ML24247A024
| ML24247A024 | |
| Person / Time | |
|---|---|
| Issue date: | 08/22/2024 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-0014 | |
| Download: ML24247A024 (1) | |
Text
Official Transcript of Proceedings
NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards NuScale Subcommittee: Open Session
Docket Number: (n/a)
Location: teleconference
Date: Thursday, August 22, 2024
Work Order No.: NRC-0014 Pages 1-154
NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433 1
1
2
3 4 DISCLAIMER
5
6
7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS
8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
9
10
11 The contents of this transcript of the
12 proceeding of the United States Nuclear Regulatory
13 Commission Advisory Committee on Reactor Safeguards,
14 as reported herein, is a record of the discussions
15 recorded at the meeting.
16
17 This transcript has not been reviewed,
18 corrected, and edited, and it may contain
19 inaccuracies.
20
21
22
23
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 W ASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 UNITED STATES OF AMERICA
2 NUCLEAR REGULATORY COMMISSION
3 + + + + +
4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
5 (ACRS)
6 + + + + +
7 NuSCALE SUBCOMMITTEE
8 + + + + +
9 THURSDAY
10 AUGUST 22, 2024
11 + + + + +
12 The Subcommittee met via Teleconference,
13 at 8:30 a.m. EDT, Walter L. Kirchner, Chair,
14 presiding.
15 COMMITTEE MEMBERS:
16 WALTER L. KIRCHNER, Chair
17 RONALD G. BALLINGER, Member
18 VICKI M. BIER, Member
19 VESNA B. DIMITRIJEVIC, Member
20 CRAIG A. HARRINGTON, Member
21 GREGORY H. HALNON, Member
22 ROBERT P. MARTIN, Member
23 SCOTT P. PALMTAG, Member
24 THOMAS E. ROBERTS, Member
25 MATTHEW W. SUNSERI, Member
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1 ACRS CONSULTANTS:
4 MYRON HECHT
6
7
8 DESIGNATED FEDERAL OFFICIAL:
9 MIKE SNODDERLY
10
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14
15
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1 P-R-O-C-E-E-D-I-N-G-S
2 8:30 a.m.
3 CHAIR KIRCHNER: Good morning. The
4 meeting will now come to order. This is a meeting of
5 the NuScale Committee of the Advisory Committee on
6 Reactor Safeguards.
7 I am Walt Kirchner, Chair of today's
8 Subcommittee meeting. ACRS members in attendance in
9 person are Ron Ballinger, Vicki Bier, Robert Martin,
10 and Craig Harrington. ACRS members in attendance
11 virtually via Teams are myself, Matt Sunseri, Vesna
12 Dimitrijevic, Greg Halnon, and Scott Palmtag. We also
13 have our consultants participating: Dennis Bley,
14 Charlie Brown, Myron Hecht, and Steve Schultz.
15 If I've missed anyone, please speak up.
16 I think that is the roster for today and attendance.
17 Michael Snodderly of the ACRS staff --
18 MEMBER ROBERTS: Hey, Walt, this is Tom
19 Roberts. I'm on.
20 CHAIR KIRCHNER: Okay. My oversight, Tom.
21 Thank you.
22 Michael Snodderly of the ACRS staff is the
23 Designated Federal Officer for this meeting.
24 No member conflicts of interest were
25 identified for today's meeting.
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1 During today's meeting, the Subcommittee
2 will receive a briefing on the staff's evaluation of
3 NuScale Power LLC's US460 Standard Design Approval
4 Application, Chapters 7, Instrumentation and Controls;
5 Chapter 9, Auxiliary Systems; Chapter 12, Radiation
6 Protection, and Chapter 18, Human Factors Engineering.
7 We will also be briefed on the status of
8 10 high-impact technical issues by the NuScale staff.
9 We previously reviewed the Certified
10 NuScale US600 design, as documented in our July 29,
11 2020 Letter Report on the safety aspects of the
12 NuScale small modular reactor.
13 Like the staff, we are performing a delta
14 review between the two designs, including a power
15 uprate from 50 to 77 megawatts electric per module.
16 We are reviewing these chapters as part of our
17 statutory obligation under Title 10 of the Code of
18 Federal Regulations, Part 52, Subpart (e), Section
19 141, Referral to the Advisory Committee on Reactor
20 Safeguards, and to report on those portions of the
21 application which concern safety.
22 The Subcommittee will hear presentations
23 by and hold discussions with the NRC staff and NuScale
24 regarding these matters.
25 A portion of the presentations by the
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1 Applicant and the NRC staff may be closed to discuss
2 information that is proprietary to the licensee and
3 its contractors, pursuant to 5 U.S. Code 552b(c)(4).
4 Attendance at the meeting that deals with
5 such information will be limited to NRC staff and its
6 consultants, NuScale, and those individuals and
7 organizations who have entered into an appropriate
8 confidentiality agreement with them. Consequently, we
9 will confirm that we have only eligible observers and
10 participants in the closed portion of the meeting.
11 The ACRS was established by statute and is
12 governed by the Federal Advisory Committee Act, FACA.
13 The NRC implements FACA in accordance with its
14 regulations found in Title 10 of the Code of Federal
15 Regulations, Part 7.
16 Per these regulations and the Committee's
17 Bylaws, the ACRS speaks only through its published
18 Letter Reports. We hold Subcommittee meetings to
19 gather information and perform proprietary work that
20 will support our deliberations and final decisions of
21 whether to issue a Letter Report at a full Committee
22 meeting. All member comments should be regarded,
23 therefore, as the individual opinion of the member
24 only, not a Committee position.
25 The rules for participation in all ACRS
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1 meetings, including today's, were announced in a June
2 13th, 2019, Federal Register Notice.
3 The ACRS section of the U.S. NRC public
4 website provides our Charter, Bylaws, member guidance,
5 subcommittee structure, agendas, Letter Reports, and
6 full transcripts of all full and subcommittee
7 meetings, includes slides presented there.
8 The meeting notice and agenda for this
9 meeting were posted there and can be easily found by
10 typing about us ACRS in the Search field in the upper
11 right corner of the website.
12 The ACRS, consistent with the agency's
13 value of public transparency in regulation of nuclear
14 facilities, provides opportunity for public input and
15 comment during its proceedings. We have received no
16 written statements or requests to make an oral
17 statement from the public today, but we have set aside
18 time in the agenda at the end of this meeting for any
19 comments from members of the public listening into
20 this meeting. The Subcommittee will consider all
21 public comments, as appropriate.
22 The Subcommittee will gather information;
23 analyze relevant issues and facts, and formulate
24 proposed positions and actions, as appropriate, for
25 deliberation by the full Committee.
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1 A transcript of the meeting is being kept
2 and will be made available.
3 Today's meeting is being held in person
4 and over Microsoft Teams for the ACRS staff,
5 Applicant, and members of the public. The Teams link
6 information with a telephone bridge line was placed in
7 the agenda on the ACRS public website.
8 When addressing the Subcommittee, the
9 participants should, first, identify themselves and
10 speak with sufficient clarity and volume so that they
11 may be readily heard. When not speaking, we request
12 that participants mute your computer microphone on
13 Teams, or phone, if you are on the bridge line, by
14 pressing star-6.
15 Please do not use any virtual meeting chat
16 features to conduct sidebar discussions related to the
17 presentations. Rather, limit the use of the meeting
18 chat function to report IT problems, such as inability
19 to hear speakers or see presentations.
20 Also, for everyone in the room, please put
21 all your electronic devices in silent mode, including
22 muting your speakers and microphone on your laptops.
23 In addition, please keep sidebar discussions in the
24 room to a minimum, since the microphones in the
25 ceiling are live for the course of the meeting.
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1 Finally, for the presenters, your
2 microphone at the tables, those of you there at
3 headquarters, are unidirectional. So, you'll have to
4 speak into the front of the microphone in order to be
5 heard online.
6 And before we proceed with the meeting, I
7 just want to reiterate Michael Snodderly's thanks to
8 all participants. We are spanning four time zones.
9 A special shout-out to our NuScale
10 colleagues. It's still dark here in Santa Fe, New
11 Mexico, and I think it's probably darker in Corvallis
12 at this moment. So, thank you very much for flexing
13 with us to meet our schedule requirements.
14 And with that, I'm going to turn to a
15 representative of the staff for opening comments.
16 MR. JARDANEH: Good morning, Chair
17 Kirchner, and good morning, ACRS Subcommittee members,
18 NuScale participants, NRC staff, and members of the
19 public.
20 I am MJ Jardaneh. I serve as the Branch
21 Chief of the New Reactor Licensing Branch responsible
22 for licensing the NuScale Units for design, the
23 Division of New and Renewed Licenses in NRR.
24 Thank you for the opportunity today for
25 the staff to present their reviews on select NuScale
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1 US460 Standard Design Approval Application on Chapters
2 7, 9, 12, and 18.
3 The staff is reviewing of all the chapters
4 of the SDAA concurrently. The staggered completion
5 dates are based on the complexity of the chapter and
6 the extend of change from the Certified NuScale US600
7 design.
8 Today, the staff will be presenting on
9 their review of the second group of SDAA chapters,
10 including Chapters 7, 9, 12, and 18.
11 The remaining chapters of the SDAA are
12 still being reviewed by the staff and we will inform
13 the ACRS on the Safety Evaluations when the remaining
14 chapters are available to the ACRS.
15 In today's meeting, the staff will focus
16 on the deltas from the Design Certification that the
17 NRC has approved and that the Committee reviewed in
18 the past.
19 Getachew Tesfaye, the lead Project Manager
20 for the NuScale SDAA, will provide us with update
21 about the project and walk us through the logistics of
22 the review.
23 Once again, thank you for the opportunity
24 and we look forward to a good discussion today.
25 MR. TESFAYE: Thank you, MJ.
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1 Good morning, Chair Kirchner, ACRS
2 Subcommittee members, and everyone that's
3 participating in today's meeting.
4 My name is Getachew Tesfaye. As my Branch
5 Chief MJ indicated, I'm the lead Project Manager for
6 NuScale's Standard Design Approval Application.
7 In the way of background for today's
8 meeting, NuScale completed the submittal of its
9 Standard Design Approval Application for US460, small
10 module reactor, that began in November 2022 to
11 December 31, 2022.
12 NuScale submitted the SD application
13 pursuant to the requirements of Title 10 of the Code
14 of Federal Regulations, Part 52, Subpart (e). The
15 application was formally accepted for an NRC review on
16 July 31st, 2023. Following NuScale's submittal of
17 supplemental information needed for docketing, the
18 application was accepted and we started the review.
19 On March 29 of 2024, we presented Chapters
20 2, 10, 11, 17, not including Section 17.4. The full
21 Committee deliberated on these four chapters in May.
22 About a month ago, we shared with the
23 Committee the final Draft Safety Evaluations for
24 Chapters 7, 9, and 12 that were still under management
25 review and the completed Advanced SE for Chapter 18.
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1 The three Draft SEs are still under legal
2 and management review and were officially submitted
3 and made public to support today's Subcommittee
4 meeting. When management review and legal review of
5 these chapters are completed, we will share the final
6 versions with the Committee, highlighting the changes
7 made to the drafts that are presented today.
8 We do not expect any material changes to
9 the technical content of the Draft SEs as a result of
10 review.
11 We do appreciate the ACRS's flexibility
12 for allowing us to present them at this time to
13 maintain our current schedule.
14 And I thank you again.
15 CHAIR KIRCHNER: Thank you.
16 Let me just reiterate, as we proceed, I'll
17 ask each speaker to clearly identify yourself as you
18 start your presentation, both since we have a large
19 contingent listening in today and we're in disparate
20 locations. So, just clearly identify yourself and
21 affiliation as we start each presentation. That also
22 benefits the court reporter, who is doing the
23 transcription of today's meeting.
24 With that, I believe we're ready to start
25 with Chapter 7 with NuScale. So, I'll turn to NuScale
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1 to begin.
2 MR. GRIFFITH: Good morning.
3 Thomas Griffith, Licensing Manager at
4 NuScale for the US460 Standard Design Approval
5 Application.
6 Background about myself: I have 15 years'
7 experience in the nuclear industry; worked in a
8 variety of roles, including safety analysis, reactor
9 engineering, in maintenance as an I&C shop manager,
10 and was a formerly licensed operator at Dresden Units
11 2 and 3.
12 NuScale is pleased to be presenting these
13 chapters to the ACRS today, and I would like to extend
14 my thanks to both the NRC staff for their thorough
15 review and to the NuScale staff for their efforts in
16 getting us to where we are today. There has been a
17 substantial effort by everyone involved to get to the
18 point we are, and I am thankful for all of that
19 effort.
20 At this point, I'll turn it over to Tom
21 Case to present Chapter 7.
22 MR. CASE: Good morning.
23 My name is Tom Case. I'm a Licensing
24 Engineering with NuScale. I've been with NuScale for
25 about two years and I've been in the nuclear industry
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1 for about 14 years, and I'm a Licensed Professional
2 Engineer.
3 Before discussing Chapter 7, this slide
4 acknowledges that NuScale is the recipient of
5 financial assistance awards from the U.S. Department
6 of Energy and is obliged to identify their support.
7 Next slide. I'll be presenting Chapter 7
8 of the final Safety Analysis Report from the NuScale
9 US460 Standard Design Approval Application. The scope
10 of this chapter covers safety-related I&C systems,
11 which are the module protection system and the neutron
12 monitoring system.
