ML24303A168
| ML24303A168 | |
| Person / Time | |
|---|---|
| Issue date: | 10/01/2024 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-0058 | |
| Download: ML24303A168 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards NuScale Design-Centered Review Subcomm.
Docket Number:
(n/a)
Location:
teleconference Date:
Tuesday, October 1, 2024 Work Order No.:
NRC-0058 Pages 1-60 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
(ACRS) 5
+ + + + +
6 NUSCALE DESIGN-CENTERED SUBCOMMITTEE 7
+ + + + +
8 TUESDAY, OCTOBER 1, 2024 9
+ + + + +
10 The Subcommittee met via Teleconference, 11 at 1:00 p.m. EDT, Vesna B. Dimitrijevic, Chair, 12 presiding.
13 COMMITTEE MEMBERS:
14 VESNA B. DIMITRIJEVIC, Chair 15 RONALD G. BALLINGER, Member 16 VICKI M. BIER, Member 17 CRAIG A. HARRINGTON, Member 18 GREGORY H. HALNON, Member 19 WALTER L. KIRCHNER, Member 20 ROBERT P. MARTIN, Member 21 SCOTT P. PALMTAG, Member 22 DAVID A. PETTI, Member 23 THOMAS E. ROBERTS, Member 24 MATTHEW W. SUNSERI, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 ACRS CONSULTANT:
1 DENNIS BLEY 2
3 DESIGNATED FEDERAL OFFICIAL:
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 C-O-N-T-E-N-T-S 1
Opening Remarks and Introduction 4
2 Discussion of NuScale's Topical Report on 3
Capabilities to Mitigate Beyond-4 Design-Basis Events 9
5 Staff's Evaluation of NuScale's Topical 6
Report on Capabilities to Mitigate 7
Beyond-Design-Basis Events 28 8
Opportunity for Public Comment 59 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 P-R-O-C-E-E-D-I-N-G-S 1
1:01 p.m.
2 CHAIR DIMITRIJEVIC: Okay. So the meeting 3
will now come to order. This is a meeting of the 4
NuScale Design-Centered Review Subcommittee of the 5
Advisory Committee on Reactor Safeguards. I'm Vesna 6
Dimitrijevic, chair of today's subcommittee meeting.
7 ACRS members in attendance in person are 8
Walt Kirchner, Ron Ballinger, Vicki Bier, Gregory 9
Halnon, Craig Harrington, Robert Martin, Scott 10 Palmtag, Dave Petti, and Thomas Roberts. ACRS members 11 in attendance virtually via Teams are Matt Sunseri and 12 myself. I did not see Matt yet on the line, but I 13 think he should be here with us any moment.
14 So we have also one of our consultants 15 participating virtually via Teams Dennis Bley.
16 Dennis, I see on the line. If I have missed anybody, 17 the ACRS member consultants, please speak up now. But 18 I think I listed everybody, so I couldn't have missed 19 anybody.
20 So Mike Snodderly of ACRS staff is the 21 designated federal officer for this meeting. No 22 member conflict of interest were identified for 23 today's meeting. And we have a quorum for today's 24 meeting.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 During today's meeting, the subcommittee 1
will receiving a briefing of the staff advanced safety 2
relation of the NuScale topical report, NuScale power 3
plant design capability to mitigate beyond design 4
basis events, DBE, defined by 10 CFR 51.55. The 5
topical report describes design capabilities to 6
mitigate beyond design basis events. Design features 7
that provide enhanced capability for coping with an 8
extended loss of electrical power, loss of normal 9
access to the normal heat sink, and loss of launch 10 area due to explosions or fire are discussed within 11 this topical report.
12 10 CFR 51.55, mitigation of beyond design 13 basis events, was put in effect in response to lessons 14 learned from the Fukushima Daiichi accident. The ACRS 15 was established by statute and is governed by the 16 Federal Advisory Committee Act or FACA. The NRC 17 implements FACA in accordance with our regulation.
18 But those regulations are the committee 19 bylaws. The ACRS speaks only through each published 20 letter reports. All members' comments should be 21 regarded as only the individual opinion of that member 22 and not as a committee position.
23 All of the information related to ARCS 24 activities such as
- letters, rules, committee 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
6 participation, and transcripts are located on the NRC 1
public website and can be easily found by tying About 2
Us ACRS in the search field of the NRC home page. The 3
ACRS consisted with agency value of public 4
transparency in regulation of nuclear facilities 5
provides an opportunity for public input and comment 6
during our proceedings. We have received no written 7
statements or requests to make an oral statement from 8
the public.
9 We have also set aside time at the end of 10 the meeting for public comments. Portions of these 11 meetings may be closed to protect sensitive 12 information as required by FACA and the government in 13 the Sunshine Act. Attendance during the closed 14 portion of the meeting will be limited to NRC staff 15 and its consultant, applicants and those individuals 16 in the organization who have entered into an 17 appropriate confidentiality agreement.
18 We will confirm that only eligible 19 individuals in the closed portion of the meeting. The 20 ACRS will gather information, analyze relevant issues 21 and facts, and formulate proposed conclusions and 22 recommendation as appropriate for deliberation by the 23 full committee. A transcript of the meeting is being 24 kept and will be posted on our website.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 When addressing the subcommittee, the 1
participants should first identify themselves and 2
speak with sufficient clarity and volume so that they 3
may be readily heard. If you are not speaking, please 4
mute your computer on Teams by pressing star-6 or by 5
pressing star-6 if you're on the phone. Please do not 6
use the Teams chat feature to conduct sidebar 7
discussions related to presentations. Rather, limit 8
use of the meeting chat function to report IT 9
problems.
10 For everyone in the room, please put --
11 sorry, I have this distraction here -- please put all 12 your electronic devices in silent mode and mute your 13 laptop microphone and speakers. In addition, please 14 keep sidebar discussion in the room to a minimum since 15 the ceiling microphones are live. For the presenters, 16 your table microphones are unidirectional and you will 17 need to speak into the front of the microphone to be 18 heard.
19 Finally, if you have any feedback for the 20 ACRS about today's meeting, we encourage you to fill 21 out the public meeting feedback form on the NRC 22 website. We will now proceed with the meeting. So 23 who will be the first presenter today from the 24 NuScale?
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 MR. CUMMINGS: Hey, this is Kris Cummings 1
from NuScale. Pete will be our presenter. But before 2
that, I'm going to ask Jim Osborn to make opening 3
statements.
4 CHAIR DIMITRIJEVIC: Okay. Thank you.
5 Please.
6 MR. OSBORN: Yes, thank you, Vesna and 7
Kris. Good morning, everyone. My name is Jim Osborn, 8
Licensing Supervisor for NuScale Power. We thank 9
everyone for this opportunity to present the 10 mitigation of beyond design basis event topical 11 report.
12 For a little background, NuScale has 13 developed this topical report to generically address 14 the requirements of 10 CFR 50.155 for a NuScale design 15 plant. It is not specific to either the U.S. 600 or 16 the U.S. 460. This topical report was developed from 17 two technical reports that were found acceptable in 18 the U.S. 600 DCA approved application.
19 However, the DCA review did not assess the 20 capacity and capability of the U.S. 600 plant beyond 21 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This topical report provides a framework 22 for a future licensee to demonstrate that their 23 NuScale plant has the design features and capability 24 to satisfy the requirements of 10 CFR 50.155 for a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 significantly extended period beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. So at 1
this point, I would like to turn the floor over to my 2
colleague, Pete Shaw, who will be the main presenter.
3 MR. SHAW: Good afternoon. My name is 4
Pete Shaw. Thank you for the introduction, Jim. I 5
have 15 years of experience in the nuclear industry, 6
10 of which were at the Vogtle 3 and 4 construction 7
site.
8 I will be the -- I'm the lead presenter on 9
mitigation on beyond design basis events topical 10 report. Next slide, please. Once again, we would 11 like to thank DOE for their award for NuScale and 12 their contributions to the project. Next slide, 13 please. So this is a brief agenda for everything that 14 will be discussing today.
15 It will include an introduction and basis 16 for the topical report. We will have a brief overview 17 of the plant features for individuals to provide 18 familiarity with the design. We'll then discuss the 19 event mitigation features for design basis events 20 which then perform for the accrediting of safety 21 features for mitigation of beyond design basis events.
22 We'll then discuss the conditions of 23 applicability and a summary of the plant systems 24 relied upon for loss of AC power, a summary of how 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
10 regulatory commitments are fulfilled, and then 1
conditions of applicability that were brought by the 2
NRC. And then last, we will have our statements.
3 Next slide, please. So as Jim said, the basis for 4
this particular topical report, we were looking at a 5
consistent approach for mitigation of beyond design 6
basis events.
7 And the importance behind this is that it 8
would be applicable to any applicant looking to adopt 9
a NuScale technology. This includes configurations 10 and that implement the design features that are 11 described. They are described at a level that is 12 consistent for all NuScale power plants.
13 The safety features extend the baseline 14 coping period past traditional reactor technologies.
15 And the topical report discusses how these are relied 16 upon for beyond design basis events. This use 17 informed approached used by NEI guidance and the 18 lessons learned from the Fukushima Daiichi accident, 19 these are developed into the NuScale design and are a 20 part of the basis that we will be discussing today.
21 An applicant will then be required to 22 discuss and describe on how they fulfill these 23 conditions for use in their applications. And as a 24 note, the duration of the extending coping period is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 proprietary NuScale information. If there are 1
questions about the duration beyond the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, we 2
can answer those in the closed session. Next slide, 3
please.
4 So familiarization with the design and 5
what the basis is. There is a NuScale reactor. It's 6
a light water reactor with enhanced passive safety 7
features. And it implements in all of its safety 8
functions passive reliability in a sequence that will 9
be discussed later in more detail.
10 At the initiation of a design basis event, 11 decay heat removal is actuated and it will be removed 12 by the steam generators and the DHRS for three days.
13 And after the three days, the decay heat will continue 14 to be removed by the containment into the ultimate 15 heat sink. This transition period is our topic of 16 discussion today.
17 Obviously, once this is completed, the 18 modules will then transfer into long term air cooling.
19 But again, that's not subject to it. These water 20 levers on this particular graphic are not indicative 21 of final plant design. They're here for illustrative 22 purposes. Next slide, please.
23 So this is an overview of an example, 6 24 pack power plant for the purposes of today's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
12 presentation. What's most important to focus on in 1
this is the control building and reactor building are 2
separate structures. Next slide, please.
3 MEMBER HALNON: Pete, before you go on, 4
this is Greg. Just wanted to get your definition of 5
indefinite on the previous slide. Is that assuming 6
nothing -- no operator reaction to refill or in any 7
way replenish the DHRS?
