention and mitigation of the postulated inadvertent actuation of the containment flood and drain system (CFDS) to flood the 9642 16-61 NuScale containment vessel with RCS temperature above the RCS temperature limit in the PTLR? Why not use safety-grade interlock on CFDS CIVs to protect RPV from higher than analyzed thermal stress?
Clarify in the Subsection 3.3.1 Bases the relationship of the MPS Instrumentation Functions, and their bypassing or 9642 16-62 NuScale enabling interlocks and permissives, to the SP controls and Channel Calibration Surveillances.
(editorial) Clarify Bases for LCO 3.0.4 and SR 3.0.4. Should 9642 16-63 NuScale initiating and terminating PASSIVELY COOLED be treated as a MODE transition within MODE 3?
Request for adding two additional references to the 9642 16-66 NuScale Specification 5.6.3 COLR methodology list, LOCA analyses for LCO 3.1.6, LCO 3.2.2, and LCO 3.4.1.
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 21
Technical Topics Open Item Listing (6 of 6)
RAI # Action Issue Acceptability of 1.5 gallons per hour leak-before-break (LBB) limit on in-containment secondary system piping leakage, specified by LCO 3.7.3.
9201 5.2.5-7 Staff The staff is reviewing March 5, 2019, supplemental response and may request additional edits to Subsection B 3.7.3.
Revised LBB analysis-related additional information in 9201 5.2.5-7 NuScale preparation.
Completion of COL sub-item listing under None NuScale COL Item 16.1-1 March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 22
Technical Topics Pending MPS design change
- 13 DHRS-initiating MPS instrumentation Functions to support new ESFAS Function of Secondary System Isolation (SSI)
SSI closes isolation valves in main steam lines and feedwater lines without an unnecessary cooldown from DHRS actuation Associated SSI operating bypass based on FWIV position indication, which is upgraded to safety-related to support bypass logic Just 4 of these 13 MPS instrumentation Functions will initiate DHRS:
- High Pressurizer Pressure (7.b)
- High Steam Pressure (17.b)
- High Narrow Range RCS Hot Temperature (13.b)
- Low AC Voltage to ELVS Battery Chargers (25.b)
- Remove uncredited ECCS actuation on RPV riser level; The safety analysis credited ECCS actuation on High Containment Water Level (23.a) is adequate to initiate ECCS by opening RVVs and RRVs March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 23
Technical Topics General Issues
- Combined License (COL) sub-item determination -
Open Item
- Disposition of NRC-approved technical specifications task force (TSTF) traveler changes - adapted or not adapted which have been incorporated in NUREG-1432 (digital), Revision 4; or approved since issuance of NUREG-1432 (digital), Revision 4.
Open Items under RAI Questions 16-2f and 16-28
- Administrative Changes -
Open Items as listed above Correction of grammatical and typographical errors Replacement of inapplicable content taken from STS Bases Addition of missing content to the Bases Clarification of submitted content in the Bases Conformance to STS style, punctuation, phrasing, formatting conventions Resolution of inconsistencies, both within DCA Part 4 and DCA Part 2 March 21-22, 2017 Chapter 16 Generic Technical Specifications and Bases 24
Chapter 16 Review Status
SUMMARY
- The NuScale Generic Technical Specifications are based upon Standard Technical Specifications; differences are a result of design differences with plant designs considered in the STS
- A thorough review of the NuScale GTS has been conducted resulting in a safety evaluation chapter that includes open items in the following areas:
- New and Revised Defined Terms and revised definitions
- MPS instrumentation Function Applicability inconsistencies
- I&C surveillance requirements & testing
- Response time testing
- Administrative and editorial issues
- Resolution of the Open Items will be accomplished with the assistance of the technical branches (i.e., ICE, SRSB, SPRA, SCVB, and SPSB)
March 20, 2019 Chapter 16 Generic Technical Specifications and Bases 25