13 It also includes the non-safety-related,
14 non-risk-significant I&C systems that perform specific
15 regulatory required functions. These systems include
16 the module control system, plant control system, plant
17 protection system, safety display and indication
18 system, in-core instrumentation system, and radiation
19 monitoring system.
20 Next slide. Chapter 7 is divided into the
21 three sections shown here. For each section, I'll be
22 highlighting changes from the NuScale's US600 Design
23 Certification and, also, covering the results of the
24 NRC audit and review of this section.
25 Next slide. In Section 7.0, the remote
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1 shutdown station is removed because the US460 design
2 does not have a dedicated remote shutdown station.
3 Instead, there are alternate operator workstations
4 that allow for plant monitoring outside the main
5 control room. And there are no audit items or RAIs
6 specific to Section 7.0.
7 As denoted here, there is one audit item,
8 Chapter 15 review, that is related to a combined
9 license item, this section. And resolution of that
10 audit item will be through the Chapter 15 audit and
11 RAI process.
12 Next slide.
13 MEMBER HALNON: This is Greg Halnon.
14 Before you go on, could you just confirm that your
15 operating staffing connects up with this change, so
16 that you have enough operators on shift to be able to
17 do the monitoring you're expecting to be able to do?
18 MR. CASE: Yes. Can I just clarify? Is
19 that with regards to the change listed here with the
20 remote shutdown station?
21 MEMBER HALNON: Correct.
22 MR. CASE: Yes. So, basically, there's
23 alternate operator workstations outside of the main
24 control room. And in the event of a main control room
25 evacuation, the operators have the ability to monitor
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1 the plant from outside the main control room in a safe
2 shutdown condition.
3 As far as staffing goes, I'd like to
4 introduce Doug Bowman to help with the staffing part
5 of that question.
6 MR. BOWMAN: So, this is Doug Bowman,
7 Plant Operations Manager from NuScale.
8 So, a staffing question, of course, we
9 answered that with our revised staffing plan, Revised
10 Staffing Validation that we performed that was done
11 for topical report that was presented to the ACRS and
12 already has an SAR on it. And that included, the
13 testing included the remote shutdown station. So,
14 there's adequate staffing to do remote shutdown and
15 monitoring.
16 MEMBER HALNON: Okay. Thanks. Thanks.
17 Yes, Doug, just one quick follow-up. Did
18 you guys also plan on installing appropriate
19 communications, hard-wire communications between those
20 stations or are you relying on repeaters, operator, I
21 mean radio operating? Or what kind of communications
22 are you looking at?
23 MR. BOWMAN: We have specified the same
24 level of communication that we get in the control room
25 for those alternate operator workstations.
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1 MEMBER HALNON: What does that mean?
2 MR. BOWMAN: It means they get, you get,
3 it gets exactly the same communication system that you
4 would have at the main control room.
5 MEMBER HALNON: Okay. So, each of the
6 alternate operator workstations will have some ability
7 to communicate with the other operating stations and
8 some central SRO who is still in charge of the plant?
9 MR. BOWMAN: Right. The SRO, the command
10 function would move to the alternate operator
11 workstation. So, there's not going to be another
12 location.
13 Really, in terms of remote shutdown,
14 there's really not much difference from the DCA to the
15 SDA. We really didn't change the procedure very much
16 The operators take the same actions before they leave.
17 And they have the same actions they can implement
18 afterwards, if they need to, as a compensatory
19 measure. And really, the only thing that's changed is
20 we not only have a dedicated remote shutdown station,
21 we now have non-dedicated remote shutdown stations
22 that we can implement, when needed, in a control room
23 evacuation scenario.
24 MEMBER HALNON: Okay. Well, and that's
25 the point. Rather than being located in a single
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1 area, you're going to be distributed around the plant,
2 which causes a little bit more communication
3 complications, coordination, and whatnot. And that's
4 what I was just exploring, is if all of that was taken
5 into consideration with hardware, as necessary.
6 MR. BOWMAN: I understand now. I'll try
7 to get into a little more detail here and try to
8 answer that.
9 There's two designated alternate operator
10 workstations. One of them is with the module
11 maintenance center, which is inside of the reactor
12 building. One is at the rad waste control room, which
13 is in the rad waste building.
14 We wouldn't implement both of those. The
15 operators would all go to a single alternate operator
16 workstations. So, it's just a matter of having
17 choices just in case. You know, we could use this
18 alternate operator workstations in a wide range of
19 events -- from the loss of a large area to just a
20 control room evacuation event. And depending on those
21 conditions, you might pick one or the other, but
22 they're not going to be at both during this event.
23 They'll be at a single alternate operator
24 workstations.
25 MEMBER HALNON: Okay. I get it.
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1 MR. BOWMAN: Okay.
2 MEMBER HALNON: Yes, thank you.
3 MR. BOWMAN: Thank you.
4 MEMBER SUNSERI: This is Matt Sunseri. I
5 had a follow-up question.
6 CHAIR KIRCHNER: Yes, go ahead, Matt.
7 MEMBER SUNSERI: So, regarding this
8 control room evacuation, if the control room is
9 evacuated in the event of a fire, please remind me if
10 there was a way to lock out the controls in the
11 control room, so that you don't get spurious
12 actuations from the controls that are there.
13 MR. BOWMAN: Yes.
14 MEMBER SUNSERI: Is that the case?
15 MR. BOWMAN: Yes, there are.
16 MEMBER SUNSERI: Okay. Thank you.
17 MR. BROWN: This is Charlie Brown. Can I
18 make a comment? Is it the appropriate time? Walt, is
19 that all right?
20 CHAIR KIRCHNER: Yes, go ahead, Charlie.
21 MR. BROWN: Okay. On the overall
22 architecture drawing, it still shows the remote
23 shutdown station, and I presume that's the station
24 you're referring to that's not going to be in place
25 anymore. Is that correct?
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1 MR. CASE: Can you clarify which figure is
2 showing that?
3 MR. BROWN: It's 7.0-1, or something like
4 that. I forgot. Hold it. I'll have it for you here
5 in a second. I've got to scroll down to the bottom of
6 your page. It's on page 7.0-36 and the figure is,
7 let's see, 7.0-1.
8 MR. CASE: Yes, I believe that is not the
9 current revision that we're presenting today. So,
10 we're presenting Revision 1 of the US460 Standard
11 Design Approval Application. And I'm looking at
12 figure 7.0-1, which is on page 7.0-32 and it shows
13 alternate operator workstations.
14 MR. BROWN: Okay.
15 MR. CASE: And it does not --
16 MR. BROWN: Okay. All right. I accept
17 that. It's just the one I was able to get when we had
18 the information.
19 But the question I have relative to that
20 is, there were manual safety shutdown switches, so you
21 could scram the plant and initiate safeguards,
22 anything you wanted to do that you needed to operate,
23 particularly tripping the plant. Do we still have
24 manual control? Or now, you're saying it's all
25 workstations? So, all remote actuation of any
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1 function would still be done software-based as opposed
2 to any manual shutdown?
3 MR. BOWMAN: So, the remote shutdown
4 station in the DCA did not have manual actuations for
5 the safety functions. That was all done in the
6 control room or had to be done locally at the MPS
7 rooms. That's the only place they've ever had those
8 safety-related manual actuations with the new design.
9 So, that shutdown station in the DCA was nothing more
10 than a monitoring, a place for the operators to go and
11 monitor.
12 MR. BROWN: Yes, I'm just saying there was
13 a Note 9 in the other ones that said you had these
14 trip switches. So, I'm just trying to clarify, do we
15 have the ability to scram the plant outside the main
16 control room without going down to a switchboard and
17 tripping all the circuit breakers?
18 MR. BOWMAN: We have the ability to
19 locally -- we have the ability to trip the plant from
20 inside the main control room and we have the ability
21 to locally trip the plant and place it in safe
22 shutdown from the MPS room.
23 MR. BROWN: Without software? It's a
24 manual switch?
25 MR. BOWMAN: Correct.
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1 MR. BROWN: But if the main control room
2 is evacuated due to fire -- relating off to the
3 previous comment -- is that still -- in other words,
4 you're relying on somebody to not haul whatever they
5 want to haul to get it out of the main control room to
6 scram plant before they leave? Is that the point?
7 MR. BOWMAN: The operators will scram the
8 plant prior to leaving the main control room.
9 MR. BROWN: Before they evacuate?
10 MR. BOWMAN: Sure.
11 MR. BROWN: Okay. All right. Thank you.
12 MR. SNODDERLY: Excuse me, Charlie. This
13 is Mike Snodderly.
14 MR. BROWN: Yes?
15 MR. SNODDERLY: This is Mike Snodderly on
16 the ACRS staff.
17 So, Charlie, in the ACRS SharePoint folder
18 for Chapter 7 for this meeting --
19 MR. BROWN: Yes?
20 MR. SNODDERLY: -- that's Rev 1 of the
21 SDAA, is in there, if you --
22 MR. BROWN: Yes, mine said, Rev 1 on it.
23 So, I'm still a little puzzled. It's okay. I got my
24 answer.
25 MR. SNODDERLY: Okay.
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1 MR. BROWN: My version says, Rev 1, but
2 it's got the little box up in the upper left-hand
3 corner.
4 MR. SNODDERLY: So, if I could ask you, go
5 to the SharePoint site for this meeting under Chapter
6 7 and --
7 MR. BROWN: That's where I got this.
8 MR. SNODDERLY: Okay. All right.
9 MR. BROWN: But don't worry about. I
10 understand what you're saying and it's not a problem.
11 Okay?
12 MR. SNODDERLY: All right. Thank you,
13 Charlie.
14 MR. BROWN: Yes.
15 MR. SNODDERLY: All right. Bye.
16 MR. CASE: Next slide, please. In Section
17 7.1, MPS setpoints are changed as a result of the
18 changes in operating pressure and temperature. In the
19 updated safety analysis for the US460 design, some of
20 the ECCS, DHRS, and RTS actuations are changed as a
21 result of the updated Safety Analysis.
22 A timer is added to automatically initiate
23 ECCS eight hours after a reactor trip to add
24 supplemental boron, if needed to maintain
25 subcriticality during long-term cooling.
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1 And Type F post-accident monitoring
2 variables are added as a result of adoption of an
3 updated regulatory guide. Type F PAM variables
4 monitor for fuel damage.
5 There was one audit item resolved in this
6 section, and that audit item removed a note from a
7 figure to clarify that the inadvertent actuation block
8 is not applicable to the reactor bin valves.
9 Next slide. For Section 7.2, information
10 from the Sensor Technical Report cited in the DCA is
11 incorporated into this section.
12 Certain level and pressure sensors, shown
13 here, are changed from digital to analog, based on
14 additional design development of the US460 sensor
15 design. And the quantity of reactor coolant system
16 temperature sensors is reduced in each quadrant based
17 on engineering evaluation that shows streaming effects
18 do not require the use of multiple sensors per
19 quadrant.
20 There are no audit items or RAIs specific
21 to Section 7.2.
22 Next slide.
23 MEMBER ROBERTS: Yes, this is Tom Roberts.
24 If I can go back to that previous slide, the second
25 subset bullet says the D3 analysis was updated in 7.1
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1 -- presumably to reflect the conversion from the
2 digital to analog sensors for pressure.
3 Yes, my question is, the analysis change
4 appeared to take away evaluation of the pressure
5 sensor as a potential common cause failure, presumably
6 because it's an analog sensor. I guess the question,
7 is that right? And then, I was wondering why that
8 would be valid because there still could be a common
9 cause failure of the sensor because it's
10 miscalibrated; there's a common cause, or something
11 like that, that would still result in a loss of the
12 pressure function due to a common cause.
13 MR. CASE: Yes, so that is correct. The
14 diversity and defense-in-depth analysis and coping
15 analyses were updated due to the change from digital
16 to analog pressure sensors. And within that D3
17 analysis, the analysis looks for digital-based common
18 cause failures. And because the pressure sensors in
19 the US460 design were changed from digital to analog,
20 they're no longer susceptible to digital-based common
21 cause failures. Therefore, the coping analysis does
22 not evaluate digital-based common cause failures of
23 those pressure sensors.
24 With respect to, I guess what I would
25 call, an analog-based common cause failure, that would
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1 be evaluated in other analyses, such as the failure
2 modes' effects and effects analysis and the hazard
3 analysis performed on the overall system.
4 So, I think it's appropriately excluded
5 from digital-based common cause failure analysis and
6 the D3 coping analyses, but failure of those sensors
7 is addressed elsewhere in other analyses in Chapter 7.
8 MEMBER ROBERTS: Okay. Thanks.
9 You're about done with this chapter. I
10 have what I call a clarifying question, and it kind of
11 relates to the question I just asked, which has to do
12 with ATWS. Give me a second here to get to this
13 document. Okay.
14 So, the SAR Section 7.1.1, which is design
15 basis, has not changed. But it says, The design meets
16 the intent of 10 CFR 50.62 by demonstrating the
17 redundancy and diversity of the MPS design, which
18 avoids common cause failures and reduces the
19 probability of a failure to scram. And then, it
20 references Section 15.8. 15.8 says the same thing in
21 three paragraphs. So, it doesn't really add to that.