8 MR. SHAW: So yes, but that's also not 9
covered in this particular topical report. So that's 10 a topic for another discussion in the design of the 11 NuScale power plants. We're focusing just on the 12 boiling and the coping period as discussed in the 13 topical.
14 MEMBER HALNON: Okay. I didn't put 15 indefinite on the slide. You did. Can you explain 16 indefinite to me?
17 MR. SHAW: I'd like for Meghan McCloskey 18 who's on the phone to potentially answer that 19 question.
20 MS. MCCLOSKEY: This is Meghan McCloskey, 21 NuScale. One key clarification here is that the 22 indefinite cooling is really focused on the module 23 conditions that are -- the modules are isolated. The 24 containments are closed. And so water inside the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 modules continues to cool the emergency core cooling 1
system. Eventually, there could potentially come a 2
point where additional liquid is needed in the pool 3
for spent fuel pool cooling. Does that clarify your 4
question?
5 MEMBER HALNON: I think so. I don't want 6
to get into a lot of detail. But I was just trying to 7
figure out at what point something had to get done.
8 You see Dr. Martin is going to jump in there and bail 9
me out or something.
10 MEMBER MARTIN: I don't know about bailing 11 you out. But I had a question for Meghan. So just 12 from what you described, we have oil off. We have 13 closed containment.
We do not expect some 14 condensation and maybe return of that coolant back 15 into the pool?
16 MS. MCCLOSKEY: The condensation and 17 return of coolant potentially occurs. It's not a 18 design feature of the reactor building. And it's not 19 credited in any of the work that we've done to 20 evaluate these types of extended cooling capabilities.
21 And the focus of this topical report really was on the 22 criteria and the design features of the NuScale plant 23 to cope with these beyond design basis conditions.
24 NuScale was not looking for approval of specific 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 methods and requirements for the thermal hydraulic 1
analysis specifically.
2 MEMBER MARTIN:
So potentially 3
conservative to kind of set design criteria.
4 MS. MCCLOSKEY: I didn't hear that 5
clearly.
6 MEMBER MARTIN: I was trying to summarize 7
and paraphrase what you're saying. Potentially 8
conservative basically to establish a method of --
9 criteria.
10 MS. MCCLOSKEY: Yes.
11 MR. SHAW: Thank you, Meghan. Next slide, 12 please. For familiarization, this is a reactor 13 building example. This is typical features of a 6 14 pack power plant.
15 The specific features are discussed in the 16 topic report are here. You have an overhead heavy 17 load handling system, single failure proof in the 18 event of a seismic event during a module move. There 19 are reactor building walls that protect those modules 20 and the ultimate heat sink.
21 The bioshields also provide an additional 22 potential layer of protection in a beyond design basis 23 event. The spent fuel pool and the ultimate heat sink 24 are communicated together as the same unit water until 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 the evaporation below the rear wall. And I'll note 1
that the ultimately heat sink is also below grade in 2
the event of a beyond design basis event.
3 That is protected from external events by 4
the reactor building. Next slide, please. And then 5
this is a graphical representation of a typical 6
NuScale power module. As you can see, it has its own 7
containment vessel.
8 It has emergency core cooling values for 9
both the vent at the top of the reactor vessel and 10 towards the bottom for recirculation and then the 11 decay hey removal system. Next slide, please. So 12 this is a brief summary of a high level of the 13 mitigation features for design basis events. When an 14 event occurs, safe and stable shutdown is achieved 15 without any operator actions.
16 These include an automatic reactor trip, 17 containment isolation that activates immediately, a 18 decay heat removal system that passively removes decay 19 heat for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If AC power is not restored within 20 those 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the emergency core cooling valves 21 activate. And then the modules rely on no additional 22 equipment to maintain core cooling.
23 The spent fuel pool shares a common water 24 source with the reactor modules, thus only a single 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 volume needs to be maintained for that coping period.
1 We do not credit this for safe-shutdown. But for 72 2
hours, the system response is assured in the main 3
control room.
4 And then last, all of these design-basis 5
mitigation features are then relied upon for beyond 6
design basis events to provide the capability for 7
successful mitigation past large light water reactor 8
design coping periods. Next slide, please. And so as 9
an extension into the mitigation of beyond design 10 basis events, these redundant reliable safety features 11 then are relied upon to mitigate or damage. The 12 immediate plant shutdown and containment isolation at 13 the initiation event prevents release.
14 The containment vessels are protected by 15 the reactor building from external events. And 16 because no operator actions are needed to initiate 17 safe shutdown, further actions aren't needed during 18 the extended coping period. To reflect this, the 19 NuScale implements a set of conditions for use an 20 adopter of this topical report.
21 Firstly, they must have described all of 22 the design features that are in scope of the topical 23 report. They must then provide a thermal analysis to 24 validate a proposed coping period. They must 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 establish a maintenance fuel program in accordance 1
with 50.65 to ensure reliability of the features that 2
are relied upon. And then last, they must have an 3
emergency plan in their application. Next slide, 4
please.
5 MEMBER KIRCHNER: Peter, just to clarify, 6
when you said an emergency plan, is that upper case 7
8 MR. SHAW: Yes.
9 MEMBER KIRCHNER: That's site specific, 10 not an emergency plan for coping?
11 MR. SHAW: Exactly.
12 MEMBER KIRCHNER: Thank you.
13 MR. SHAW: Next slide, please. And so 14 this is a brief summary of the plant systems that are 15 relied upon for a loss of AC power event. The reactor 16 building protects modules from design basis and 17 natural phenomena. And it houses the ultimately heat 18 sink.
19 The control building is relied upon to 20 protect operators from design basis natural phenomena.
21 And augmented direct power system provides continuous 22 DC power for the equipment. The module protection 23 system automatically actuates safe-shutdown functions 24 and don't require new operator actions.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
18 The plant protection system monitors 1
reactor, refueling, and spent fuel conditions, and 2
they're common to all modules. And the safety display 3
and indications system provides non-safety-related 4
accident monitoring.
5 Next slide, please. So, further, the 6
containment isolation is required so that the passive 7
systems can reliably initiate safe-shutdown and 8
prevent an uncontrolled release and also dissipate 9
decay heath.
10 The ultimate heat sink provides common 11 water source for maintaining safe-shutdown. And it 12 also has a Category 1 assured water make-up line. A 13 decay heat removal system removes decay heat from the 14 reactor into the ultimate heat sink. Emergency core 15 cooling circulates reactor coolant into containment to 16 initiate long term passive cooling.
17 A reactor building and control room 18 building ventilation system maintain safe atmospheres 19 for respective environments. An overhead heavy load 20 handling system supports the module during lifts 21 during design-basis natural phenomena. And then last, 22 a communication system used to coordinate operator 23 response and respond to a beyond design basis event.
24 Every NuScale reactor configuration power plant would 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 have to have these features in it to adopt the topical 1
report.
2 CHAIR DIMITRIJEVIC: Hi, this is Vesna.
3 So let me just ask you -- let me just summarize where 4
we are now. So you're basically -- the premise is you 5
can deal with just install existing equipment. You 6
don't need any extra equipment to deal with those 7
events, right?
8 And you can rely just on the plant 9
equipment. And then you describe all of this plant 10 systems which you rely on. So let me ask you a couple 11 of the things.
12
- Like, for
- example, the containment 13 isolation, what does it -- in your TR, it says the 14 containment isolation occur on the loss of AC power 15 instantly. So is that something -- there is no signal 16 needed for containment isolation. It's something 17 which comes automatically with loss of AC power?
18 MR. SHAW: So that is given by the -- so 19 described earlier in the previous slide, the module 20 protection system automatically actuates the 21 containment isolation for the reactor.
22 CHAIR DIMITRIJEVIC: For every trip or 23 just loss of AC power.
24 MR. SHAW: I would actually like 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 clarification on that from Meghan McCloskey.
1 MS. MCCLOSKEY: This is Meghan McCloskey, 2
NuScale. During a loss of AC power event, the module 3
protection system will actuate containment isolation 4
and then -- and reactor trip and DHRS actuation.
5 Containment isolation could also be actuated on a 6
number of other signals such as high containment 7
pressure or I think --
8 (Simultaneous speaking.)
9 CHAIR DIMITRIJEVIC: The normal signals, 10 right.
11 MS. MCCLOSKEY: The normal signals that 12 you're familiar with.
13 CHAIR DIMITRIJEVIC: Yeah, okay. So the 14 loss of AC power will automatically actuate. Let me 15 ask you then also you -- I saw that on your augmented, 16 the DC system, the two channels provide the -- remove 17 power from emergency cooling system valves and open 18 them about one hour. Does that mean it's assumed that 19 decay heat removal will get pressure down enough to 20 enable opening the ESSC valves in one hour?
21 MS. MCCLOSKEY: I'm not sure where the one 22 hour2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> comes from. If AC power supplied to the chargers 23 is not restored, then the module protection system 24 will actuate ECCS after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 CHAIR DIMITRIJEVIC: Sorry, that's what I 1
meant 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. And if I said the one hour, that was 2
a mistake. So you have those two channels, you know 3
EDAS, IMS, right, which provide ECCS hold mode. And 4
after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, they will sort of, like, de-energize 5
those valves, right? Well, my question is would the 6
pressure differential be low enough to enable opening 7
that valve in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?
8 MS. MCCLOSKEY: Yes, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of normal 9
DHRS cooling is very effective to reduce the module 10 pressure. And so for designs that have inadvertent 11 actuation blocks on all of the valves, we would expect 12 the ECCS values to open at that time. And then the 13 design -- the NuScale design that was submitted with 14 the SDAA does not have inadvertent actuation blocks on 15 the vent valves. And so the vent valves would open 16 when actuated and continue depressurizing the reactor 17 coolant system.
18 CHAIR DIMITRIJEVIC: Okay, thanks. So let 19 me then ask you following which is connected with this 20 slide. And that's why I'm asking this question here.
21 You said you don't discuss this extended coping time.
22 But we can discuss 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, right, in the open 23 session? Okay.
24 So what's happening, the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 basically the batteries run out, right? We lost the 1
ventilation system and lights in the control room, 2
right? After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, it will be like a total 3
blackout, right? But we don't really need operator in 4
instrumentation that's a main premise, right?
5 (Simultaneous speaking.)
6 CHAIR DIMITRIJEVIC: I mean, after 72 7
hours.
8 MS. MCCLOSKEY: I think Dan Lassiter here 9
can provide some additional information from the 10 system design perspective.
11 MR. LASSITER: This is Dan Lassiter, 12 NuScale design engineering. You're asking about the 13
-- with regard to the time for ECCS operations?
14 MS. MCCLOSKEY: No, the --
15 (Simultaneous speaking.)
16 CHAIR DIMITRIJEVIC: No, no. I'm asking 17 about the --
18 (Simultaneous speaking.)
19 CHAIR DIMITRIJEVIC: -- instrumentation 20 and control room and things like that because I see 21 here you have it here that somebody apply a system 22 rely upon during loss of AC power. Some of those 23 systems, you rely on after in the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not for 24 that extended time, right?
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 MR.
LASSITER:
- Yeah, there's no 1
requirement for operator action within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 2
design basis event is the bottom line.
3 CHAIR DIMITRIJEVIC: So let me just put my 4
question a little more clear. We have two times which 5
we're going to discuss. One is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> which is what 6
has been done before, right?
7 And now you have introduced another 8
specified extended time which you said we can only 9
discuss in the closed section, right? So I'm only 10 talking now about 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when we're 11 going to enter this other specified time to be 12 discussing in the closed session, there would not be 13 the control room, the ventilation systems, the PAM.
14 And everything will be gone because the 15 augmented DC power system will be gone in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
16 Is that a true statement? After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, we will not 17 have any instrumentation and control room will not be 18 able to be habitable, right?
19 MS. MCCLOSKEY: Just a second.
20 CHAIR DIMITRIJEVIC: Sure.
21 MR. LASSITER: It is correct that the 22 control room habitability system and the PAM battery 23 provisions -- excuse me, post-accident monitoring 24 battery provisions for the instrumentation is designed 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
24 for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, we expect there to be 1
time for operators to take appropriate actions based 2
on the accident.
3 MS. MCCLOSKEY: And this is Meghan again.
4 The plant is in a safe -- the modules are in a safe, 5
stable condition throughout that 72-hour duration and 6
the spent fuel is being cooled. And so the plant 7
personnel have that time frame to respond to the event 8
as appropriate for whatever is going on.
9 CHAIR DIMITRIJEVIC: I see. Okay. Well, 10 then I sort of -- that was my impression. You're 11 qualifying all of this for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not for that 12 other extended specified time which we cannot discuss 13 now. So all right. Well, okay, all right. I 14 understand this now. So let me also ask you why do 15 you have to rely on this load handling system for the 16 loss of AC power?
17 MS. MCCLOSKEY: I think Erwin may have 18 something.
19 MR. LAUREANO: Yeah, this is Erwin 20 Laureano, Plant Office Manager for NuScale. The 21 overhead heavy load handling system, that's just being 22 mentioned in the event that a NuScale power module is 23 in transition to the containment plant tool or it's 24 being held in the containment plant tool which is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 seismically qualified.
1 MR. LASSITER: Yeah, it's not an active 2
safety system. It's just designed to keep the module 3
in a save condition in the event of loss of power in 4
that scenario.
5 CHAIR DIMITRIJEVIC: Okay. All right.
6 Thanks.
7 MR. LAUREANO: You're welcome.
8 MR. SHAW: Next slide, please. And so 9
following from the loss of AC power mitigations, the 10 filament of the other regulations actually follows a 11 very similar mitigation. In this case, the station 12 blackout has the same mitigations of a safe stable 13 shutdown as said within that 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, no operator 14 reactions. Because the spent fuel pool shares a 15 common water source, only a single volume needs to be 16 maintained for that extended coping period.
17 For 10 CFR 50.155(b)(2) and (c), exterior 18 concrete walls for the reactor building are designed 19 to withstand aircraft impacts in similar events for 20 loss of a large area. Next, the training requirements 21 for these components are not needed because they are 22 in the described systems that they operators will 23 already be trained and qualified for. And then last 24 for spent fuel pool monitoring until the final fuel 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
26 removal from the reactor vessel, 51.55(e), a spent 1
fuel pool monitoring is common to the ultimate heat 2
sink and the instrumentation needs the guidance of NEI 3
1202. Next slide, please.
4 So the review with the staff, there were 5
24 audit questions that were asked, one docketed 6
response. And there were no RAIs for the review.
7 During their review, the NRC identified several site 8
specific considerations based on plant location.
9 Given that our extended coping period is so long, 10 additional considerations were needed. For these 11 limitations and conditions, most of these would 12 normally be a part of a site selection and they would 13 be --
14 An adopter would also supply why no 15 preplanning is needed using this methodology 451.55.
16 And if not, an adopter must identify water for an 17 ultimate heat sink, necessary mode of equipment such 18 as pumps and generators, equipment for debris removal, 19 address the statement of operators past the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
20 Address all personnel, will ascertain plant conditions 21 in a post-event, and address all spent fuel level 22 instrumentation. Power sources will be replaced post-23 event.
24 And applicant must also provide a fire 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
27 protection in accordance with 50.48 and a training 1
program. That includes this relied upon for beyond 2
design basis event. Next slide, please. And so as 3
summary, NuScale power plant design provides extended 4
coping capabilities for mitigating beyond design basis 5
events. Operator actions are not necessary to 6
establish the safe shutdown condition. And the 7
topical report establishes a consistent approach for 8
remedy at a NuScale power plant to demonstrate 9
compliance to 10 CFR 50.155.
10 CHAIR DIMITRIJEVIC: Okay. Thank you.
11 And who will be presenting from NRC?
12 MR. HAYDEN: That's me, Tommy Hayden.
13 CHAIR DIMITRIJEVIC: Okay.
14 MR. SNODDERLY: We need maybe just a few 15 minute break or we need to move out the NuScale folks 16 and then move in staff able to come and put their 17 slides up. So give us one minute and we'll let you 18 know when we've made that --
19 (Simultaneous speaking.)
20 CHAIR DIMITRIJEVIC: Okay, okay.
21 MR. SNODDERLY: Thank you.
22 (Pause.)
23 MEMBER KIRCHNER: Are you ready, Tom?
24 MR. HAYDEN: Yes, I'm ready.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 MEMBER KIRCHNER: Vesna, so I think the 1
NRR staff is ready. So go ahead.
2 CHAIR DIMITRIJEVIC: Okay. Go ahead, 3
please.
4 MR. HAYDEN: Yeah, sure thing. This is 5
Tommy Hayden from the NRR DNRL NRLB branch, new 6
reactor licensing branch. I'll be presenting the 7
staff's evaluation of the topical report.
8 So for a number of you, NuScale submitted 9
the topical report for the mitigation of beyond design 10 basis events, Rev. 0 on September 11, 2023 and Rev. 1 11 on June 26, 2024. As NuScale mentioned, the NRC 12 regulatory audit was performed on September 2023 to 13 March 2024. And 24 issues were generated. Eleven of 14 those were resolved via audit responses and 13 were 15 resolved via limitations and conditions.
16 NuScale submitted one piece of 17 supplemental information to the docket and no RAIs 18 were issued. Staff completed the topical report 19 review and issued an advanced safety evaluation to 20 support today's ACRS subcommittee meeting. Here's the 21 technical reviewers for the topical report: Hanry 22 Wagage, Raul Hernandez, Angelo Stubbs, Josh Miller, 23 Ryan Nolan, John Hughey, Thinh Dinh, Marie Pohida, 24 Sheila Ray, Nick Hansing, and myself and Getachew 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
29 Tesfaye with our project managers for this topical 1
report.
2 DR. BLEY: Hey, this is Dennis Bley. Can 3
you go back two slides from there? One more, yeah.
4 Is it pretty unusual, you have no RAIs? Were you able 5
to -- I guess the audits gave you the flexibility to 6
resolve issues that were concerning you. Is that 7
right?
8 MR. HAYDEN: That's correct.
9 DR. BLEY: Okay. Go ahead.
10 MR. HAYDEN: Okay. Here's a layout of the 11 safety evaluation for this topical report. You'll see 12 the bolded sections are the sections that we've 13 created slides for that we'll go into a little more 14 depth on. So Section 2 is the background and then 15 Section 3 is technical evaluation and each of the 16 sections within that evaluation.
17 Section 4 is the conclusion. You'll 18 notice Section 5, limitations and conditions, is not 19
-- we don't have a slide made out for that explicitly.
20 But inside each of the technical evaluation slides, 21 the applicable L&C is discussed there.
22 Section 2
background, so 2.1 has 23 regulatory requirements and regulatory and industry 24 guidance. And then Section 2.2 is the summary of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
30 technical information. So the TR is applicable only 1
to NuScale's small modular reactor designs with 2
specific characteristics.
3 As mentioned in NuScale's presentation, 4
there's also conditions of use that were laid out in 5
the topical report. And the TR contains a description 6
of the NuScale design capabilities and features, 7
including passive safety systems capable of 8
maintaining core cooling containment and cooling 9
functions. And the large reactor pool serving as the 10 ultimate heat sink.
11 The topical report specifies how these 12 features enable a design to mitigate beyond design 13 basis events for a specified extended duration without 14 the need for AC power, special equipment, or 15 additional guidelines and strategies. 3.1 is the 16 plant baseline coping criteria for loss of all AC 17 power. 3.1.1 is the assessment of the electrical 18 power.
19 In the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of a loss of all AC 20 power is identical to a station blackout and no AC 21 power is relied upon for performing safety functions.
22 The initial conditions and assumptions in NE 12-06, 23 Rev. 4 and Reg Guide 1.226, Rev. 0, assume that 24 station batteries would remain available following a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
31 beyond design basis event since they are considered 1
robust. In the TR, Section 3.1.5, Initial Event 2
Conditions and Assumptions, NuScale states that for 3
the baseline coping capability, station batteries and 4
associated DC buses remain available for the design 5
operating of the station batteries, and installed 6
electrical distribution system, including inverters 7
and battery chargers, remain available provided they 8
are seismic Cat. 1.
9 Initial assumptions in the TR following a 10 beyond design basis external event are consistent with 11 NE 12-06, Rev. 4 guidance, endorsed by Reg Guide 12 1.226. 3.1.1 continued, the ultimate heat sink 13 monitoring, including spent fuel pool level, is 14 assured for a time frame in excess of the 7 days post-15 event minimum in NEI 12-02. The staff notes that the 16 ultimate heat sink and spent fuel pool level 17 instruments can be powered by multiple means.
18 The capacity and capability of such 19 multiple means are not discussed. And therefore, an 20 applicant must satisfy L&C 5.2. L&C 5.2 addresses how 21 plan operators will ensure, during the initial coping 22 phase, that the following can be achieved to provide 23 makeup to the ultimate heat sink.
24 The source of water is identified and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 available in sufficient quantity. Necessary motive 1
equipment such as pumps and generators and the 2
required electrical power and fuel, can be obtained, 3
staged, and implemented. And required debris removal 4
will be accomplished to support placement of equipment 5
and access to site connections.
6 3.1.2 is plant design capabilities.