22 You issued a Technical Report back in 2013
23 on ATWS. It was entitled, NuScale Power Plant Designs
24 for ATWS and 10 CFR 50.62 Regulatory Compliance, and
25 Technical Report 2196 was the number.
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1 And it had two reasons or two rationales
2 for the ATWS approach. One of them was the diversity
3 and defense-in-depth in the MPS. The other one was
4 preliminary RELAP calculations. It showed that the
5 NuScale would meet ATWS success criteria, even if the
6 event happened. And it quoted that the consequence
7 would still maintain equivocal geometry; maintain
8 radical and pressure boundary integrity; maintain
9 containment integrity. And your SAR Chapter 19 has a
10 similar discussion in 19.2.2.
11 And so, that leaves me a little bit
12 confused as to what the technical basis is for the
13 ATWS approach. The design of diversity and defense-
14 in-depth in the MPS appears to be based on an
15 assessment of digital common cause failures, which the
16 ATWS rule predates digital I&C and was concerned about
17 more generic common cause failures.
18 And your technical basis does show that
19 the consequence of the event, if it occurred, would
20 not be called catastrophic, which would support that
21 being considered to be sufficient.
22 So, my question is, if you didn't have
23 that analysis, or if that analysis had come up with a
24 different conclusion in terms of the consequence,
25 would you still consider the diversity and defense-in-
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1 depth of the MPS to be sufficient? Or are you really
2 counting, also, on this separate study of the
3 consequence of the ATWS?
4 MR. CASE: So, I think the justification
5 provided in the SAR is what we're relying on to
6 support our evaluation of 10 CFR 50.62 as it pertains
7 to ATWS. And it specifies demonstrating the
8 redundancy and diversity of the MPS design, which is
9 digital, avoids digital-based common cause failures
10 and it reduces the probability of the failure to
11 scram. So, I think that's the justification we're
12 providing in the application and the associated
13 exemption from certain portions of that regulation
14 with respect to ATWS.
15 With respect to the history of the
16 additional analysis provided in there, I can't really
17 speculate on how we would present this without that
18 analysis, but what we are presenting here in the
19 application includes the analysis that was done.
20 MEMBER ROBERTS: Right, and I'll have a
21 similar question for the staff because they cited that
22 analysis in Chapter 19 as part of their basis for
23 accepting at least the exempt for part of the
24 requirements there. Again, I'll ask staff, but it
25 seems like they leverage that work as part of their
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1 basis for acceptance. So, it kind of leaves the
2 question.
3 And we don't need to really ask the
4 question for this plant design because the fact is you
5 have both attributes. And so, if you didn't
6 necessarily credit the diversity and defense-in-depth
7 to the extent that it was initially intended by the
8 ATWS rule, you've got the consequences as bounded.
9 And so, the degree of diversity and defense-in-depth
10 is probably more than sufficient for the ATWS rule,
11 given the consequence of the event.
12 So, I just wonder -- I'm not concerned
13 about NuScale -- my bigger concern is if there's a
14 subsequent plant design that didn't have your plant
15 characteristics and couldn't make that statement,
16 would we all come out in a different place? And so,
17 I just wanted to ask the question and I think you've
18 answered it. So, thank you.
19 MR. CASE: Next slide. So, the module
20 protection system is based on the Highly Integrated
21 Protection System Platform described in the topical
22 report shown here. This topical report was approved
23 in June of 2017. There are no changes to the topical
24 report and no changes since ACRS engagement in 2023.
25 Next slide. The Instrument Setpoint
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1 Methodology Technical Report shown here is reviewed as
2 part of Chapter 7. There have been no changes to the
3 Technical Report since previous ACRS engagement in
4 2023.
5 Next slide. With respect to combined
6 license items, or COL items, there are no changes to
7 the COL items in Chapter 7. As previously mentioned,
8 there is an audit item in Chapter 15 related to a
9 Chapter 7 COL item, and that's being resolved through
10 the Chapter 15 audit.
11 And that concludes my presentation of
12 Chapter 7. I could take any additional questions.
13 CHAIR KIRCHNER: Well, hearing none at
14 this point, Mike, we will turn to the staff, is that
15 correct?
16 MR. SNODDERLY: That is correct. So, if
17 staff could come to the --
18 MR. HECHT: This is Myron Hecht. I was on
19 mute. I'm just wondering if I can ask a question.
20 CHAIR KIRCHNER: Go ahead, Myron.
21 MR. HECHT: Okay. All of the
22 communications with both the auxiliary workstation and
23 the primary control room are done using the HIPS
24 communications modules, is that correct?
25 MR. CASE: So, can I just make sure I
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1 understand the question? It's the communication
2 between alternate operator workstations that we were
3 --
4 MR. HECHT: Right.
5 MR. CASE: -- discussing previously?
6 MR. HECHT: Right.
7 MR. CASE: Yes. So, that communication is
8 shown in figure 7.0-1 of the overall instrumentation
9 and control system architecture.
10 The communication flow path is from the
11 plant control system. So, the plant control system
12 feeds plant control system workstations and alternate
13 operator workstations. And the plant control system
14 gets its information through the safety display and
15 indication system hub, which gets its information from
16 the module protection system. So, that's kind of the
17 flow path of information from the module protection
18 system through to the alternate operator workstations.
19 I will just note that, outside of the
20 module protection system, the safety display and
21 indication system hub, the plant control system, and
22 the alternate operator workstations are all non-
23 safety-related.
24 MR. HECHT: But they do rely on the HIPS
25 communication systems in order to get that, in order
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1 to get data from the plant to the workstations, is
2 that correct?
3 MR. CASE: Yes, that's correct. The
4 information originates from the module protection
5 system, which is based on the HIPS platform.
6 MR. HECHT: Okay. Thank you.
7 CHAIR KIRCHNER: Okay. Then, with that,
8 we'll transition to the staff's evaluation of Chapter
9 7.
10 MR. VIVANCO: Good morning to the ACRS and
11 staff and members of the NRC and NuScale.
12 My name is Ricky Vivanco. I am the
13 Chapter PM for Chapter 7 of the NuScale Standard
14 Design Approval Application.
15 To provide an overview, NuScale submitted
16 Chapter 7, Instrumentation and Controls, Rev 0, on
17 December 31st, 2022, and Revision 1 on October 31st,
18 2023.
19 The NRC performed an audit of Chapter 7
20 from March 2023 to August 2023. Here, I would like to
21 note that during its review the staff recognized that
22 the I&C architecture was, essentially, not changed
23 between the US460 and the US600 design.
24 Use of the Highly Integrated Protection
25 System Platform, diverse field programmatic gamma rays
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1 technologies, the number of groups, the number of
2 divisions, data communications schemes, redundancy of
3 data buses within a division, control of access design
4 features, et cetera, in the US460 plant are the same
5 as the Certified Design.
6 Because of this, only one audit issue was
7 raised during the audit and resolved. NuScale
8 submitted supplemental information to address the
9 audit issue and no RAIs were issued as a result.
10 Joe Ashcraft and Dinesh Taneja are the
11 technical reviewers for this chapter. Again, I am the
12 responsible PM, Ricky Vivanco. Getachew Tesfaye is
13 the Project Manager for the overall SDAA review.
14 Sections of the SAR:
15 Section 7.0 is the introduction and
16 review process.
17 Section 7.1 is the fundamental design
18 principles of the I&C system.
19 Section 7.2 describes the system
20 characteristics of the I&C system.
21 To move to changes from the DCA to the
22 SDA, the NuScale power module power uprate and safety
23 analysis resulted in changes to the reactor trip
24 setpoints and engineered safety feature actuation
25 system actuation logic and analytical limits.
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1 The setpoint methodology described by TR-
2 122844-P documents the new analytical limits and
3 setpoints.
4 There is a change to the common cause
5 failure coping analysis due to the reductions in
6 digital sensors.
7 And the US460 design added Type F post-
8 accident monitoring variables.
9 Again, while there were some differences
10 between the US460 and the US600 designs, the staff
11 found that the Applicant provided sufficient
12 information to support the safety findings and that
13 all applicable regulatory requirements were addressed.
14 And that concludes the presentation for
15 Chapter 7 from the NRC staff.
16 MEMBER ROBERTS: Yes, this is Tom Roberts.
17 I had the same clarification question on
18 the section that didn't change from the previous one.
19 I just want to understand what your basis is.
20 Pretty deep into Section -- let me find
21 this -- I think it was page 7-63 in the SAR talks
22 about your basis for accepting the ATWS approach. And
23 it leverages pretty heavily the sensitivity study that
24 the Applicant documents in Chapter 19 that shows that,
25 for the more frequent events that could happen,
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1 complicated by an ATWS, the plant would take care of
2 itself. And so, there's probably half a page on that,
3 and then, you ended that, Based on this evaluation,
4 the NRC staff finds the Applicant's assertion above to
5 be reasonable. And then, you document a study the NRC
6 did to confirm the analysis that was in Chapter 19.
7 So, it seems like, from the staff's
8 perspective, an important part of the basis for
9 accepting the approach to ATWS is the fact that the
10 plant performance is not that bad; that as long as you
11 have what would be characterized as an anticipated
12 event, the plant will take care of itself. And so,
13 the design of the I&C system certainly should be
14 diverse, but that's kind of, you know, either
15 secondary or ancillary to that.
16 So, I made a comment on that. Did I
17 misunderstand what your intent was for that part of
18 the discussion or is that really part of the basis for
19 acceptance?
20 MR. TANEJA: Yes, this is Dinesh Taneja.
21 So, the requirement of the 50.62 of having
22 a diverse means of achieving reactor trip and some of
23 these other support functions, aux feedwater, which
24 doesn't apply to NuScale, and turbine trip -- so, when
25 we looked at the protection system design, it has
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1 diversity. So, that was a big review basis for that.
2 And also, in addition to that, the
3 probability of achieving these functions, you know, so
4 the regulatory requirements are that it needs to be
5 better than 10 to the minus 5, I believe. And that's
6 where the independent assessment was done by our PRA
7 folks to assure that, yes, it also meets that
8 requirement as well, achieving that function. So, I
9 think that was the basis for approving this exemption
10 request.
11 So, you know, it hasn't really changed at
12 all from our DCA to the SDA. It's those features are
13 still the same. They still have the same exact
14 architecture for the HIPS platform and the
15 arrangements. You know, it's using the same type of
16 setup here. It's using a redundant setup of the
17 equipment interface modules and the reactor trip
18 breakers are aligned. So, it really is having the
19 diverse FPGAs in the different divisions. So, it's
20 not susceptible any kind of common cause failure and
21 the reliability is pretty high.
22 MEMBER ROBERTS: Yes, thanks, Dinesh. I
23 appreciate all that. And the question I'm asking
24 would be hypothetical and it's probably not worth
25 spending more time in this meeting talking about it.
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1 But if you had a design where you didn't
2 have that assessment in Chapter 19, or if the
3 assessment showed completely opposite results, that if
4 you had the event, the ATWS event, that the
5 consequences would be really bad, you know, it seems
6 like, from the way you've written the SER, that would
7 be something that would certainly be a consideration
8 in terms of whether it's acceptable, given that you've
9 written that as part of the basis for acceptability.
10 If they did the side study which you validated, that
11 if the event were to happen, that it wouldn't be that
12 bad, and that's what --
13 MR. TANEJA: Yes, but, you know, the thing
14 is, when we looked at that rule 50.62, it was really
15 specific to the reactor designs of the time. So, they
16 were talking about the PWRs by Westinghouse and CE and
17 their design features, and the BWRs. So, how do you
18 diverse means of achieving that function was needed.
19 And I think it resulted from some events
20 that happened at Salem, is that correct?
21 MR. VIVANCO: Yes, that is correct.
22 MR. TANEJA: So, like now, if you think
23 about it from that point of view, that what if I do
24 have a scenario where I have a transient without a
25 trip, how does a plant cope with it?
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1 So, you know, now we are looking at
2 passive plants that have really a margin of safety and
3 a margin of time that's there. So, all these factors
4 really play into you achieving the safe shutdown
5 condition or the ability to trip the plant, and there
6 is a lot of time available, as opposed to the large
7 light water reactors that we have had experience with.
8 So, you know, yes, it is a good way of
9 asking this question, how do we cope with it, but,
10 then, adequacy of defense-in-depth, I think we are
11 looking at all the advanced non-light water reactors.
12 Our safety goals are still to see that there is
13 adequate defense-in-depth in achieving these safety
14 functions. So, really, the fundamentals are still
15 there. I don't think we are going away from that.
16 MEMBER ROBERTS: Yes, thanks, Dinesh. I
17 think that answers the question.
18 Again, it would be hypothetical at this
19 point to look at a plant design where a side study
20 showed that the consequences of the event were really
21 bad. Then, you would have to think about, okay, is
22 the adequacy of either the plant design or the
23 diversity in the trip system sufficient?
24 But, for NuScale, there's a pretty solid
25 story, and you cite that in the Safety Evaluation,
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1 that there's kind of this two prongs of good diversity
2 in the trip system, and if the event were to happen,
3 the plant mitigates a consequence without the trip
4 system, which it seems to me that both have a role.
5 And that's really the point I wanted to make. And it
6 sounds you're saying the same thing.
7 So, thank you.
8 MR. TANEJA: Yes.
9 MR. VIVANCO: Are there any additional
10 questions on Chapter 7?
11 (No response.)
12 MR. VIVANCO: Thank you all again for your
13 time.
14 CHAIR KIRCHNER: Thank you. Okay. With
15 that, then, I think we should move on to Chapter 9.