7 Following a loss of all AC power event, automatic 8
responses of safety-related equipment establish and 9
maintain core cooling, containment, and spent fuel 10 pool cooling by placing the reactor modules and spent 11 fuel into a safe, stable, shutdown state with passive 12 cooling. 3.1.2 also introduces L&C No. 5.1.
13 To accomplish this, an adopter must 14 satisfy Conditions of Use in TR Section 1.3, by 15 providing a plant specific design, as described within 16 the report, plant specific thermal analysis, including 17 configuration of the plant, number of modules, spent 18 fuel pool capacity for all modes of operation, a 19 maintenance rule program and an emergency plan. And 20 I bolded the second Condition of Use there as that is 21 the portion of L&C 5.1 that the safety evaluation in 22 Section 3.1 captures. 3.2, plant systems and 23 responses to a loss of all AC power event.
24 10 CFR 50.155 requires that equipment have 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 the capacity and protection to be used to support 1
mitigation of beyond design basis events. TR Section 2
4.0 lists proposed equipment for a NuScale SMR design 3
as NuScale just showed. The TR also lists the 4
protection of this equipment.
5 The design specific and location specific 6
nature of the MBDBE necessitates analysis from a COL 7
applicant. The analysis would need to confirm the 8
assumptions in the TR remain true for the proposed 9
plant as well as contain analysis to show that the 10 proposed plant systems would be able to respond to the 11 progression of the event for the proposed period of 12 time. Again, L&C 5.1 is referenced here, and this was 13 added to address the need for the connection of the 14 analysis and the plant system responses for the event.
15 3.3 is safety functions during a loss of 16 all AC power. 10 CFR 50.155 states that applicants 17 shall develop, implement, and maintain strategies and 18 guidelines to maintain or restore core cooling, 19 containment, and spent fuel pool cooling capabilities.
20 The TR discusses the strategies for maintaining these 21 functions during the event and how it is accomplished 22 with a NuScale SMR.
23 The site-specific nature of hazards and 24 potential site layouts and locations requires that a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
34 COL applicant address the potential events at their 1
specified site. L&C 5.1 addresses the need for an 2
applicant to provide analysis and potential guidelines 3
and strategies dependent on that analysis. 3.3.2, 4
indefinite maintenance of core cooling, containment, 5
and spent fuel pool cooling capabilities.
6 The regulation requires that offsite 7
assistance and resources be acquired indefinitely or 8
until functional capabilities can be maintained 9
without the need for mitigating strategies. The 10 endorsed guidance in NEI 12-06 recognizes that FLEX 11 strategies and offsite resources do not need to be 12 explicitly planned for the period beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
13 The TR proposes that no pre-planning for mitigating 14 actions is required because a NuScale SMR design 15 provides beyond design basis external event mitigating 16 capability beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
17 3.3.2, continued. However, the guidance 18 in NEI 12-06 also presumes that initial coping 19 mitigating strategies have been established such that 20 staging areas to receive offsite equipment are 21 identified; means are established to transport 22 equipment to deployment areas; the ability of an 23 offsite organization to provide support has been 24 ensured; and standard FLEX equipment such as 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 mechanical and electrical connectors that are 1
compatible with site connections are obtained. L&Cs 2
5.2, 5.3, 5.4, and 5.5 address the requirement for 3
long term mitigating capabilities associated with the 4
NuScale SMR design discussed in the TR.
5 MEMBER HALNON: Tommy, this is Greg. It 6
appears that -- trying to reconcile their statement 7
that no pre-planning is required, yet in their TR, 8
they also say that pre-planning is required. Is that 9
how you all took it? I mean, staging areas, means are 10 established, stability offsite, that's all pre-11 planning stuff. When they say no pre-planning is 12 required, what do you they mean? How did you take it?
13 MR. HAYDEN: Josh or John, if you have 14 that -- if you're able to answer that. There's a mic 15 right there.
16 MR. HUGHEY: John Hughey, PRA division, 17 Oversight Branch. So my understanding of the TR is 18 that the assertion was no pre-planning was required.
19 And the basis for that was their design was able to 20 cope with the -- cope without any planning for at 21 least the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period.
22 So the guidance says after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> we 23 acknowledge that you're going to be able to continue 24 to indefinitely get resources. You don't have to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
36 explicitly plan that. However, with the NuScale 1
design, there was no pre-planning considered necessary 2
at all.
3 Therefore, those characteristics that 4
would have been in place with the initial mitigating 5
strategies will not be in place according to the 6
topical report. So the TR was essentially saying no 7
pre-planning was required at all in RSE where we are 8
essentially saying some pre-planning is necessary 9
unless for some reason an applicant is at a site and 10 for one or more or all of the elements, they can make 11 a justifying argument for why they in their particular 12 situation don't need pre-planning.
13 MEMBER HALNON: Okay. Thank you. That 14 explains it.
15 DR. BLEY: This is Dennis Bley. In past 16 discussions, and I think there was no commitments that 17 I remember about FLEX safer and info. And then the 18 statements certified by FLEX essentially a requirement 19 if you're going to (audio interference).
20 CHAIR DIMITRIJEVIC: Dennis, you are 21 breaking up. I cannot hear you.
22 MEMBER HALNON: Dennis, it sounds like 23 your connection is poor. We can't catch any of the 24 words.
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37 (Audio interference).
1 DR. BLEY: They're working on it.
2 MEMBER HALNON: We got that they're 3
working on it.
4 CHAIR DIMITRIJEVIC: Yes, Dennis, it's 5
tough to understand what you're saying, so --
6 DR. BLEY: (Audio interference.)
7 MEMBER HALNON: Put it in chat, then we 8
can read it.
9 MEMBER KIRCHNER: Dennis, this is Walt.
10 If you can hear me, if you could just text your 11 question to Mike Snodderly, we'll loop back and try 12 and address it. But we're having problems with your 13 audio. Tom, why don't you go on and then we'll try 14 and loop back to Dennis' question.
15 MR. HAYDEN: Sure thing. Perhaps seeing 16
-- and I'm seeing that are referenced on this slide, 17 on the next might shed some light. So L&C 5.2 18 specifically addresses inventory makeup to the 19 ultimate heat sink. So identification of an available 20 and sufficient source of water; acquisition, staging, 21 and implementation of necessary motive equipment; and 22 provision of debris removal capability to deploy 23 offsite equipment.
24 L&C 5.3 specifically addresses the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
38 assigned control room monitoring function. So 1
sustaining control room operators for the 72-hour 2
period after the start of the beyond design basis 3
external event, provision of debris removal to allow 4
egress from the control room. L&C 5.4 specifically 5
addresses the capability of site support personnel to 6
ascertain plant conditions to determine necessary 7
coping requirements once onsite power systems are 8
depleted. And L&C 5.5 specifically addresses long-9 term support related spent fuel pool level monitoring 10 instrumentation. And as John mentioned, a provision 11 is included in the above L&Cs to allow an applicant to 12 provide a justification that supports not implementing 13 the elements of that L&C.
14 MEMBER HALNON: Tommy, this is Greg. I 15 would assume that you'd want at least the equivalent 16 capabilities of those L&Cs.
17 MR. HAYDEN: Right. And that would be 18 encompassed in the justification.
19 MEMBER HALNON: You're talking about pre-20 planning the implementation. They need to be able to 21 do those. That would be in the justification. We 22 don't have to pre-plan for it. Is that correct?
23 MR. HAYDEN: Could you repeat that?
24 MEMBER HALNON: Yeah, you mentioned if 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 they had justification to where they did not have to 1
implement those L&Cs. My point is that they're asking 2
for that you don't have to pre-plan those elements.
3 But you still need to be capable of doing those 4
things.
5 MR. HAYDEN: Yeah, so I'd have to refer 6
back to the L&C as they're specifically written. But 7
that might be --
8 MEMBER HALNON: In general, you want at 9
least the equivalent capability to be able to restore 10 or replenish or whatever case --
11 MR. HAYDEN: Sure.
12 MEMBER HALNON: Yeah, okay. So it's not 13 that you don't want to be able to.
14 (Simultaneous speaking.)
15 MEMBER HALNON: Sustaining operations for 16 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. But you're not saying that they can justify 17 not doing it for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. They just don't have to 18 pre-plan for it. The way that I read the TR was that 19 they didn't have to do pre-planning. I would assume 20 that you would expect in that justification at least 21 the equivalent capabilities to do those things.
22 MR. HAYDEN: Yeah, I understand your 23 question. I think I'd want to re-read specifically 24 how the L&C is worded --
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40 MEMBER HALNON: Okay. That's fair enough.
1 MR. HAYDEN: -- before I answer that.
2 John, if you're capable of answering that now, that's 3
fine. Or I can take that and I can make sure I have 4
an answer by the end.
5 MR. HUGHEY: No, I think I can answer.
6 MR. HAYDEN: Thanks, John.
7 MR. HUGHEY: You are correct. So for an 8
example, the first L&C talks about identifying a 9
makeup water source, an inventory source for the 10 ultimate heat sink. So for example, if one of these 11 plants was located next to a very large river, very 12 large lake, the ocean, somewhere where it was obvious 13 that was the water source. And in their operating 14 procedures, they already had procedural direction for 15 how to get water from there.
16 It was protected for many environmental --
17 so if they already had things like that built into the 18 design that were obvious, then there would be no need 19 for that pre-planning. But you are correct. The 20 issue is pre-planning. The plant still needs to be 21 able to perform the mitigating capability.
22 MEMBER HALNON: Thank you.
23 MR. HAYDEN: Yeah, I think it may be 24 misleading how that last bullet there is worded, what 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 is in the L&C. You can justify not accomplishing them 1
in their entirety, but the pre-planning part.
2 MEMBER HALNON: Right. So I just want to 3
make sure we're focused on just the pre-planning.
4 MR. HAYDEN: Okay. Thank you. Okay.
5 3.4, capability to respond to a loss of large area due 6
to explosion or fire event as required by 10 CFR 7
50.155(b)(2). That requirement, in part, is in 8
regards to a LOLA. Strategies and guidelines must 9
address in a three-phase approach, Phase 1, enhanced 10 firefighting capabilities, Phase 2, measures to 11 mitigate damage to fuel in the spent fuel pool, and 12 Phase 3, measures to mitigate damage to fuel in the 13 reactor vessel and to minimize radiological release.
14 The TR follows guidance in NUREG-0800, 15 Temporary Instruction 2515/168, and NEI 06-12. For 16 Phase 1, the TR lists design features that cope with 17 potential loss of large area events, but it did not 18 provide plant-specific features to demonstrate 19 adequate LOLA coping capability. For Phase 2, the TR 20 provides generic NuScale plant spent fuel pool design 21 features that mitigate damage to fuel in the spent 22 fuel pool. But it did not provide certain plant-23 specific details, for example, minimum spent fuel pool 24 water level.
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42 MR. SNODDERLY: Excuse me. This is Mike 1
Snodderly from the ACRS staff. I have Dennis Bley's 2
question. His question is, so would a NuScale 3
referencing this topical report, would they have to 4
have access to a safer type facility to obtain the 5
FLEX equipment or not?
6 MR. HAYDEN: I think the answer is not.
7 They don't need a safer center. They have the 8
equipment on site.
9 MR. WAGAGE: Yes, NuScale does not rely on 10 FLEX equipment or the coping capability.
11 MEMBER HALNON: Hanry, you need to state 12 your name for the court reporter.
13 MR. WAGAGE: My name is Hanry Wagage from 14 NRR. NuScale does not need any FLEX capabilities for 15 this presented strategies.
16 MEMBER HALNON: And that includes onsite 17 equipment and safer beyond --
18 MR. WAGAGE: Because you can then talk in 19 the closed session that for a long time.
20 MEMBER HALNON: I think that was Dennis' 21 question was all the commitments the industry made for 22 FLEX equipment safe for Phoenix and Memphis or really 23 where those things are -- that's not required here.
24 MR. WAGAGE: It's not requiring FLEX 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
43 equipment.
1 CHAIR DIMITRIJEVIC: Well, maybe all of 2
this will be easier to be discussed in the closed 3
section because now I'm sort of, like, everything is 4
for here, Like what is for closed or for the open 5
session. But one of the main ideas in the TR is that 6
after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, you're still not in a hurry. There's 7
still time available to get all equipment you need and 8
organize those things.
9 So it is a slightly -- and this is where 10 they will introduce this addition of time in analysis 11 available for the -- to take an action, get organized, 12 and see what needs to be done. So I think that maybe 13 this will be better discussed in the closed session 14 where we can discuss this extended after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />' 15 time. That's opinion about the subject. All right.
16 MR. HAYDEN: Sure. So I'll continue on 17 with 3.4, capability to respond to a loss of a large 18 area event. For Phase 3, the TR provides generic 19 NuScale design features that are used to mitigate 20 damage to fuel in the reactor vessel and minimize 21 radiological release by plant key safety functions.
22 But it did not provide plant-specific design features.
23 So for those reasons, L&C 5.1 for Phases 24 2 and 3 and L&C 5.6 for Phase 1 were written into the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 evaluation. And 5.1, we've gone over. But it 1
addresses plant-specific design features, including 2
key design features that would preclude the need for 3
an enhanced spent fuel pool mitigation capability such 4
as diverse or portable makeup capability to mitigate 5
a LOLA event and maintain key safety functions to 6
mitigate potential fuel damage and radiological 7
release due to a LOLA event.
8 And L&C 5.6 for Phase 1, and that 9
addresses the Fire Protection Program, including 10 plant-specific design features and procedures, that 11 would provide assurance of adequate LOLA coping 12 capability. 3.5 capacity, capability, and protection 13 of equipment associated with mitigation of events 14 described in the rule, as required by 10 CFR 15 50.155(c). This regulation requires the equipment 16 that is relied on for the mitigation strategies and 17 guidelines have sufficient capacity and capability to 18 perform the intended functions and be reasonably 19 protected from the effects of natural phenomena that 20 are equivalent in magnitude to the phenomena assumed 21 for developing the design basis of the facility.
22 The TR provides the equipment relied upon 23 for the mitigation strategies and guidelines. The TR 24 also describes the
- capacity, capability, and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
45 protection of the equipment from natural phenomena.
1 Subject to satisfaction of L&C 5.1, and applicant or 2
licensee adopting the TR for a NuScale design could 3
demonstrate compliance with 10 CFR 50.155(c). 3.6, 4
training requirements as defined by 10 CFR 50.155(d).
5 NEI 12-06, Section 11.6 describes the 6
training to be provided to address this regulation.
7 The TR provided justification why specific training is 8
not needed to address 10 CFR 50.155(d), but did not 9
specifically address the required activities related 10 to replacement of spent fuel pool level monitoring 11 power supply. For that reason, the staff introduced 12 L&C 5.7 which requests an adopter of the topical 13 report to confirm that the training program includes 14 the required activities related to replacement of 15 spent fuel pool level monitoring power supply.
16 3.7, spent fuel pool monitoring after 17 final fuel removal from the reactor vessel. The staff 18 has endorsed NEI 12-02 which describes the design 19 criteria for spent fuel pool monitoring system that 20 meets the requirements of 10 CFR 50.155(e). The TR 21 discussed the level instruments design criteria and 22 indicated that they meet the guidance of NEI 12-02, 23 but it did not address the training requirements.
24 For that reason, L&C 5.7 was used in this 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
46 section of the evaluation as well. So finally, 1
Section 4, our conclusion. Based upon its review as 2
discussed above, subject to the limitations and 3
conditions described in Section 5 of the evaluation, 4
the NRC staff concludes that this topical report 5
provides a reasonable methodology for an applicant or 6
licensee to demonstrate the NuScale plant design 7
capability to mitigate beyond design basis events as 8
defined by 10 CFR 50.155.
9 MEMBER HALNON: Tommy, what is the process 10 now for their topical? I mean, a lot of those limits 11 and conditions were what I would consider they 12 should've been there. Someone should've thought it 13 out. It's pretty -- establish nuclear cultures to 14 take it one step further and make sure that these 15 capabilities are all there.
16 Are they going to revise the topical so 17 that the revised one will have those items? Or is it 18
-- because the SER itself is not part of a licensing 19 basis. It's the topical report and their response to 20 the SER. Are you expecting a response to the SER with 21 a revision or an acceptance letter or something?
22 MR. HAYDEN: I would let NuScale answer 23 that. But I do not expect that.
24 MR. CUMMINGS: This is Kris Cummings with 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 NuScale. So I'd actually say the SER is part of the 1
topical report. So we'll get the SER. We've seen it.
2 We've provided our comments on it.
3 We will get the final SER, and then we 4
will submit to the NRC for verification that the 5
approved topical report will include the SER. It will 6
also include the one audit response that we docketed 7
and the topical report itself. So when we have an 8
approved topical report, it will include the SER as 9
part of that.
10 MEMBER HALNON: Okay. So in the future if 11 an applicant uses it, they're going to have all three 12 of those. It's going to be one on the one title 13 cover. The topical report will have the SER and 14 question and answers in there?
15 MR. CUMMINGS: That's correct. And an 16 adopter of the topical report would need to when they 17 adopt it show how they meet the limitations and 18 conditions and the conditions of use --
19 (Simultaneous speaking.)
20 MEMBER HALNON: I just want to make sure 21 that was clear because there's a lot of -- 13 limits 22 and conditions is a lot for this focused report, I 23 think. It felt like you had to tell them a lot.
24 MR. HAYDEN: I hear what you're saying.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
48 And I think from NuScale's perspective many of these 1
would be or will be answered in the site specific 2
application.
3 MEMBER HALNON: Yeah, okay.
4 MR. HAYDEN: And so for them, that was 5
part of their consideration.
6 MEMBER HALNON: So they roll it into the 7
8 (Simultaneous speaking.)
9 MR. HAYDEN: And our evaluation would 10 obviously look to ensure that satisfies the L&Cs, 11 whatever is put into that specifically.
12 MEMBER HALNON: Okay. Thank you for that 13 clarification. Thank you.
14 MEMBER PALMTAG: This is Scott Palmtag.
15 I just had a question about the FLEX. You said that 16 the NuScale reactors won't have FLEX?
17 MR. WAGAGE: The answer is yes.
18 (Simultaneous speaking.)
19 MR. WAGAGE: My name is Hanry Wagage.
20 MEMBER KIRCHNER: Yeah, just pull the 21 microphone closer to you.
22 MR. WAGAGE: I said yes. And the question 23 is how do you need the reminder for going to -- the 24 question is -- the answer is yes.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
49 MR. HAYDEN: It will not have it?
1 MR. WAGAGE: It will not. It will not 2
need any FLEX. No, sir.
3 MR. HAYDEN: He kind of explained it to me 4
because it seemed like the -- I don't understand the 5
exact regulations and FLEX, I'm not familiar with 6
them. But it seems like FLEX was for beyond design 7
basis for some unanticipated situation. It also 8
seemed like very low hanging fruit.
9 MR. WAGAGE: Yes.
10 MR. HAYDEN: Why would you know want FLEX?
11 MR. WAGAGE: I think you must be thinking 12 about operating reactors at three phases. So 13 addressing mitigation of strategies. The first one is 14 Phase 1 is to use plant equipment.
15 Then after using the plant equipment, then 16 all the capability is gone. And they will rely on 17 FLEX equipment. FLEX equipment is other equipment 18 available at the facility.
19 They can use, then the third phase is when 20 getting help from -- getting resources from outside to 21 help with the situation. But NuScale is passive 22 reactor. It can more than what other operating 23 reactors do because we think that needs a pinpoint 24 that can keep between actors in the reactivities and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
50 the reason that passive design -- and they'll need --
1 MR. HAYDEN: I understand it's passive.
2 But it just seems like FLEX is that extra layer in 3
case something unanticipated comes along.
4 MR. WAGAGE: Yes, but the adopter of this 5
topical report has to show that core cooling 6
containment and spent fuel pool can be maintained for 7
the expected duration. So they have to show by 8
thermal analysis.
9 MEMBER HALNON: Scott, that point is 10 really aimed at the large light water reactors where 11 it could possibly have problems in those three areas.
12 And this topical shows that those three areas are 13 covered by analysis and in plant design. So there 14 would be no reason -- those vulnerabilities don't 15 exist.
16 MEMBER PALMTAG: I understand that. And 17 it's not a requirement. I understand. It just seemed 18 like that'd be a low hanging fruit to have that just 19 in case. But if it's not required, it's not required.
20 CHAIR DIMITRIJEVIC: Well, the basic idea 21 they're trying to say after initiating safe shutdown 22 in the beginning, no active systems are required.
23 Everything is managed possibly without operator 24 actions. So basically, therefore there is no FLEX 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 equipment that'd be necessary for this situation.
1 I'm not sure here why you call this TR 2
that describes methodologies. It's not really 3
methodology. It's sort of like the plant design which 4
enables this to be a situation.
5 So in this statement, it says provides 6
reasonable methodology.
This is not really 7
methodology. But this is just like some comments on 8
that statement.
9 MR. CUMMINGS: This is Kris Cummings from 10 NuScale. I just wanted to respond to Scott's comment 11 about the safer facility since that's on the record in 12 the open session. I want to be able to address that.