16 We'll begin, again, with NuScale, and then we'll hear
17 from the staff. And depending on how much time that
18 takes, we'll decide on taking a break. But let's
19 proceed to Chapter 9 with NuScale's presentation.
20 MS. TURMERO: Good morning.
21 My name is Sarah Turmero and I'm a
22 Licensing Engineer with NuScale Power. I've been with
23 the company for about two years and I was previously
24 a reactor engineer at Waterford 3.
25 MS. AHMED: Good morning.
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1 My name is Freeda Ahmed. I've been with
2 NuScale as a Licensing Engineer and I've been with
3 NuScale for a little over two-and-a-half years. I
4 have a degree in nuclear engineering and I've had over
5 a decade of experience in the industry.
6 MR. GREEN: Good morning.
7 My name is Jordan Green. I'm a
8 professional engineer licensed in Texas. I've been
9 the with the Plant Systems Engineering Mechanical
10 Group at NuScale for about two years. Before that, I
11 was a systems engineer and programs engineer for about
12 11 years at South Texas Project.
13 MS. TURMERO: So, Chapter 9 is our
14 auxiliary systems, and it's systems including the fuel
15 storage handling, water systems, process auxiliary
16 systems, air conditioning, heating and ventilation,
17 and other auxiliary systems such as the lighting,
18 communication systems, fire protection, and the fire
19 hazards analysis.
20 Next slide. For Section 9.1, the focus is
21 on fuel storage and handling, which includes the
22 criticality of fuel storage and handling; the pool
23 cooling and cleanup system; the fuel handling
24 equipment, and the overhead heavy load handling
25 system.
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1 For Section 9.1.1, the criticality safety
2 of fresh and spent fuel storage and handling, the
3 criticality safety of fuel storage is addressed by COL
4 Item 9.1-1, which is shown on the slide.
5 There were three audit questions. One
6 focused on the design of the reactor flange tool and
7 the criticality of the fuel while the reactor pressure
8 vessel is in the reactor flange tool, and then, other
9 audit questions were on the topic of pool criticality.
10 Section 9.1.2, the new and spent fuel
11 storage. The storage is addressed by COL Item 9.1-2,
12 which is shown on this slide. There were two RAI
13 questions on the topic of pool inventory.
14 Section 9.1.3, pool cooling and cleanup
15 system. The systems described in the DCA were
16 consolidated as subsystems for the overarching pool
17 cooling and cleanup system. The major components in
18 subsystem remains the same as the DCA with the
19 exception of the pumps and heat exchangers, which were
20 reduced five trains to three trains.
21 And there were no specific audit or RAI
22 questions for Section 9.1.3.
23 For 9.1.4, the fuel handing system, the
24 design changes from the DCA, the new fuel elevator is
25 now capable of handling irradiated fuel assemblies for
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1 inspection purposes. There are vertical travel
2 limits. That ensures adequate shielding of the spent
3 fuel assemblies.
4 And the new fuel jib crane classification
5 changed from ASME NUM-1 Type 2 to ASME NUM-1 Type 1A
6 with single failure-proof features.
7 And there were no specific audit questions
8 to Section 9.1.4.
9 For the overhead heavy load handling
10 system, the design changes include an increase of the
11 reactor building crane capacity. The module-lifting
12 adapter from the DCA was removed and is now integral
13 to the crane itself.
14 The reactor building crane auxiliary hoist
15 capacity increased. There were changes to the reactor
16 building crane automated control system software that
17 reduces probability of operator error.
18 Additional jib cranes were added to the
19 overhead heavy load handling system and they are
20 designed to ASME NUM-1 Type 1A.
21 And the heavy load exclusion zone above
22 the spent fuel pool was removed.
23 There was one change to the COL Item 9.1-5
24 in relation to the heavy load exclusion zone. That
25 wording was replaced with safe load paths.
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1 MEMBER BALLINGER: This is Ron Ballinger.
2 We looked at this a very long time ago and
3 we had a long discussion about the heavy load, the
4 crane system itself. And I asked the question then,
5 but I don't remember the answer. With respect to the
6 safe load paths, are there mechanical stops for the
7 crane as opposed to electronic interlocks that would
8 prevent the crane from deviating?
9 MS. TURMERO: There are not mechanical
10 stops.
11 MEMBER BALLINGER: There are no mechanical
12 stops?
13 MS. TURMERO: That's correct. There are
14 physical stops at end of travel, but nothing that
15 would prevent going over the spent fuel pool
16 physically.
17 MEMBER BALLINGER: Thank you.
18 DR. SCHULTZ: This is Steve Schultz.
19 This is a particular area where you have
20 made changes that significantly affect the safety
21 related to the operations of the crane, the capability
22 of the crane, and so forth.
23 With regard to the COL item, can you
24 describe the type of guidance that is provided by
25 NuScale to the applicant to help them move through
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1 what you're asking them to do in the COL stage?
2 MS. TURMERO: For COL Item 9.1-5?
3 DR. SCHULTZ: Yes.
4 MS. TURMERO: We don't prescribe specific
5 standards to be followed. However, we do anticipate
6 that the zero applicant would follow Reg Guide 1.244,
7 and that would be our recommendation for this COL
8 item.
9 DR. SCHULTZ: Thank you.
10 MEMBER SUNSERI: This is Matt. I had a
11 question here also. Matt Sunseri.
12 You made changes to the automated control
13 software to reduce the probability of operator error.
14 And as I recall on the previous version, operator
15 error was a significant contributor to the risk of the
16 station. So, in light of that, are there any
17 limitations on whether or not the automated controls
18 can be in override and the operator can take manual
19 control, the operator and reintroduce those risks, or
20 are there going to be administrative controls to
21 prohibit or otherwise limit operator manual
22 engagement?
23 MS. TURMERO: There are administrative
24 limits. So, there's an override key that would be
25 required and that would be administratively
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1 controlled.
2 MEMBER SUNSERI: And I'm presuming that
3 use of that administrative control would require
4 additional oversight or something of that nature to
5 mitigate the incurred risk?
6 MS. TURMERO: That's correct. That would
7 be part of the operating and maintenance procedures in
8 zero Item 9.1-5.
9 MEMBER SUNSERI: Okay. Thank you.
10 MS. TURMERO: Next slide, please. For the
11 audit and RAI results for Section 9.1.5, there were
12 three audit questions on the use of ASME NUM-1,
13 including demonstrating a compliance with Regulatory
14 Guide 1.244, position C.1; one RAI on the deviations
15 taken on ASME NUM-1 for the reactor building crane,
16 and one RAI on the elimination of the heavy load
17 exclusion zone terminology.
18 And that's all I have for Section 9.1.
19 Before I turn it over to Freeda, are there any
20 additional questions?
21 (No response.)
22 MS. TURMERO: Thank you.
23 MS. AHMED: Section 9.2 is water systems,
24 and it consists of nine separate sections. The
25 systems that are in bold text are the systems that
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1 have had changes from the DCA. All the systems that
2 are not in bold have no major design changes. I will
3 only be discussing the bolded systems, which is the
4 potable and sanitary water system; ultimate heat sink;
5 site cooling water system, and utility water system.
6 However, I would be happy to answer any questions
7 about any of the systems in Section 9.2.
8 And I would also like to point out that,
9 other than the ultimate heat sink, all the systems are
10 non-safety, non-significant systems.
11 Next slide, please. Section 9.2.4 is the
12 potable and sanitary water systems and we had one
13 change, which was that the potable and sanitary water
14 systems piping, including the loop seals penetrating
15 the control room envelope change from a Seismic
16 Category II to a Seismic Class I. And the reason for
17 this change was to ensure that the loop seals at the
18 penetration could remain intact, thereby maintaining
19 the leak tightness of the control room on load after
20 a seismic event.
21 In addition to this change, we also
22 removed two COL items. Because of the potable water
23 system design, the Reg Guide criteria 1.206 and NUREG-
24 800 are no longer applicable and all the general
25 design criteria is included in the SDA. And we were
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1 able to successfully resolve one audit item.
2 Next slide, please. All right. Section
3 9.2.5 is the ultimate heat sink and we had a couple of
4 changes from the DCA.
5 The first change would be that we had a
6 smaller footprint of the UHS and the reason for this
7 was that we went from 12 modules to 6 modules, so the
8 lower inventory; and also, the UHS level was lowered
9 from 68 feet to 53 feet from the bottom of the module.
10 And we had no audit or RAI questions
11 specific to this section.
12 MEMBER SUNSERI: So, this is Matt. I have
13 a question about this also.
14 I mean, I understand reducing the
15 footprint and the boundary of the pool, but why would
16 you lower the depth of the pool? I mean, it seems
17 like that's giving away margin that would be easy to
18 maintain.
19 Did you understand the question?
20 MS. AHMED: Could you just clarify one
21 more time, please?
22 MEMBER SUNSERI: So, I understand why the
23 total inventory of the UHS might go down because of
24 the smaller footprint, but I don't understand why you
25 would choose to lower the level from 68 feet to 53 in
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1 the remaining footprint, because that appears to me to
2 be margin, giving away margin that is fairly cheap to
3 maintain, unless I'm missing something.
4 MS. TURMERO: This is Sarah Turmero.
5 So, the pool level was reduced to help
6 with the operation of DHRS. So, there was an issue of
7 overcooling. And so, the lower level helps with the
8 performance of DHRS.
9 MEMBER SUNSERI: Oh, okay. All right.
10 Thank you.
11 CHAIR KIRCHNER: This is Walt.
12 Could you elaborate on that last point
13 about the DHRS? You suggested that there was an
14 overcooling problem.
15 MS. SWANSON: Hi. This is Mara Swanson,
16 NuScale Power. I'm an engineer in the Mechanical
17 Systems Group.
18 CHAIR KIRCHNER: Yes?
19 MS. SWANSON: Yes. I'm sorry, did you
20 want to finish?
21 CHAIR KIRCHNER: No, go ahead, please.
22 MS. SWANSON: Yes. The lowering of the
23 pool level was a result of overcooling concerns. So,
24 during certain accident scenarios with an ECCS and
25 DHRS actuation, the containment module would cool down
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1 too quickly, compromising the integrity of the
2 containment module over long periods of time through
3 lots of thermal cycling. So, the reduction of the
4 pool level decreases the thermal stresses on the
5 containment module.
6 CHAIR KIRCHNER: Thank you.
7 What I'm struggling with here, quickly, is
8 the DHRS, basically, is immersed in the pool. Are you
9 suggesting that going from 68 feet to 53 feet keeps
10 some of it above the water level line?
11 I mean, pretty much, that heat exchanger
12 affixed to the external part of the containment vessel
13 will see a uniform temperature. So, if it's immersed,
14 I'm not following how it could result in an
15 overcooling transient. Your normal cooldown mechanism
16 would not be using DHRS, would it?
17 MS. SWANSON: No, this scenario does not
18 apply to a normal cooldown, and you're correct, the
19 DHRS do remain completely immersed for a minimum of 72
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> during accident scenarios. So, it is for the
21 ECCS system actuating, which involves the entire
22 exposed surface area of the containment system. So,
23 it's reducing the surface area of the UHS and the
24 containment system itself, which is the conductive
25 cooling that ECCS uses to --
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1 CHAIR KIRCHNER: Correct.
2 MS. SWANSON: -- cool the primary fluid.
3 CHAIR KIRCHNER: But that would be an
4 unusual event, for your ECCS system to be actuated.
5 MS. SWANSON: That's correct, that
6 scenario applies only to very specific cooldowns.
7 CHAIR KIRCHNER: So, from what you've
8 provided, like Matt, my concern is -- it's not a
9 concern; it's just it seems to me you would have more
10 margin.
11 So, the vessel, could you just kind of
12 give us a feel for how much of the vessel, then, is
13 immersed? The vessel, the containment vessel, is,
14 roughly, 85-feet high -- I'm doing this from memory,
15 which is not a good way --
16 MEMBER SUNSERI: I think he's got it
17 there. I think it's on the chart there. The vessel
18 would be 53-feet deep instead of 68-feet deep.
19 CHAIR KIRCHNER: Yes. Yes. Yes,
20 essentially, roughly two-thirds of the vessel is
21 immersed in water.
22 MS. SWANSON: That's correct, yes.
23 CHAIR KIRCHNER: Is part of the rationale
24 just thermal performance; that if it's entirely
25 immersed or more immersed at 68 feet, obviously,
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1 another 15 feet of water, that you're impacting your
2 thermal efficiency? Well, you would have more heat
3 loss to the pool during normal operation if it's more
4 fully immersed.
5 I'm just trying to understand what design
6 objective was being, what design objective was reached
7 by reducing the water level.
8 MS. SWANSON: It was, yes, the design
9 objective is to prevent the containment vessel failure
10 over long periods of thermal cycling.
11 Regarding the other points you brought up,
12 the thermal losses during normal operation, because
13 our containment is kept at a vacuum, is pretty
14 minimal. So, those --
15 CHAIR KIRCHNER: Pretty minimal? Okay.
16 MS. SWANSON: That parameter is largely
17 unchanged from the DCA design.
18 CHAIR KIRCHNER: I see. All right.
19 MEMBER SUNSERI: So, maybe this is a
20 question, then. I presume that, with the lower
21 inventory in the pool, the water in the UHS is going
22 to be warmer than it would be at 68 feet. Is that a
23 fair assumption?
24 MS. SWANSON: The normal operating
25 temperature is unchanged from 100 degree Fahrenheit.