13 You're right. We don't anticipate needing 14 to subscribe to the safer -- the FLEX facilities. At 15 the end, you spend a lot of time and effort to do for 16 the light water reactors.
17 And that's because it's a NuScale design 18 itself, the plant and the extended coping period that 19 we'll talk about in the closed session that provides 20 the real capacity to mitigate beyond design basis 21 events. So that extended duration allows you to have 22 a lot of time to be able to use something on a plant 23 specific basis whereas the FLEX facilities in the 24 safer centers were to be able to get equipment to the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 large light water plants that needed it very, very 1
quickly. They don't have those large time frames. So 2
we think this is a very appropriate response because 3
of the design of the NuScale plant.
4 MEMBER PALMTAG: Thank you for the 5
explanation.
6 CHAIR DIMITRIJEVIC: All right. Any more 7
questions from the members? And then --
8 MEMBER ROBERTS: This is Tom Roberts. I'm 9
trying to follow up on Scott's question that there's 10 probably, I'm trying to count, 13, 14 systems in the 11 applicant's slides that all are accredited to avoid 12 having to rely on any of the FLEX equipment. It seems 13 like it'd cause low hanging fruit to have the 14 capability of acting in one of those systems that are 15 accredited. I was wondering if they had thought of 16 that --
17 (Simultaneous speaking.)
18 MEMBER ROBERTS: -- reactor building is 19 probably the close to obvious one. The assumption is 20 that the reactor is capable of holding the UHS water 21
-- spent fuel pool water and protecting all the SSEs 22 inside the reactor building. And so if you had some 23 sort of hole in the floor, that's suddenly going to 24 take a lot of the water away potentially. And I 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
53 assume the reason why you don't think that can happen.
1 But having the ability to bring in your equipment 2
would seem like come from one of those unknowns.
3 CHAIR DIMITRIJEVIC: I propose we can 4
continue discussing this in the closed session when we 5
will have a little better feeling what is the main 6
idea in this proposal. I think NuScale doesn't want 7
to make a comment on this latest proposal. Then I 8
propose that we open for public comments.
9 MR. CUMMINGS: Well, Vesna, this is Kris.
10 I guess I'm elected to let that comment stand without 11 a response. So let me just comment, right? 51.55 is 12 promulgated after Fukushima.
And that was 13 specifically to consider a loss of all AC power, 14 right?
15 And the specific event by which you lose 16 AC power is not considered, right? But there are 17 certain things that you can credit within the context 18 of 51.55. And that includes reliance upon your 19 safety-related equipment.
20 So the UHS is safety-related. It's 21 designed to cite us in the Category 1. So as we 22 describe in the topical report, that sort of a 23 structure, the reactor building is expected to survive 24 whatever the beyond design basis event is because it's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 designed to essentially survive.
1 So that's the context of the rule. And so 2
that's the context of the topical report. So we want 3
to consider specifically that sort of a situation in 4
this particular topical report for addressing 51.55.
5 MEMBER ROBERTS: Okay. Thank you. I 6
assume if a NuScale reactor found itself in an 7
unexpected position they would use resources available 8
9 MR. CUMMINGS: That's right.
10 MEMBER ROBERTS: -- to get whatever they 11 needed to --
12 MR. CUMMINGS: There's an emergency plan.
13 There's FEMA. There's ways to get other equipment on 14 site, things to mitigate events that go beyond that.
15 MEMBER ROBERTS: Okay. Thank you.
16 MEMBER HALNON: Tommy, before we go back 17 to Vesna, could you go back to your slide 13 here? I 18 just wanted to go back to the crux of Dennis' question 19 ant that's standard FLEX equipment such as mechanical 20 connectors are compatible. We've been talking about 21 FLEX is not necessary. So why do that?
22 MR. HAYDEN: I guess I'll answer that by 23 saying that this is meant to show what the guidance in 24 the NEI, it proposes, right?
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 MEMBER HALNON: Okay. So that's the NEI 1
guidance, not --
2 (Simultaneous speaking.)
3 MR. HAYDEN: And then we talk about the 4
L&Cs that we are having adopted.
5 MEMBER HALNON: So that's fine. I think 6
that will clarify for Dennis.
7 MR. HAYDEN: Okay.
8 MEMBER HALNON: So that's the NEI 9
guidance. And we're saying that they don't need to do 10 that.
11 MR. HAYDEN: Right. In the previous 12 slide, it talks about what the TR provided. This 13 slide, it talks about however we do understand, we 14 recognize what this guidance states. And so here are 15 the four L&Cs that we've adopted.
16 MEMBER HALNON: Okay. I think that 17 addresses Dennis' question.
18 MR. HAYDEN: Okay, great.
19 MR. CUMMINGS: So Scott, I just had one 20 more. This is a minor detail. But at the very 21 beginning, you said this TR applies to every NuScale 22 reactor. I'm not sure what the exact wording was.
23 MR. HAYDEN: If that was the wording, that 24 was not necessarily intended.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 MR. HUGHEY: It's specifically the 460 and 1
the 600.
2 MR. HAYDEN: No, that is not correct.
3 It's not specific to either of those two designs. But 4
it's not specific to -- you couldn't say any design.
5 There is a criteria that's established in the topical 6
report itself. And it's proprietary, but --
7 MR. HUGHEY: Any NuScale reactor with 8
these criteria, if I remember right?
9 MR. HAYDEN: Yes, I see Milton. I just 10 wanted to make sure I answer Scott's question first.
11 MR. HUGHEY: I just wondered if that 12 should be -- because they come up with a new change, 13 major change.
14 MR. HAYDEN: Right, if they have a new 15 change and then that change somehow removes one of 16 these systems or design features that this topical 17 report relies upon, then it would not be a topical 18 report.
19 MEMBER KIRCHNER: The area would be a 20 thermal analysis of what's going on with the UHS.
21 Assuming that we're looking at a nominal NuScale 22 module. Limiting actor here would be the passive 23 cooling and the UHS capacity, right?
24 We should discuss some aspects, I guess, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
57 further in a closed meeting. But I think it is in the 1
open that the, for example, NuScale or the latest SDAA 2
has changed the level in the UHS. But it would seem 3
to me that in general that you, the NRC, would look at 4
whatever the proposed level is and then look at how 5
that supports the coping analysis to see if they can 6
use 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and then go beyond as part of the COL 7
review.
8 MR. HUGHEY: Right, yeah. L&C 5.1 9
references those four conditions of use. And those 10 conditions of use are housed in the TR itself. So the 11 TR has to -- as part of -- likely as part of their 12 application -- include that thermal analysis.
13 MR. HAYDEN: So for example, if they had 14 enough power for ten years down the road, your 15 criteria would cover that. They would have to show 16 that would be applicable?
17 MR. HUGHEY: Yeah, within that change 18 process, yes. Go ahead.
19 MR. VALENTIN: I'm sorry. I'm Milton 20 Valentin. I am the supervisor for the containment and 21 plant systems branch. And I just wanted to provide 22 some context.
23 When the staff was looking at this report, 24 we need to keep in mind the possibilities of the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 different design features and the site characteristics 1
that will then define what will be needed at the 2
specific site, either the use of FLEX equipment, 3
either the reliance of certain systems or components 4
that are now being generally created for these 5
capabilities. So we understand that we are going here 6
assuming generic things that might not be as 7
descriptive as ideally we have seen for operating 8
reactors. But the reality is that we don't have the 9
information right now.
10 This is something that we're trying to 11 capture in a way that in the future when we see the 12 specific designs for these sites using these designs 13 at the sites, all the information, then anyone will 14 take on and be fair on looking at all of the details 15 that will be important to make a safety finding. So 16 here this is a very unconventional approach that we --
17 again, we're trying to come up from a place of giving 18 credit for what we know but also keeping in mind other 19 possibilities that could happen once we have the 20 details. So I hope that helps everyone understand 21 with some of the challenges that the staff has dealt 22 with in trying to look at this from a more futuristic 23 or, like, thinking about what will be important to 24 look at when we actually have the design information 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
59 and all of the analysis that you typically see at this 1
point in time.
2 MR. HAYDEN: Thanks, Milt. At this time, 3
I think Jim Osborn gave it in his opening statement 4
that this is a framework and that the TR when it is 5
adopted in the future will obviously be supported by 6
analysis that ensures that the TR is applicable.
7 MR. CUMMINGS: I'll just add. Kris 8
Cummings from NuScale. We certainly have two designs, 9
the U.S. 460 and U.S. 600, that we have those two, a 10 6 pack and a 12 pack. As long as they meet the plant 11 equipment that are identified and you can go through 12 the thermal analysis that you have the extended coping 13 time as described in the topical report, then you 14 could apply this and say you basically rely on the 15 plant itself to be the mitigating strategy with a few 16 things that need to be done on a site or plant 17 specific basis.
18 MR. VALENTIN: And Milton again. And the 19 fact that they are saying that FLEX is not needed 20 doesn't prevent them from actually using it. Like, 21 somebody could come in and identify the need for FLEX.
22 And then they will have that resource available for 23 them. But that is something that the COL applicant 24 will have to decide when they do the actual 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
60 evaluation. Hope that helps.
1 CHAIR DIMITRIJEVIC: Okay. Any more 2
questions? All right. So shall we open for the 3
public comments? Anybody from the public wish to make 4
a comment? Please let us know.
5 MEMBER KIRCHNER: Vesna, this is Walt.
6 We're not seeing any hands raised.
7 CHAIR DIMITRIJEVIC: I was just looking at 8
that.
9 MEMBER KIRCHNER: Let me reiterate your 10 request. If anyone on the line wishes to make a 11 comment, please signal so. Speak up.
12 CHAIR DIMITRIJEVIC: Unmute yourself. All 13 right. So hearing none, I think we can go on the 14 break and then come back in closed section. So it is 15 now 2:27. So let's come back at 2:45. And we will 16 sign in on the closed section for the people who are 17 virtual connection. Okay. So see you in 17 minutes.
18 (Whereupon, the above-entitled matter went 19 off the record at 2:28 p.m.)
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- ACRS Subcommittee Meeting (Open Session) Mitigation of Beyond Design Basis Events Topical Report, PM-174186, Revision 0
LO-174187 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com ACRS Subcommittee Meeting (Open Session) Mitigation of Beyond Design Basis Events Topical Report, PM-174186, Revision 0
1 PM-174186 Rev. 0 Copyright © 2024 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)
October 1, 2024 Presenter: Peter Shaw Mitigation of Beyond Design Basis Events Topical Report
2 PM-174186 Rev. 0 Copyright © 2024 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.
This presentation was prepared as an account of work sponsored by an agency of the United States (U.S.)
Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda
- Introduction
- Overview of Plant Features
- Event Mitigation Features for Design-Basis Events
- Crediting of Safety Features for MBDBE
- Conditions for Applicability
- Summary of Plant Systems Relied Upon During Loss of AC Power
- Summary of Fulfilling Regulatory Commitments
- Conditions of Applicability (NRC)
- Conclusion
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Introduction
- Topical report establishes a consistent approach for mitigation of beyond-design-basis events (MBDBE) at a NuScale Power Plant to demonstrate compliance with 10 CFR 50.155
- Applicable to NuScale Power Plant configurations implementing design features described within the topical report o NuScale Power Plant passive safety features extend baseline coping period past traditional reactor technologies o Topical report describes how the systems are relied upon in a BDBE
- Informed approach using lessons learned from 10 CFR 50.155 guidance o NEI 06-12 B.5.b Phase 2 & 3 Submittal Guideline o NEI 12-06 Diverse and Flexible Coping Strategies (FLEX) Implementation Guide o IAEA The Fukushima Daiichi Accident Technical Volume 1/5 Description and Context of the Accident o NEI 12-02 Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
- An applicant decides how to fulfill the conditions of use (e.g. thermal analysis)
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 NuScale Reactor Module for Mitigation of Beyond-Design-Basis Events Subject of Topical
- water level and decay heat are not indicative of final plant design
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Plant Layout Overview*
Control Building Reactor Building
- Figures are representative of a typical 6 pack power plant
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Reactor Building Overview*
Overhead Heavy Load Handling System Reactor Building Walls Bioshield Spent Fuel Pool Ultimate Heat Sink
- Figures are representative of a typical 6 pack power plant
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 NuScale Power Module Overview*
Containment Vessel Decay Heat Removal System
- Figures are representative of a NuScale Power Module Emergency Core Cooling System Reactor Vent Valve Emergency Core Cooling System Reactor Recirculation Valve
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Event Mitigation Features for Design-Basis Events
- Safe and stable shutdown is achieved without operator actions o Automatic reactor trip o Containment isolation activates automatically o Decay heat removal system passively removes decay heat for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> o If AC power supply is not restored by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, emergency core cooling activates o Modules rely on no additional equipment to maintain core cooling
- Spent fuel pool shares common water source with reactor modules, thus only a single volume need be maintained for coping period
- Although not credited for safe-shutdown, system response is assured for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the main control room
- All design-basis mitigation features apply to BDBEs providing capability for successful mitigation past large light water reactor design coping periods
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Crediting of Safety Features for MBDBE
- NuScale Power Plants implement redundant and reliable safety features that mitigate core damage o Immediate plant shutdown and containment isolation at initiation of event o Containment vessels are protected by the reactor building from external events o No operator actions needed to initiate safe-shutdown
- NuScale Conditions of Use o Maintain plant specific design features described in report o Provide a thermal analysis to validate proposed coping period o Establish a maintenance rule program in accordance with 10 CFR 50.65 o Establish an emergency plan in application
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Summary of Plant Systems Relied Upon During Loss of AC Power (10 CFR 50.155(a))
- Reactor building protects modules from design-basis natural phenomena and houses the ultimate heat sink for passive core cooling
- Control building protects operators from design-basis natural phenomena
- Augmented direct current power system provides continuous DC power for equipment
- Module protection system automatically actuates safe-shutdown functions without need for operator actions
- Plant protection system monitors reactor, refueling, and spent fuel pool conditions, common to all modules
- Safety display and indication system provides nonsafety-related accident monitoring
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Summary of Plant Systems Relied Upon During Loss of AC Power (Continued)
- Containment isolation so passive safety systems can reliably initiate safe-shutdown (prevent uncontrolled release, dissipate decay heat)
- Ultimate heat sink provides common water source for maintaining safe-shutdown, assured water make-up line
- Decay heat removal system removes decay heat from shutdown reactor into ultimate heat sink
- Emergency core cooling system circulates reactor coolant into containment to initiate long term passive cooling
- Reactor building and control room building ventilation systems maintain safe atmospheres for respective environments
- Overhead heavy load handling system supports a module during lifts for design-basis natural phenomena
- Communication system used to coordinate operators and responders after a BDBE
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Summary of Compliance with Regulatory Requirements
- 10 CFR 50.155(b)(1) o Loss of AC power is identical to station blackout, safe and stable shutdown is achieved without operator actions o Spent fuel pool shares common water source with reactor modules, thus only a single volume is maintained for coping period
- 10 CFR 50.155(b)(2) and (c) o Exterior concrete walls designed for aircraft impacts and similar events
- Training requirements as defined by 10 CFR 50.155(d) o No additional training needed
- Spent fuel pool monitoring until final fuel removal from the reactor vessel, per 10 CFR 50.155(e) o Spent fuel pool monitoring in common heat sink o Instrumentation meets guidance NEI 12-02
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Limitations and Conditions (NRC)
- NRC review completed with 24 audit questions, with one docketed response and no RAIs
- NRC review identified site-specific considerations based on plant location
- These considerations are normally part of site selection, and will be documented as part of an application
- Additional Limitations and Conditions state an adopter must:
o Justify why no pre-planning is needed using the topical reports methodology for 10 CFR 50.155, if not an adopter must identify:
Water for ultimate heat sink Necessary motive equipment such as pumps and generators Equipment for debris removal Address sustainment of operators for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> post-event Address how personnel will ascertain plant conditions post-event Address how spent fuel level instrumentation power source will be replaced post-event o Provide a fire protection program in accordance with 10 CFR 50.48 o Provide a training program that includes equipment relied upon in a BDBE
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Summary
- The NuScale Power Plant design provides extended coping capability for mitigating BDBEs.
- Operator actions are not necessary to establish a safe-shutdown condition
- Topical report establishes a consistent approach for MBDBE at a NuScale Power Plant to demonstrate compliance to 10 CFR 50.155
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AC Alternating Current DC Direct Current DHRS Decay Heat Removal System BDBE Beyond-Design-Basis Event NEI Nuclear Energy Institute MBDBE Mitigation of Beyond-Design-Basis Event
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NuScale Nonproprietary Template #: 0000-21727-F01 R10 17 Questions?
Non-Proprietary Presentation to the Advisory Committee on Reactor Safeguards Subcommittee Staff Review of NuScales Power Plant Design Capability to Mitigate Beyond Design-Basis Events Defined by 10 CFR 50.155 October 1st, 2024 (Open Session) 1 NuScale TR-141299-P, Revision 1
Non-Proprietary Acronyms and Definitions AC - Alternating Current ACRS - Advisory Committee on Reactor Safeguards APLB - PRA Licensing Branch B APLC - PRA Licensing Branch C APOB - PRA Oversight Branch BDBE - Beyond Design Basis Events BDBEE - Beyond Design Basis External Events CFR - Code of Federal Regulations CR - Control Room DEX - Division of Engineering and External Hazards DC - Direct Current DNRL - Division of New and Renewed Licenses DRA - Division of Risk Assessment DSS - Division of Safety Systems EDAS - Augmented DC power system EEEB - Electrical Engineering Branch EMIB - Mechanical Engineering and Inservice Testing Branch GDC - General Design Criteria L&C - Limitation and Condition LOLA - Loss of Large Plant Area due to Explosion or Fire MBDBE - Mitigation of Beyond Design Basis Events NPM - NuScale Power Module NRLB - New Reactor Licensing Branch NRR - Office of Nuclear Reactor Regulation PRA - Probabilistic Risk Assessment SBO - Station Blackout SCPB - Containment and Plant Systems Branch SNRB - Nuclear Methods Systems and New Reactors Branch TR - Topical Report UHS - Ultimate Heat Sink 2
Non-Proprietary NuScale MBDBE Topical Report Review NuScale submitted Topical Report TR-141299-P, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155, Revision 0 on September 11, 2023, and Revision 1 on June 26, 2024.
NRC regulatory audit of the TR was performed from December 2023 to March 2024 24 audit issues were generated. 11 were resolved via audit response. 13 were resolved via Limitations and Conditions.
NuScale submitted supplemental information to address one issue raised during the audit No RAIs were issued Staff completed Topical Report review and issued an advanced safety evaluation to support todays ACRS Subcommittee meeting 3
Overview
Non-Proprietary NuScale MBDBE Topical Report Review Technical Reviewers
- Hanry Wagage, NRR/DSS/SCPB
- Raul Hernandez, NRR/DSS/SCPB
- Angelo Stubbs, NRR/DSS/SCPB
- Josh Miller, NRR/DSS/SNRB
- Ryan Nolan, NRR/DSS/SNRB
- John Hughey, NRR/DRA/APOB
- Thinh Dinh, NRR/DRA/APLB
- Marie Pohida, NRR/DRA/APLC
- Sheila Ray, NRR/DEX/EEEB
- Nick Hansing, NRR/DEX/EMIB Project Managers
- Tommy Hayden, Lead PM, NRR/DNRL/NRLB
- Getachew Tesfaye, PM, NRR/DNRL/NRLB 4
Contributors
Non-Proprietary NuScale MBDBE Topical Report Review Section 1.0 - Introduction Section 2.0 - Background Section 3.0 - Technical Evaluation 3.1 - Plant Baseline Coping Criteria for a Loss of All AC Power 3.2 - Plant Systems and Responses to a Loss of All AC Power 3.3 - Safety Functions during a Loss of All AC Power 3.4 - Capability to respond to a LOLA event, as required by 10 CFR 50.155(b)(2) 3.5 - Capacity, Capability, and protection of equipment associated with mitigation of events described in the rule, as required by 10 CFR 50.155(c) 3.6 - Training requirements as defined by 10 CFR 50.155(d) 3.7 - SFP Monitoring after final fuel removal from the reactor vessel, as required by 10 CFR 50.155(e)
Section 4.0 - Conclusion Section 5.0 - Limitations and Conditions Section 6.0 - References 5
Sections
Non-Proprietary NuScale MBDBE Topical Report Review 2.1 - Regulatory Requirements and Relevant Regulatory and Industry Guidance 2.2 - Summary of Technical Information
- The TR outlines applicability only to NuScale small modular reactor designs with specific characteristics and Conditions of Use an adopter of the TR must provide
- The TR contains description of NuScale design capabilities and features including passive safety systems capable of maintaining core cooling, containment, and spent fuel cooling functions and a large reactor pool serving as the UHS
- TR specifies how these features enable a design to mitigate BDBEs for a specified extended duration without the need for AC power, special equipment or addition guidelines and strategies.
6 Section 2.0 - Background
Non-Proprietary NuScale MBDBE Topical Report Review 3.1.1 - Assessment of Electrical Power
- During the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of a loss of all AC power is identical to an SBO and no AC power is relied upon for performing safety functions.