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1 During refueling operations, we did increase the
2 maximum allowable pool temperature to 120 degree
3 Fahrenheit.
4 MEMBER SUNSERI: So, the impact on the
5 containment vessel is merely the temperature that
6 metal would get based on only having 53 feet of it in
7 the water versus 68 feet in the water. Is that --
8 MS. SWANSON: Yes, the rate of cooling
9 during an ECCS actuation is that, yes.
10 MEMBER SUNSERI: Okay. All right.
11 Thanks.
12 MEMBER HARRINGTON: This is Craig
13 Harrington.
14 Not to belabor the point, but it seems
15 like that portion of the vessel that's below water
16 level, the temperature isn't really going to be
17 different. So, how does that change the thermal
18 cycling that I think you said you were attributing to
19 essential failure over a number of cycles?
20 MS. SWANSON: I'm sorry, could you repeat
21 the question? I'm not sure I understood.
22 MEMBER HARRINGTON: Well, by reducing the
23 water level, the portion of the vessel out above the
24 water level, that temperature may change, be
25 different, but the part that's below water, it seems
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1 as it's immersed, would still be the same. So, I'm
2 trying to figure out where the difference in thermal
3 cycling that could lead to some vessel failure comes
4 from.
5 MS. SWANSON: The concern is less with the
6 temperature reached by the vessel and more related to
7 the rate of change at which it cools. So, the
8 increased surface area from a higher water level would
9 cool the vessel down much more quickly than the
10 reduced water level.
11 MEMBER HARRINGTON: It just seems like a
12 bulk wall average temperature argument, as opposed to
13 local temperatures below water and above water. Maybe
14 I'm just missing it.
15 MS. SWANSON: Well, I think, for this
16 level of detail, I'd need to defer this question until
17 after the presentation. I can't speak to the
18 temperature profile over the whole course of the
19 cooling scenario.
20 MEMBER HARRINGTON: Okay. Thanks.
21 MEMBER SUNSERI: But you have an analysis
22 that supports the change that will occur, is that
23 correct?
24 MS. SWANSON: That's correct, yes.
25 MEMBER SUNSERI: Okay. Thanks.
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1 MS. SWANSON: I see, in talking to some of
2 the other staff, I think that this question at this
3 level of detail is probably best addressed in the
4 closed session.
5 MEMBER SUNSERI: Okay. We'll try to
6 remember to ask then.
7 MS. SWANSON: Okay. Thank you.
8 MS. AHMED: Section 9.2.7 is the site
9 cooling waters. And we only had one design change
10 from the DCA. And that was we went from an open-loop
11 system design in the DCA to a two-loop closed hybrid
12 system in the SDA. And the reason for this is the
13 design change was to better maintain the water quality
14 for plant users.
15 We also removed two COL items, and the
16 reason we removed the COL items is that, in the
17 standard design, all (audio interference) system
18 functions. And that will manage corrosion and
19 fouling. And thus, there will no longer be a COL.
20 And we were able to successfully resolve on audit
21 question.
22 Next slide, please. Section 9.2.9 is the
23 utility water systems. And we had a change from the
24 DCA, and that was the removal of COL Item 9.2-5, which
25 concerns the identification of the site-specific water
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1 source.
2 And the water treatment system was
3 removed. And it was removed because there are no
4 regulation requirements that pertain to the COL item,
5 and the information is included in the SDA.
6 Additionally, we received two audit
7 questions that we were able to successfully resolve.
8 Next slide, please. Section 9.3 is
9 process auxiliary systems, and it consists of seven
10 different sections. However, I will only be talking
11 about the chemical and volume control system and the
12 containment evacuation system, as they are the systems
13 that had changed from the DCA.
14 I would also like to point out that, other
15 than the DCS, all systems in this section are non-
16 safety, non-risk-significant systems.
17 Next slide, please. Okay. Section 9.3.4
18 is chemical and volume control systems. And we had
19 one change from the DCA and that was we removed -- the
20 module heatup system was modified to use an electric
21 heater in lieu of two heat exchangers that were
22 providing steam from the auxiliary boiler.
23 And we were able to successfully resolve
24 one audit item and three RAIs.
25 Next slide, please.
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1 DR. SCHULTZ: Could you go back to that
2 slide just for a moment, please?
3 Looking at 9.3.4, the additional design
4 information in the last bullet there associated with
5 flow-restricting venturis supporting probabilistic
6 risk assessment, could you expand on that? What was
7 provided and how did it affect the PRA? What was the
8 question associated with that? How was the PRA
9 affected as a result of the discussion?
10 MR. GRIFFITH: Thomas Griffith, Licensing
11 Manager, NuScale.
12 So, what was added into Section 9.3.4 here
13 was specific to the actual descriptions of the
14 venturis, that the actual venturis themselves were not
15 described as well as they could have been. So, we
16 provided additional clarification in Section 9 related
17 to the location of those components and a description
18 of those components.
19 The venturis are operated in the PRA for
20 line breaks where the venturis are installed, and they
21 do provide a restriction for flow in those instances.
22 DR. SCHULTZ: That's going to affect the
23 analyses that you perform, the heat associated with
24 the LOCA and break, and so forth?
25 MR. GRIFFITH: So, it affects the small
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1 line break analysis.
2 DR. SCHULTZ: Thank you.
3 MEMBER DIMITRIJEVIC: Hi. This is Vesna
4 Dimitrijevic.
5 This is for the LOCA's outside
6 containment, right, line breaks for the LOCA's outside
7 containment? That's where those restrictive things
8 are installed? It's my question.
9 MR. GRIFFITH: This is Thomas Griffith,
10 the Licensing Manager.
11 That is correct. It is for the small line
12 breaks outside of containment.
13 MEMBER DIMITRIJEVIC: And did that change,
14 did they change the likelihood of those events?
15 MR. GRIFFITH: They did not.
16 MEMBER DIMITRIJEVIC: I mean did they
17 think --
18 MR. GRIFFITH: Their impact is in the
19 thermal-hydraulic analysis, not on the likelihood of
20 an event.
21 MEMBER DIMITRIJEVIC: Okay.
22 MEMBER SUNSERI: This is Matt Sunseri.
23 These were installed as a result of the
24 decision to remove the check valves in the line?
25 That's my question.
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1 MR. GREEN: This is Jordan Green.
2 So, the venturis and the two CIVs, and
3 then, the third isolation valve, those are all
4 considered containment system components. So, we're
5 going to have more information on those in Chapter 5
6 -- Chapter 6, I'm sorry, 15 and 19.
7 The valves moved out of the nuclear power
8 module bay are the CVCS valves and they were done so
9 in support siting considerations, maintenance
10 considerations, and reducing congestion. So, I think
11 there is only one excess flow check valve, I think is
12 the one that you're alluding to, and that was replaced
13 with an AOV. And so, these materials were not added
14 to compensate for that.
15 MEMBER SUNSERI: Okay. All right. Thank
16 you.
17 CHAIR KIRCHNER: So, just for
18 clarification -- this is Walt Kirchner -- we'll
19 discuss this particular subject when we cover Chapter
20 6? Or later today in the closed session?
21 MR. GREEN: All the information on the
22 venturis, their classification, their location, is
23 included in Chapter 6 in the containment section.
24 CHAIR KIRCHNER: Okay. All right. Thank
25 you.
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1 MS. AHMED: Section 9.3.6 is the
2 containment evacuation system. And we had a couple of
3 changes.
4 The first change is that the CES inlet
5 pressure instrumentation and connecting piping tube,
6 up to and including the isolation valves, is designed
7 to Seismic Category I standards. It was SC-III in the
8 DCA. And this ensures that these components maintain
9 capability to perform their function during and after
10 safe shutdown only.
11 And the second change that happened is
12 that same section of piping mentioned was increased to
13 containment design pressure. And it was to act as a
14 diverse, independent backup to CIVs in support of PRA.
15 And we were able to successfully resolve three audit
16 questions.
17 Next slide, please. Section 9.4 is air
18 conditioning, heating, cooling, and ventilation
19 systems. And it includes five separate sections.
20 None of the sections are bolded because none of them
21 had design changes. However, I would like to say that
22 they had no design changes, but the physical size of
23 the ventilation systems was altered.
24 And then, also, we removed four COL items,
25 COL Items 9.4-1 through 9.4-4. And these COL items
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1 were in regard to the need for periodic testing and
2 inspection requirements, and they were incorporated
3 into the SDA, and thus, no longer needed as COL items.
4 And Section 9.5 is other auxiliary
5 systems, which contains the lighting systems,
6 communication systems, and fire protection. All three
7 systems had some changes from the DCA. So, I will be
8 discussing all three.
9 Next slide, please. Section 9.5.3 is
10 lighting systems. And the main change from the DCA
11 was that the main central room has a dedicated
12 emergency lighting that is continuously on. It had an
13 auto-transfer in the DCA, but now in the SDA it is
14 continuously on. And we were able to successfully
15 resolve two audit items -- two RAIs and one audit
16 item.
17 Next slide, please. And Section 9.5.2 is
18 the communication system. And the changes in that
19 were that the sound-powered telephone system was
20 removed. The health physics network was added to the
21 communication system, and we removed COL Item 9.5-2
22 because it is now part of the Standard Design
23 Application and not needed for a COL item.
24 And there were no audit or RAIs specific
25 to this section.
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1 Next slide, please. And I will now turn
2 it over to Jordan to discuss Section 9.5.1, fire
3 protection and 9A.
4 MR. GREEN: Thanks, Freeda. Section 9.5.1
5 is the fire protection program and reflects the
6 building layout change. We successfully resolved
7 three audit questions.
8 Next slide, please.
9 DR. SCHULTZ: Jordan, this is Steve
10 Schultz.
11 Could you just expand on your last bullet
12 there, that you discussed structural and electrical
13 raceway fire barrier requirements? What was the topic
14 of discussion and what were the results of the
15 discussion that you had?
16 MR. GREEN: As I recall, it was just a
17 reiteration that the raceway design meets requirements
18 of ASME 119 and FPA 251 and is the responsibility of
19 the Applicant. and per our Reg Guide 1.189 table
20 9.5.1-2.
21 DR. SCHULTZ: Thank you.
22 MR. GREEN: Section 9A is the fire hazards
23 analysis. And again, this reflects the building
24 layout change.
25 We successfully resolved one audit
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1 question and one RAI -- with that last RAI being
2 pretty extensive. We went through the fire hazards
3 analysis, and then, explicitly stated each fire area
4 that did not contain safe shutdown equipment and how
5 propagation is mitigated.
6 And that concludes our presentation on
7 Chapter 9. If there are any additional questions,
8 we'll take those now.
9 CHAIR KIRCHNER: Members and Consultants,
10 any further questions of NuScale on Chapter 9?
11 MEMBER SUNSERI: Walt, this is Matt. I
12 don't have any.
13 CHAIR KIRCHNER: Okay. Thank you, Matt.
14 Hearing none, I see it's coming up on
15 10:00 Eastern time. While we transition to the staff,
16 let's take a break until 10:10 a.m. Eastern time, and
17 then, we'll pick up with the staff's evaluation of
18 Chapter 9.
19 So, we are in recess until 10:10 Eastern
20 time.
21 (Whereupon, the above-entitled matter went
22 off the record at 9:56 a.m. and resumed at 10:10 a.m.)
23 CHAIR KIRCHNER: Okay, we are back in
24 session and we are turning, next, to the staff's
25 presentation their evaluation of Chapter 9, Auxiliary
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1 Systems.
2 And I'll turn it to the staff.
3 MR. CRANSTON: Good morning. My name is
4 Greg Cranston. I'm a Project Manager on the NuScale
5 project, here for Chapter 9, Auxiliary Systems. I'm
6 with the Division of New and Renewed Licenses,
7 Licensing and Regulatory Infrastructure Branch.
8 Next slide, please. In conjunction with
9 the overview, NuScale submitted Chapter 9, Rev. 0,
10 SDAA FSAR on December 31st and Revision 1 on October
11 31st, 2023.
12 The NRC regulatory audit of Chapter 9 was
13 performed in March 2023 through August 2023,
14 generating 33 audit items.
15 When NuScale went through their
16 presentation, they did identify those audit items and
17 what systems they were associated with and, also,
18 associated RAIs.
19 There were 10 audit issues opened,
20 resulting in the NRC submitting supplemental
21 information, and 13 RAIs were issued. All those
22 issues have been resolved, and as a result, we do have
23 four confirmatory items were, when the FSAR is
24 updated, those confirmatory items will close to
25 incorporate the feedback we got to resolve the audit
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1 issues and the RAIs.
2 Next slide, please. As was pointed out by
3 NuScale, this auxiliary systems, there's quite a lot
4 of differences to expand. I'm certainly not going to
5 talk about them all today, either.
6 So, the number of contributors is probably
7 quite a bit more than most other chapters. So, it was
8 quite involved; a lot of people reviewed, basically,
9 the 29 systems associated with the auxiliary systems.
10 Next slide, please. As mentioned
11 previously, the main areas are fuel storage and
12 handling; water systems; process auxiliaries, HVAC,
13 and auxiliary systems. And then, Appendix 9 covers
14 the fire hazards analysis.
15 Next slide. In conjunction with some of
16 the significant items that the staff chose to
17 identify, many of these, again, were previously
18 discussed in the NuScale presentation.