- The initial conditions and assumptions in NEI 12-06, Revision 4 and RG 1.226, Revision 0, assume that station batteries would remain available following a BDBE since they are considered robust.
- In the TR, section 3.1.5, Initial Event Conditions and Assumptions, NuScale states that for the baseline coping capability,
- 1) station batteries and associated DC buses remain available for the designed operating time of the station batteries, and
- 2) installed electrical distribution system, including inverters and battery chargers, remain available provided they are seismic Category I.
- Initial assumptions in the TR following a BDBEE are consistent with the NEI 12-06, Revision 4, guidance, endorsed by RG 1.226.
7 3.1 - Plant Baseline Coping Criteria for Loss of All AC Power
Non-Proprietary NuScale MBDBE Topical Report Review 3.1.1 - Assessment of Electrical Power (cont.)
- UHS monitoring, including SFP level, is assured for a timeframe in excess of the "7 days post-event" minimum in NEI 12-02.
- The capacity and capability of such multiple means, are not discussed and therefore, an applicant must satisfy L&C no. 5.2
- L&C no. 5.2: Address how plant operators will ensure, during the initial coping phase, that the following can be achieved to provide makeup to the UHS:
- Source of water identified and available in sufficient quantity
- Necessary motive equipment such as pumps and generators, and the required electrical power/fuel, can be obtained, staged, and implemented
- Any required debris removal will be accomplished to support placement of equipment and access to site connections 8
3.1 - Plant Baseline Coping Criteria for Loss of All AC Power
Non-Proprietary NuScale MBDBE Topical Report Review 3.1.2 - Plant Design Capabilities
- Following a loss of all AC power event, automatic responses of safety-related equipment establish and maintain core cooling, containment, and SFP cooling by placing the reactor modules and spent fuel into a safe, stable, shutdown state with passive cooling.
- L&C no. 5.1, an adopter must satisfy Conditions of Use in TR Section 1.3, by providing
- Plant specific design, as described within the report
- Plant specific thermal analysis (including configuration of the plant, number of modules, SFP capacity, for all modes of operation)
- A maintenance rule program IAW 10 CFR 50.65
- An emergency plan IAW 10 CFR 50.160 or 50.47(b) and App. E describing communications and coordination with local, state, federal, and tribal agencies.
9 3.1 - Plant Baseline Coping Criteria for Loss of All AC Power
Non-Proprietary NuScale MBDBE Topical Report Review 10 CFR 50.155 requires that equipment have the capacity and protection to be used to support MBDBE The TR section 4.0 lists proposed equipment for a NuScale SMR design The TR also lists the protection of this equipment The design specific and location specific nature of the MBDBE leads to needed analysis from a COL applicant. The analysis would need to confirm the assumptions in the TR remain true for the proposed plant as well as contain analysis to show that the proposed plant systems would be able to respond to the progression of the event for the proposed period of time.
L&C no. 5.1 was added to address the need for connection of analysis and the plant system responses for the event.
10 3.2 - Plant Systems and Responses to a Loss of All AC Power Event
Non-Proprietary NuScale MBDBE Topical Report Review 10 CFR 50.155 states that applicants shall develop, implement and maintain strategies and guidelines to maintain or restore core cooling, containment and spent fuel pool cooling capabilities.
The TR discusses the strategies for maintaining these functions during the event and how it is accomplished with a NuScale SMR.
The site-specific nature of hazards and potential site layouts and locations requires that a COL applicant address the potential events at their specified site.
L&C no. 5.1 address the need for an applicant to provide analysis and potential guidelines and strategies dependent on the analysis.
11 3.3 - Safety Functions during a Loss of All AC Power
Non-Proprietary NuScale MBDBE Topical Report Review The regulation requires that offsite assistance and resources be acquired indefinitely or until functional capabilities can be maintained without the need for mitigating strategies.
The endorsed guidance in NEI 12-06 recognizes that FLEX strategies and offsite resources do not need to be explicitly planned for the period beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
TR proposes that no pre-planning for mitigating actions is required because a NuScale SMR design provides BDBEE mitigating capability beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
12 3.3.2 - Indefinite maintenance of core cooling, containment, and SFP cooling capabilities
Non-Proprietary NuScale MBDBE Topical Report Review However, the guidance in NEI 12-06 also presumes that initial coping mitigating strategies have been established such that:
- Staging areas to receive offsite equipment are identified;
- Means are established to transport equipment to deployment areas;
- The ability of an offsite organization to provide support has been ensured;
- Standard FLEX equipment mechanical/electrical connectors that are compatible with site connections are obtained.
L&Cs nos. 5.2, 5.3, 5.4 and 5.5 address the requirement for long term mitigating capabilities associated with the NuScale SMR design discussed in the TR.
13 3.3.2 - Indefinite maintenance of core cooling, containment, and SFP cooling capabilities
Non-Proprietary NuScale MBDBE Topical Report Review L&C no. 5.2 specifically addresses inventory makeup to the UHS:
Identification of an available and sufficient source of water; Acquisition, staging, and implementation of necessary motive equipment; Provision of debris removal capability to deploy offsite equipment.
L&C no. 5.3 specifically addresses the assigned control room monitoring function:
Sustaining CR operators for the 72-hour period after the start of the BDBEE; Provision of debris removal to allow egress from the control room.
L&C no. 5.4 specifically addresses the capability of site support personnel to ascertain plant conditions to determine necessary coping requirements once onsite power systems are depleted.
L&C no. 5.5 specifically addresses long-term support related to SFP level monitoring instrumentation.
A provision is included in the above L&Cs to allow an applicant to provide a justification that supports not implementing the elements of that L&C.
14 3.3.2 - Indefinite maintenance of core cooling, containment, and SFP cooling capabilities
Non-Proprietary NuScale MBDBE Topical Report Review 10 CFR 50.155(b)(2) requires, in part, in regards to a LOLA, strategies and guidelines must address in a three-phase approach:
- Phase 1 - Enhanced firefighting capabilities
- Phase 2 - Measures to mitigate damage to fuel in the SFP, and
- Phase 3 - Measures to mitigate damage to fuel in the reactor vessel and to minimize radiological release The TR follows guidance in NUREG-0800, Temporary Instruction 2515/168, and NEI 06-12. For Phase 1, the TR lists design features that cope with potential LOLA events, but it did not provide plant-specific features to demonstrate adequate LOLA coping capability.
For Phase 2, the TR provides generic NuScale plant SFP design features that mitigate damage to fuel in the SFP, but it did not provide certain plant-specific details (e.g., minimum SFP water level).
15 3.4 - Capability to respond to a LOLA event, as required by 10 CFR 50.155(b)(2)
Non-Proprietary NuScale MBDBE Topical Report Review For Phase 3, the TR provides generic NuScale design features that are used to mitigate damage to fuel in the reactor vessel and minimize radiological release by plant key safety functions, but it did not provide plant-specific design features L&C no. 5.1 (for Phase 2 and 3):
- addresses plant-specific design features, including key design features, that would:
- preclude the need for an enhanced SFP mitigation capability such as a diverse or portable makeup capability to mitigate a LOLA event.
- maintain key safety functions to mitigate potential fuel damage and radiological release due to a LOLA event.
L&C no. 5.6 (for Phase 1): addresses the Fire Protection Program, including plant-specific design features and procedures, that would provide assurance of adequate LOLA coping capability.
16 3.4 - Capability to respond to a LOLA event, as required by 10 CFR 50.155(b)(2)
Non-Proprietary NuScale MBDBE Topical Report Review 10 CFR 50.155(c) requires providing the equipment relied on for the mitigation strategies and guidelines have sufficient capacity and capability to perform the intended functions and be reasonably protected from the effects of natural phenomena that are equivalent in magnitude to the phenomena assumed for developing the design basis of the facility.
The TR provides the equipment relied upon for the mitigation strategies and guidelines. The TR also describes the capacity, capability, and protection of equipment from natural phenomena.
Subject to satisfaction of L&C no. 5.1, an applicant or licensee adopting the TR for a NuScale design could demonstrate compliance with 10 CFR 50.155(c).
17 3.5 - Capacity, Capability, and protection of equipment associated with mitigation of events described in the rule, as required by 10 CFR 50.155(c)
Non-Proprietary NuScale MBDBE Topical Report Review NEI 12-06, section 11.6 describes the training to be provided to address 10 CFR 50.155(d).
The TR provided justification why specific training is not needed to address 10 CFR 50.155(d), but did not specifically addressed the required activities related to replacement of SFP level monitoring power supply. Therefore, that staff introduced L&C no. 5.7.
L&C no. 5.7 requests an adopter of the topical report to confirm that the training program includes the required activities related to replacement of SFP level monitoring power supply.
18 3.6 - Training requirements as defined by 10 CFR 50.155(d)
Non-Proprietary NuScale MBDBE Topical Report Review The staff has endorsed NEI 12-02, which describes the design criteria for a SFP monitoring system that meets the requirements of 10 CFR 50.155(e).
The TR discussed the level instruments design criteria and indicated that they meet the guidance of NEI 12-02, but it did not address the training requirements.
L&C no. 5.7 addresses the training of the operators to make the necessary connections to establish replacement power sources.
19 3.7 - SFP Monitoring after final fuel removal from the reactor vessel, as required by 10 CFR 50.155(e)
Non-Proprietary NuScale MBDBE Topical Report Review Based upon its review as discussed above, subject to the limitations and conditions as described in section 5.0 of the SE, the NRC staff concludes that TR-141299-P, Revision 1, provides a reasonable methodology for an applicant or licensee to demonstrate the NuScale plant design capability to mitigate BDBEs as defined by 10 CFR 50.155.
20 Section 4.0 - Conclusion
Attendance List for Open NuScale Subcommittee Meeting on October 1, 2024 Michael Snodderly Thomas Dashiell Shandeth Walton Meghan McCloskey (NuScale)
Erwin Laureano (NuScale)
Jim Osborn James Cordes - Court Reporter Augi Cardillo (NuScale)
Wendy Reid (NuScale Power)
Ron Ballinger Vesna B Dimitrijevic Dave Midlik Dan Lassiter (NuScale)
Larry Burkhart Iulia Jianu River Rohrman (She/Her)
Derek Widmayer Tyesha Bush Tim Polich Angelo Stubbs Robert White Sean McCloskey Robert Martin (He/Him)
Dennis Bley Sarah Horacek Thomas Hayden Gene Eckholt - NuScale Tammy Skov Alissa Neuhausen Gregory Halnon Sandra Walker Dominik Muszynski Joy Jiang Thinh Dinh Mahmoud -MJ-Jardaneh