19 On the fire barrier rating, there was a
20 change from some areas being two-hour rates to now
21 three-hour rated. And in both cases, the ratings were
22 within code, whether it was associated with the DCA or
23 the SDA.
24 And with the US460, there are only six
25 nuclear power modules rather than 12 in the DCA. And
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1 as was discussed previously, the related parameters
2 used in the analysis of the spent fuel pool and
3 ultimate heat sink changed accordingly.
4 The reactor building crane capacity was
5 increased to better handle the modules and a dry dock
6 jib crane was added for refueling.
7 Next slide. The design included
8 consolidating the spent fuel pool cooling and cleanup
9 systems and combined the ultimate heat sink cooling
10 systems into a single system that cools both the spent
11 fuel pool and the ultimate heat sink.
12 And as discussed previously, the pool
13 water level is lower and the operating temperature has
14 increased. The fuel storage rack and design analysis
15 will be done by the COL applicant.
16 And also, increased core thermal power
17 will impact both the spent fuel pool and ultimate heat
18 sink cooling, due to the increased heat loads
19 associated with the nuclear power modules and spent
20 fuel assemblies.
21 MEMBER SUNSERI: Hey, this is Matt. I
22 have a question on that slide. Matt Sunseri.
23 MR. CRANSTON: Yes?
24 MEMBER SUNSERI: So, you're stating here
25 in the third bullet UHS's operating temperature
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1 increased. But yet, when I asked the question during
2 the NuScale presentation, they said it remained
3 unchanged. Can you explain that difference?
4 MR. STUBBS: Yes. What is referred to
5 there is the maximum allowable operating temperature.
6 And I think they increased that up to 120 degrees.
7 Oh, and my name is Angelo Stubbs. I'm
8 with Plant Systems, Containment and Plant Systems.
9 MEMBER SUNSERI: Okay. So, the maximum
10 allowed did not increase, but the operating
11 temperature did increase? That's what I would have
12 thought, anyway. But okay. Thank you.
13 MR. HERNANDEZ: This is Raul Hernandez
14 from Plant Systems.
15 It's the other way around. The normal
16 temperature stays the same. The maximum allowable for
17 evaluating the limits of the system, that is what
18 increased from 110 to 120.
19 MEMBER SUNSERI: Okay. I got it. Thanks.
20 MR. CRANSTON: And the chemical and volume
21 control system reconfiguration moved valves outside,
22 again, as was discussed previously by NuScale, and the
23 flow restrictors were added in conjunction with pipe
24 breaks. And the flow-restricting venturis are
25 credited in both Chapter 15 and 19, where they will be
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1 discussed further.
2 Regarding the SDA reactor building crane
3 description, one item that we wanted to note was that,
4 by going to controllers rather than total manual
5 operation, that it had an significant improvement on
6 core damage frequency, based on the calculated drop
7 probability. And that was a big improvement, which,
8 again, was discussed previously.
9 The application does not include specific
10 spent fuel pool criticality safety design information
11 and corresponding criticality safety analysis for the
12 SDA. This is addressed in two COL items requiring the
13 COL applicant to perform the criticality analysis for
14 the new fuel and spent fuel pool.
15 So, in conclusion, while there are some
16 differences between the DCA and the SDA, the staff
17 found the Applicant provided sufficient information to
18 support the staff's safety finding, which is that
19 staff found that all applicable regulatory
20 requirements were adequately addressed.
21 And that concludes my presentation.
22 MEMBER DIMITRIJEVIC: Hi. This is Vesna
23 Dimitrijevic.
24 Can you go to the previous slide on the
25 crane? So, they say that changes actually -- it
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1 resulted in operator error probability being
2 negligible. So, how? Because this was the add-on of
3 the Commission. So, how do those changes actually use
4 the standard of the Commission?
5 MR. CRANSTON: Marie?
6 I was looking for a staff member who could
7 --
8 MR. TANEJA: This is Dinesh Taneja.
9 MR. CRANSTON: Oh, Dinesh has got it. Go
10 ahead.
11 MR. TANEJA: Hi. This is Dinesh Taneja.
12 I am the I&C technical reviewer.
13 So, we audited the crane control system
14 design. And basically, the crane is designed to meet
15 the ASME Code requirements. And in accordance with
16 that, the control system is designed to failsafe and
17 all the safety features of the crane are implemented
18 in a segmentation that is independent from the control
19 system features that are used for controlling the
20 crane.
21 So, by having that independence between
22 safe follow-up protection on the control functions,
23 and having the failsafe features implemented, I guess
24 you can say that the reliability of the safety
25 features and the control features is significant, you
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1 know.
2 I think, you know, from that point of
3 view, you can kind of come to -- and in addition to
4 that, the other thing that we were trying to confirm
5 was that the software life cycle development
6 activities for the control system are going to be
7 performed in accordance with the software two-way
8 program that is documented in Chapter 7.
9 Specifically, I think it's Section 7.2.1. So, that
10 really, also, kind of adds to the reliability of the
11 development activities.
12 MEMBER DIMITRIJEVIC: So, I guess, then,
13 when we review the Chapter 19, we're going to see this
14 event developed in more details. So, we can see
15 actually what you're talking about reflected in this
16 probability. Is that what we should expect when we go
17 to Chapter 19? Because that event wasn't developed in
18 enough details to see something like this, you know.
19 And I guess, also, I was just really
20 wondering, I mean, in operator actions, we're not
21 really so specifically involved that you can see how
22 the controls can prevent these events. I mean, you
23 know, I was just wondering, are we going to see this
24 in a little more detail somewhere?
25 MS. POHIDA: Good morning.
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1 This is Maria Pohida, a Senior Reliability
2 Risk Analyst in the Division of Risk Assessment.
3 Yes, we will be discussing the reliability
4 and the reactability of the crane control system in
5 the Chapter 19 presentation.
6 NuScale developed a special PRA for the
7 reactability of the crane control system. I reviewed
8 that with the I&C Branch and their technical expertise
9 to verify the conclusions that operator errors are
10 negligible contributors to module drop, given the
11 assumptions of the reliability of the programmable
12 logic control and the control system itself. But that
13 will be discussed in Chapter 19.
14 MEMBER DIMITRIJEVIC: Okay. Thanks.
15 MS. POHIDA: You're welcome.
16 CHAIR KIRCHNER: While you have this slide
17 up -- this is Walt Kirchner -- criticality safety is
18 a COL item. I understand that.
19 This is a question for both the staff and
20 the Applicant. I presume, in lowering the level of
21 the ultimate heat sink, there were compensatory design
22 changes to the spent fuel pool which communicates
23 through a weir, such that the same margins in terms of
24 shielding and water for cooling of the spent fuel, the
25 assemblies in the spent fuel pool was maintained. Is
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1 that an accurate assessment of what was done with
2 regard to the ultimate heat sink and the spent fuel
3 pool?
4 MR. HERNANDEZ: This is Raul Hernandez
5 from Plant Systems.
6 While we were reviewing the NuScale 600,
7 the one with the higher level --
8 CHAIR KIRCHNER: Right.
9 MR. HERNANDEZ: -- the minimum water level
10 needed for cooling was 55 feet, not 68.
11 CHAIR KIRCHNER: All right.
12 MR. HERNANDEZ: The additional water level
13 was credited to provide buoyancy. So, the reactor
14 building crane could handle the fully loaded nuclear
15 power module.
16 CHAIR KIRCHNER: Right.
17 MR. HERNANDEZ: With increasing capacity
18 of the crane, NuScale no longer needs to credit
19 buoyancy.
20 CHAIR KIRCHNER: Right.
21 MR. HERNANDEZ: So, that additional margin
22 of water was not credited for thermal or radiation
23 protections. That additional margin was for buoyancy
24 to move the fuel.
25 CHAIR KIRCHNER: Right. Hence, the
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1 compensatory --
2 MR. HERNANDEZ: So, there is a -- sorry,
3 go ahead.
4 CHAIR KIRCHNER: Well, so I understand the
5 compensatory increase in load capability for the heavy
6 lift crane because you would have less buoyancy. But
7 my question was more directed towards the spent fuel
8 pool and whether, if you're lowering the overall level
9 in the ultimate heat sink and you're connecting the
10 two pools via a weir, did they drop the spent fuel
11 pool, the bottom of it, to the same level? Or maybe
12 it's always been that, the same level as the bottom of
13 the bay that accommodates the refueling and the six
14 modules?
15 MR. HERNANDEZ: No, the bottom of the pool
16 wasn't dropped. But what I'm saying is that the
17 additional water was never accounted on the safe
18 evaluation. We never credited that additional water.
19 CHAIR KIRCHNER: I see. Okay.
20 MR. HERNANDEZ: When we did our
21 evaluation, the minimum water level was 55 feet,
22 though they operated up to 68.
23 CHAIR KIRCHNER: Okay, okay. So, there's
24 been no diminishment of margin, so to speak, with
25 regard to the spent fuel pool?
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1 MR. HERNANDEZ: There was a lower -- the
2 water level has dropped. The minimum water level now
3 is 48 feet for the thermal analysis.
4 CHAIR KIRCHNER: Okay. Yes.
5 MR. HERNANDEZ: So, there is still
6 sufficient margin. And NuScale wants to talk, too.
7 MS. TURMERO: This is Sarah Turmero. With
8 the flow level drop, we still maintain the required
9 shielding above the fuel without any (audio
10 interference).
11 CHAIR KIRCHNER: Okay. And then, heat
12 load is not an issue?
13 MS. TURMERO: That's correct.
14 CHAIR KIRCHNER: Okay. It's shielding is
15 the dominant design consideration then, not heat
16 loading of the spent fuel pool?
17 MS. TURMERO: For criticality or?
18 CHAIR KIRCHNER: No, not for criticality.
19 For just heat removal.
20 MS. TURMERO: No, shielding is not a
21 dominant design factor.
22 Could you repeat your question to clarify?
23 CHAIR KIRCHNER: Okay. Thank you.
24 Members, further questions of the staff on
25 Chapter 9?
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1 MEMBER SUNSERI: None from me. This is
2 Matt.
3 CHAIR KIRCHNER: Okay. Yes, I neglected
4 to make this observation earlier for everyone
5 participating. We assign leads amongst the members
6 for each of the chapters. Our lead for I&C is Member
7 Roberts. Our lead for the auxiliary systems is Matt
8 Sunseri. Our lead for Chapter 12 on radiation
9 protection is Dave Petti, and our lead on Chapter 18
10 for human factors is Vicki Bier.
11 And with that, I think we're ready to turn
12 back to NuScale and Chapter 12.
13 (Pause.)
14 CHAIR KIRCHNER: NuScale, if you're ready
15 to proceed, go ahead.
16 MR. SLOBE: Hello. My name is Erik Slobe.
17 I'm a Licensing Engineer with NuScale Power. I've
18 with NuScale for a little over a year and have been in
19 the nuclear industry for eight years. I'm a Licensed
20 Professional Engineer in Pennsylvania.
21 I'll be presenting on Chapter 12 for
22 radiation protection of NuScale's Standard Design
23 Approval Application. I will be focusing on the
24 differences in the SDAA and the Design Certification.
25 In Chapter 12, we discuss ALARA and
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1 radiation shielding requirements. Chapter 12 consists
2 of five different sections. I'll be discussing each
3 of the sections today.
4 This includes Section 12.1 on ensuring the
5 occupational radiation exposures as low as reasonably
6 achievable; Section 12.2 on radiation sources; Section
7 12.3 on radiation protection design features; Section
8 12.4 on dose assessment, and Section 12.5 on the
9 operational radiation protection program.
10 We will be starting with Section 12.1 on
11 ensuring the occupational radiation exposures as low
12 as reasonably achievable. This section used the same
13 methodology as what was used in the Design
14 Certification Application. There were no significant
15 changes made to this section.
16 Next, we'll be talking about Section 12.2
17 on radiation sources. This section also uses the same
18 methodology as the DCA. The source term information,
19 including tables 12.2-1 through -31, updated changes
20 in source term information, are based mostly on the
21 change in cycling, the increase in burnup rate, the
22 change in thermal power, and the change in the number
23 of NuScale power modules.
24 For source terms for systems shared
25 between NuScale power modules, the tech spec design
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1 basis failed fuel fraction is applied to one NuScale
2 power module, and the realistic failed fuel fraction
3 is applied to the remaining NuScale power modules, as
4 discussed in Section 11.1.
5 There were a few audit questions from the
6 NRC that resulted in an update to the SDA. One audit
7 question involved dose rates for workers on the fuel
8 handling machine. The SDA was updated and clarified
9 that the workers are not exposed to dose rates above
10 2.5 millirem per hour.
11 Another audit question involves
12 clarification of the decay of N-16 in the CVCS and
13 when it decays to insignificant levels. This occurs
14 before the degasifier in the liquid rad waste system.
15 A third audit question involves the source
16 terms for the low conductivity waste processing skid.
17 Some of the source terms of the components of the LCW
18 processing skid are now shown separate in the total of
19 the skid in tables 12.2-12D and 12.2-13D.
20 Next, we'll be talking about Section 12.3
21 on radiation protection design features. As with the
22 other sections, the methodology used is the same as
23 the DCA methodology.
24 One difference from the DCA is there are
25 now no very high radiation areas. This is due to more
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1 realistic modeling of the solid waste prepared for
2 shipping and changes in the handling of the failed
3 fuel fraction, and the reduction in the number of
4 NuScale power modules.
5 There's also a reduction in the number of
6 reactor building and rad waste building shield doors.
7 This reduction in shield doors is due to the use of
8 labyrinths, some lack of end room radiation sources,
9 and some shielding analyses.
10 The number of fixed radiation monitors was
11 also reduced based on an analysis of which fixed
12 radiation monitors are necessary for providing
13 adequate coverage of the localized areas of the plant.
14 The reduction in fixed radiation monitors does not
15 interfere with the plant operations, equipment, or
16 personnel monitoring.
17 There were two COL items that were removed
18 from this section. The first is COL Item 12.3-5,
19 which concerned design criteria for additional area
20 radiation monitors. The design criteria and
21 regulatory criteria for area radiation monitors is
22 included in Section 12.3.4.2, which makes this COL
23 item unnecessary.
24 The second COL item that was removed is
25 12.3-8, which concerned radiation shielding for shield
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1 wall penetrations. This COL item is associated with
2 a carve-out into the DCA, which is now resolved. The
3 COL item is no longer needed due to the completion of
4 more detailed radiation shielding analyses. SDA
5 Section 12.3.2.4.1 has been updated to clarify that
6 the calculations of the penetration show adequate
7 protection from radiation streaming.
8 There were audit questions that impacted
9 Section --
10 DR. SCHULTZ: Erik, can you just back up?
11 When you say a more detailed shielding analyses, is
12 that new methodology or is it more detail associated
13 with the layout of the input to the methodology that
14 you've used in the past?
15 MR. SLOBE: Yes, so this is more detail of
16 analyses using similar methodology as other shielding
17 analyses are done for the plant, but it did take in
18 more detail of the layout of the plant.
19 DR. SCHULTZ: Okay. Thank you.
20 MR. SLOBE: In terms of audit questions
21 that impacted 12.3, there was some language that was
22 updated, based on a design change that replaced some
23 break pot tanks and vent lines with hooded vents.
24 There were two RAIs associated with this
25 section. The first concerned shielding. The language
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1 of the RAI was updated to clarify that the shielding
2 was based on nominal concrete equivalent gamma
3 attenuation.
4 The second RAI concerned radiation
5 monitoring post-accident. It was clarified that the
6 radiation monitors under the bioshield are PAM system
7 Type F variables, which provide primary information to
8 accident management personnel to estimate fuel damage
9 and the effects of fuel damage. This is in addition
10 to the Type B and C variables that were in the DCA.
11 There are additional Type E PAM variable radiation
12 monitors provided throughout the plant to effectively
13 monitor the accident progression.
14 And next, we will talk about Section 12.4,
15 which is on the dose assessment. The dose assessments
16 were completed using the same methodology as the DCA.
17 The results of the dose assessments differ from the
18 DCA primarily due to changes in cycle length; increase
19 in burnup rate; change of thermal power; the change in
20 the number of NuScale power modules; building layout
21 changes, and operational optimizations. The results
22 of these are shown in tables 12.4-1 through -7 of
23 Section 12.4.
24 Another change is that the vital areas for
25 post-accident actions no longer include areas for
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1 initiating combustible gas monitoring. This was also
2 related to a carve-out in the DCA regarding post-
3 accident doses for the leakage associated with
4 combustible gas monitoring systems. As discussed in
5 Chapter 6, the SDA does not rely on combustible gas
6 monitoring to assess core damage due to the use of the
7 passive auto-analytic recombiners.
8 COL Item 12.4-1 was also updated as part
9 of an audit question. The COL item was updated to say
10 that the dose to construction workers from collocated
11 existing NuScale power plants is the responsibility of
12 the Applicant.
13 For Section 12.5 on the operational
14 radiation protection program, there were no changes
15 from the DCA.
16 And that concludes my prepared for Chapter
17 12. Are there any other questions?
18 CHAIR KIRCHNER: This is Walt Kirchner.
19 On behalf of Dave Petti, who is not with
20 us today, I would like to ask a question going back to
21 the change in level in the ultimate heat sink and the
22 reactor pool. What impact does that have, or is it a
23 negligible impact, on radiation, starting first with
24 dose to the people doing the refueling operation? Is
25 that a material impact or is streaming the largest
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1 component?
2 By streaming, what I mean is coming from
3 the module itself rather than through the pool. Is
4 that the dominant radiation exposure for -- source, I
5 should say, for the refueling operation? You've
6 removed the bioshields on that particular unit. So,
7 I would presume streaming from the actual containment
8 vessel in the vertical direction would be the major
9 dose component?
10 MR. OSBORN: This is --
11 MR. BRISTOL: Go ahead, Jim.
12 MR. OSBORN: Yes, I'm sorry. This is Jim
13 Osborn, Licensing Supervisor with NuScale.
14 No, the major contribution to worker dose
15 in a refueling outage is going to be the evaporation
16 off of the pool water itself. With the reactor shut
17 down and removal of the bioshield, streaming -- and,
18 of course, the containment is flooded; the reactor
19 vessel is flooded. So, that provides sufficient
20 shielding from any streaming pathways from the reactor
21 core itself.
22 So, I think -- and Jon Bristol can
23 corroborate me or correct me, if I'm wrong -- but I
24 believe that the major contribution for a worker, you
25 know, after removing the bioshield is going to be from
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1 the airborne in the evaporation from the pool.
2 MR. BRISTOL: This is Jon Bristol,
3 NuScale, Engineering Radiological Manager.
4 For refueling, operator dose is a
5 contribution of NPM component activation and source
6 term in the ultimate heat sink from opening up the
7 NPM. And then, there is a little bit of shining from
8 the operating NPMs through the bioshield.
9 Ultimately, the lowering of the ultimate
10 heat sink water level changed the shielding design
11 criteria for the bioshield, which is adjusted to
12 account for that water level difference.
13 But, in our design and analysis, we
14 analyzed a variety of sources to operator dose.
15 CHAIR KIRCHNER: Okay. Well, it was
16 mentioned by your colleague that atmospheric source
17 from the pool was a contributor to that worker
18 exposure. Previously, in an earlier presentation, we
19 talked about the operating temperature of the pool.
20 So, is that a factor in the evaporation rate?
21 MR. BRISTOL: So, there is evaporation
22 from the ultimate heat sink and it does create
23 airborne particulate in the area. We have the HVAC
24 system designed to handle that, so that it's not a
25 qualified airborne area.
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1 Does that answer your question?
2 CHAIR KIRCHNER: Well, I'm just asking, if
3 you operate at a higher pool temperature, then you're
4 likely to have more evaporation; hence, more airborne
5 particulate. You've got an operating HVAC system, I
6 presume, continually during operation of the plant
7 that may be adequate to keep the dose, the buildup of
8 material in the atmosphere of the reactor building,
9 prevent that from increasing dose. I presume that's
10 the strategy there.
11 MR. BRISTOL: Correct.
12 CHAIR KIRCHNER: Let me reduce it to a
13 simple question. Does operating the pool at a higher
14 temperature increase the dose, the occupational dose?
15 Or is that a negligible factor?
16 MR. BRISTOL: It's a negligible factor.
17 The nominal ultimate heat sink temperature is
18 consistent between the DCA and SDA. The upper
19 allowable bound, if utilized, would be controlled
20 through the operational programs to account for the
21 radiological hazards of increased evaporation.
22 CHAIR KIRCHNER: Okay. And then, the
23 second half of my question, then, is -- I'll make it
24 a more direct question. Decreasing the water level,
25 does that materially increase the dose to the
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1 refueling operating crew?
2 MR. BRISTOL: No.
3 CHAIR KIRCHNER: Okay. Thank you.
4 MEMBER MARTIN: To follow up on that, on
5 the previous one, for clarification: so, if you've
6 increased the maximum, that's usually where you would
7 do your safety analysis. You've said that the
8 evaporation is the largest contributor. I would
9 otherwise expect that that's what shows up in those
10 safety analyses.
11 You would have updated your analysis with
12 120, right, and that would have factored into what you
13 presented?
14 MR. OSBORN: Yes, this is Jim Osborn.
15 Jon, I'm thinking that our normal
16 operating temperature for the UHS didn't change.
17 MR. BRISTOL: Right, right.
18 MR. OSBORN: And I think we evaluated our
19 evaporation for normal operations and normal worker
20 doses at 100 degrees Fahrenheit, is that correct, Jon?
21 MR. BRISTOL: That's correct. And for
22 Chapter 12.4, operator dose, you're not doing it at
23 design basis conditions. It's a realistic
24 representation of the operator exposure to radiation
25 in the plant.
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1 MEMBER MARTIN: Okay, well, I'm not a
2 technical person, but I am a safety person, so I'm
3 used to seeing that sort of thing being done at the
4 limits, but I appreciate you said that.
5 MR. OSBORN: Right, this is not a strict
6 cite to the analysis. This is radiation protection.
7 This is different.
8 CHAIR KIRCHNER: Members, further
9 questions? Okay, then I think we're ready to turn to
10 the staff again, and their evaluation of Chapter 12.
11 Just bear with us. It will take a moment to change
12 out.
13 (Pause.)
14 MR. TESFAYE: Okay, the staff is ready.
15 Can we start?
16 CHAIR KIRCHNER: Yes, go ahead.
17 MR. TESFAYE: Thank you, Chair. So just
18 an overview of Chapter 12. NuScale submitted the
19 Chapter 12, Radiation Protection, Revision 0 on
20 December 28, 2022, and Revision 1 on October 31, 2023.
21 NRC regulatory audit of Chapter 12
22 performed March 2023 to August 2023, generating 13
23 audit items. Eleven audit issues were resolved in the
24 audit. Nine audit issues resulted in NuScale
25 submitting supplemental information to address
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1 questions raised during the audit. Two Requests for
2 Additional Information were issued and resolved.
3 Staff completed Chapter 12 review and issued an
4 advanced safety evaluation to support today's ACRS
5 Subcommittee meeting.
6 Technical reviewer and the final presenter
7 today is Ed Stutzcage. Project manager, Chapter PM is
8 Alina Schiller. Unfortunately, she's not here. She's
9 on well-deserved annual leave this month. And I'm the
10 lead PM sitting in for Alina.
11 So this section has been presented by
12 NuScale. These are the various sections of Chapter
13 12. There's no point in going over them again. With
14 that, I will turn over the presentation to Ed.
15 MR. STUTZCAGE: All right, thanks,
16 Getachew. I am Ed Stutzcage in the Radiation
17 Protection and Accident Branch. I have, essentially,
18 five issues that we decided were significant to
19 include in the presentation slide for changes.
20 The first two are related to carve-outs
21 that were in the DCA. The first one here is related
22 to the hydrogen and oxygen monitoring. If you recall,
23 in the DCA, there was a carve-out due to the potential
24 radiological implications of performing hydrogen and
25 oxygen monitoring because you had to open up an --
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1 isolate containment following a major accident and
2 have a loop of the containment atmosphere through the
3 CVCS system, the sampling system, and I think the
4 containment flood and drain system.
5 In the SDAA, NuScale has requested an
6 exemption from hydrogen and oxygen monitoring
7 requirements. And that's being reviewed under Chapter
8 6, primarily, and I think Chapter 19, too, as well.
9 For the purpose of Chapter 12, the vital
10 area mission dose requirements, we assume that
11 exemption is going to be approved. It's still under
12 active review, but it's -- I believe it's on a path.
13 It's an issue that's changing, but it's on a path to
14 resolution, I think. And they want me to do that
15 hydrogen and oxygen monitoring. So, therefore, they
16 want me to do this analysis, so that kind of resolves
17 the carve-out that we had in the DCA.
18 Next slide, please. The second issue was
19 -- and NuScale spoke to this -- this is the radiation
20 -- the penetrations through the module bays. In the
21 DCA, the radiation streaming through the penetration,
22 what penetrations weren't explicitly considered. In
23 the SDAA, they did update. They did provide
24 calculations and assess the impacts of major
25 penetrations. And we audited the calculations and,
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1 essentially, they weren't very significant to the
2 dose. The major penetrations were really high in the
3 module bay area and they were adequately assessed in
4 the shielding and dose calculations.
5 Next slide, please. The third here is
6 related to N-16. In the SDAA design, N-16 travels
7 through the reactor coolant loop and the reactor
8 vessel a lot quicker, significantly quicker than in
9 the DCA. And, because of that, N-16 kind of makes it
10 outside the CVCS lines in more significant
11 concentrations than it did in the DCA. And we audited
12 the NuScale calculation as part of the audit and made
13 sure that the implications of this were assessed.
14 And, essentially, NuScale accounts for N-
15 16 up until 10 half-life, when it's no longer
16 significant. So the shielding and any radiological
17 impacts to equipment qualification or anything, they
18 assess the N-16 dose appropriately.
19 So that's that item. Next slide, please.
20 Did we skip one? I think we might have skipped one.
21 There we are. We skipped one earlier. Yeah.
22 This is just all the source terms and
23 associated analysis with radiation shielding and doses
24 and everything that were changed due to changes in the
25 reactor power level, the cycle length, the number of
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1 units, all that. So, all the source terms, shielding
2 calculations, doses analyses were all updated to
3 account for this. And we reviewed the source term
4 shielding based on the safety significance and found
5 them to be acceptable.
6 Next item. And then the last one here is
7 the change to the shielding approach for -- the way
8 it's described in Chapter 12. So, Chapter 12 provides
9 all the major radiation shielding for the plant. And,
10 besides the bioshield shielding and water in the pool
11 and stuff, the shielding is all described in terms of
12 the concrete shielding thickness. It's evaluated for
13 the appropriate -- for radiation attenuation based on
14 concrete.
15 NuScale has updated their application to
16 kind of say that if they want to use different
17 shielding materials -- and through the audit we found
18 that there may be a few cases where they want to use
19 different shielding materials -- that it will meet the
20 Chapter 12 radiation zoning and will meet all
21 applicable regulatory requirements associated with
22 whatever the shield wall is necessary for, just, you
23 know, the shielding is needed for. So there's just
24 changes in the FSAR language to describe the criteria
25 that need to be met if an alternative shielding to
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1 concrete is specified. I think that's all for this
2 issue.
3 So those are the significant items that we
4 identified. There were some other changes that, most
5 of them, NuScale described in their presentation that
6 we didn't think were very significant, so we didn't
7 include them in the slides.
8 Any questions?
9 DR. SCHULTZ: Edward, this is Steve
10 Schultz. Just to go back to the catalytic converter.
11 You mentioned that that issue is still being
12 evaluated. I presume that what needs to be identified
13 is how effective the converter will be and what's its
14 reliability and availability, interaction, and
15 conditions. Are those are the things that you're
16 exploring?
17 MR. STUTZCAGE: It's not my review. I
18 relied completely on those combustible gas reviewers
19 in the Chapter 6 review. But I believe that's --
20 yeah, that's what they're -- that would be what
21 they're looking into. And if, for whatever reason,
22 that exemption from hydrogen and oxygen monitoring was
23 determined to be -- you know, they still needed to do
24 it, it couldn't be exempt from it, they would let me
25 know, and this would then be an open item in Chapter
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1 12 and Chapter 15.
2 So that's kind of where we are. Right
3 now, we're in this spot where we think we'll get to
4 where we won't need to do this monitoring, and we're
5 assuming that for Chapter 12.
6 DR. SCHULTZ: As you understand, that
7 review is on track?
8 MR. STUTZCAGE: I know there's been some
9 ongoing review and changes going on there. I can't
10 speak to specifics of that review. That would be in
11 Chapter 6.
12 MR. TESFAYE: This is Getachew Tesfaye.
13 There is no significant issue associated with that
14 review that we know of. We presented these chapters
15 earlier on. That's with the understanding we would go
16 back. If something changed, we'd go back and reassess
17 all the chapters as presented to you. So, that will
18 happen when we complete the entire chapters that
19 includes Chapter 6, Chapter 19, Chapter 15. We do
20 anticipate some minor changes, to go back and tell you
21 what the delta is. So when we evaluate Chapter 6, we
22 will come back. If there's anything negative impact
23 to Chapter 12 we will assess that then.
24 DR. SCHULTZ: But there's no expectation
25 that there's going to be problems associated with
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1 this?
2 MR. TESFAYE: As it stands, I don't have
3 any significant issue.
4 DR. SCHULTZ: Thank you.
5 MR. STUTZCAGE: So, yeah, I guess we
6 didn't really do to the conclusion slide. While there
7 is some differences, as we've discussed, we found that
8 they provided sufficient information to support our
9 safety finding, and that all applicable regulatory
10 requirements were adequately addressed for the design
11 within the scope of this chapter. And the COL items
12 that are provided for the program and site-specific
13 aspects are mostly similar to the DCA, and they're
14 appropriately provided in the SDAA for the applicant
15 to address.
16 MR. TESFAYE: That concludes our
17 presentation. Any additional questions from members?
18 CHAIR KIRCHNER: Members, any further
19 questions?
20 Actually, then, if there are no further
21 questions, I'll use this opportunity to make a
22 statement on behalf of the Committee. The Committee,
23 April 28, 2020, wrote a letter on this topic that we
24 were just previously discussing, combustible gas
25 monitoring. And we raised serious concerns about the
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1 approach that was contemplated during the DCA, which
2 would have required opening a large pipe, the
3 containment evacuation system line, to set up a loop
4 so that a realistic grab sample could be obtained to
5 monitor hydrogen and oxygen content in the
6 containment.
7 It was the view of the Committee that this
8 presented a much greater risk of exposure to the
9 workers, and potentially offsite, than the information
10 that would have been gained in trying to assess what
11 the hydrogen and oxygen content was.
12 So what the applicant has done, I think,
13 addresses -- is a much better approach. We will
14 review that subsequent to the exemption request and
15 the details of the radiolytic -- I misspoke, the
16 catalytic recombiner performance, in a subsequent
17 chapter. But this change in design approach by
18 NuScale and the review by the staff addresses a major
19 concern that had been raised during the DCA.
20 And so a preliminary -- I won't draw any
21 preliminary conclusions on behalf of the Committee,
22 but I do just want to acknowledge the applicant's
23 effort here to address what appeared to be a major
24 risk-contributing process or procedure for post-
25 accident monitoring. And so I'd just put that in the
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1 record.
2 Any further questions?
3 (No response.)
4 CHAIR KIRCHNER: Well, with that, let's
5 turn then back to NuScale and Chapter 18. And bear
6 with us again while we change both the presenters and
7 the slides. Thank you.
8 (Pause.)
9 CHAIR KIRCHNER: Doug, if you're ready,
10 please proceed.
11 MR. BOWMAN: Thanks, Walt. Good morning,
12 everybody. My name is Doug Bowman. I'm Plant
13 Operations Manager for NuScale Power. I've been at
14 NuScale for 10 years now. All of my work at NuScale
15 has been in the area of either human factors
16 engineering, plant procedures, or training.
17 Prior to coming to NuScale, I spent 24
18 years in commercial nuclear power. I was SRO-licensed
19 at both D.C. Cook and Byron Station, and also served
20 many other positions along that path.
21 Next slide, please. So, Chapter 18 is
22 broken up into 12 sections, and these sections are
23 broken up as NUREG-0711 breaks them up, by element.
24 The bolded sections you see there will be the ones we
25 will cover, as they're the ones that have changes in.
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1 The rest of them we will not cover. So those are
2 functional requirements, analysis of functional
3 allocation, task analysis, staffing and qualification,
4 treatment of important human actions, and human-system
5 interface design.
6 You can go on to the next slide. So, one
7 of the big things we wanted to discuss here is, one of
8 the things we did change from the US600 DCA to the
9 SDAA is what we are submitting for this portion. So
10 in the DCA, previously, we submitted a number of RSRs.
11 We submitted all the elements, all the RSRs for the
12 elements necessary get to the point of performing ISV.
13 And then we performed the integrated system validation
14 during the review phase of the DCA.
15 For this submittal, we did change that up
16 a bit. We are submitting implementat ion plans.
17 Implementation plans are allowed. They are strictly
18 a description of methodology we will use, whereas
19 Results Summary Reports describe both the methodology
20 and results. And NUREG-0711 does allow for either
21 one.
22 So we did take certain elements and we
23 pulled them back to where we were only submitting IAP
24 (phonetic) for the SDA, and this was done for a number
25 of different reasons. One was a bit of lessons
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1 learned from the DCA. Another is that, if you look
2 over what your human factors engineering program is
3 supposed to cover, your human factors engineering
4 program is supposed to be in place very early on in
5 the design process, but it really needs the entire
6 design complete in order to finish its portions of it
7 to really understand how the humans will interface
8 with it.
9 So we really wanted to move back to a more
10 traditional model of human factors engineering, rather
11 than the way we did it previously in the DCA. And we
12 understood why we had to do that. We had novel design
13 concepts and novel staffing plan, and we really had to
14 be able to demonstrate that to the NRC to show that
15 that would be acceptable. And now that we've done all
16 that work, we feel like we can do our work in a more
17 traditional manner.
18 Next slide, please. So our first section
19 to discuss, then, is 18.3, which is functional
20 requirements analysis and function allocation. The
21 purpose of this element is to take those functions
22 that the plant design has already been broken down
23 into and understand how the operators will complete
24 those functions. And, really, this breaks down into
25 either it's going to be a manual operation by the
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1 operator to complete that function, a shared condition
2 where you have both automation and manual actions, or
3 a fully automatic function.
4 And the big thing we changed here is, as
5 we matured our process, previously in the DCA we had
6 separate databases tracking FRA/FA, task analysis, and
7 procedures. And what we did was really combine all
8 those together in one single interlinked database.
9 And this is really an advantage for us for an
10 efficiency and auditability standpoint. We now have
11 direct connections from procedures to training, and
12 procedures to human factors engineering elements. So
13 it really becomes a very powerful tool for us to use
14 down the road if we can continue to maintain human
15 factors engineering programs as we move through the
16 design process.
17 Next slide, please. 18.4 is similar. We
18 actually use a very similar process to what we did in
19 the DCA for task analysis. However, this, again, was
20 combined as part of that large single interlinked
21 database as we previously described in Section 18.3.
22 Next slide, please. 18.5, staffing and
23 qualifications. S&Q determines the number and
24 qualifications of licensed operators required for safe
25 and reliable plant operation. Our minimum staffing
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1 numbers for the SDAA are one licensed reactor operator
2 and two licensed senior reactor operators. All of
3 these changes are a result of the topical reports that
4 were submitted and was previously reviewed by both the
5 staff and the ACRS. That staffing plan, that topical
6 report number is listed there, and we also have the
7 reviews by the ACRS, the ML numbers, back in April of
8 2021. So we essentially just leaned on that new
9 revised staffing and plant validation methodology to
10 develop our minimum operator-licensed staffing for
11 this design.
12 Next slide, please. 18.6, treatment of
13 important human actions. So identification of
14 important human actions, we go and we look at Chapters
15 7, 15, and 19. Chapter 7, we're looking at the D3
16 analysis, looking to see if there's any operator
17 actions listed there. In Chapter 15, we're obviously
18 looking at the accident analysis to see if there's any
19 actions that the operators would have to perform
20 there. And then 19, we're looking for anything that
21 is identified as risk-important in Chapter 19 would
22 rise to that level of risk-importance of a human
23 action. And the US460 standard design, unlike the
24 DCA, has two important human actions. The U.S.
25 standard design does not have any important human
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1 actions.
2 Importantly, that last bullet there I'd
3 like to highlight. No matter what the action is, if
4 we assume an action takes place, especially in PRA in
5 Chapter 19, we cover that action in procedures,
6 training, and development of the human-system
7 interface. So, again, to reiterate, if there is an
8 action assumed in Chapter 19, we still address those
9 actions and still make sure the operators are capable
10 of performing them correctly. So even though we
11 eliminated two important human actions, those same two
12 actions are still addressed by our procedures set and
13 our HSI.
14 MEMBER DIMITRIJEVIC: This is Vesna
15 Dimitrijevic. You didn't eliminate those actions; you
16 just didn't classified them as important. But those
17 actions are still in the PRA, right? Those were
18 actions to make up, you know, for the coolant in the
19 case of isolation and the LOCAs outside of
20 containment. So those actions are still in the PRA,
21 but they didn't show as important in the new Chapter
22 19. That's what you're saying?
23 MR. BOWMAN: That's exactly what I'm
24 saying. That's an excellent --
25 MEMBER DIMITRIJEVIC: All right. One of
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1 the points which I find a little strange is these
2 actions that are identified as important to the Level
3 2 because of the LOCAs outside of containment. And
4 based on some of the design changes, the math
5 mentioned in, you know, the containment isolation, you
6 would assume that LOCAs outside of containment become
7 more important, but, actually, based on what you are
8 saying, if those actions are not important anymore, it
9 seems that LOCAs outside of containment are not any
10 more significant contributors to the Level 2.
11 MR. BOWMAN: I'll say they're not as
12 significant of a contributor. And I'm going to go
13 into a little bit of detail and then I'm probably
14 going to state that we should probably ask this
15 question in Chapter 19. But there were changes made,
16 especially in ECCS. So, we had changes in actuation
17 set points, and what we learned through our process in
18 PRA is that, if we had both trains of ECCS actuated,
19 even with a containment bypass event, i.e., a LOCA
20 outside of containment, we did not run into core
21 damage, even if no operator actions were taken.
22 So that change in ECCS capability, and
23 also the change in the actuation set point, it now
24 comes off of riser level and it actually hits much
25 earlier in the event, so you wind up not needing to
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1 take operator action, or it's not required to take the
2 operator action in that case to prevent core damage.
3 Now, oddly enough, if we had those
4 conditions, by our emergency operator procedures, we
5 would still take the action to go ahead and initiate
6 CBCS injection in those cases. So we still have --
7 MEMBER DIMITRIJEVIC: So, basically, what
8 you're saying, maybe those actions are not anymore
9 modeled in the PRA. They're not necessary for your
10 assumed mission, then, right?
11 MR. BOWMAN: Correct. Right.
12 MEMBER DIMITRIJEVIC: I may be wrong in my
13 previous statement that those actions are still in the
14 PRA, because what you are saying, they may not be in
15 the PRA anymore.
16 MR. BOWMAN: No, I'm sorry. They are
17 still in the PRA. They are still assumed in the PRA,
18 but the changes in the design allow us to -- even if
19 we don't take the operator actions, if we take down
20 the path, the fault tree path where the operators
21 don't complete the actions, we still don't wind up in
22 that core damaging situation, if we have both trains
23 of ECCS actuated.
24 MEMBER DIMITRIJEVIC: I see. Okay. Well,
25 we will pay attention to that in reviewing Chapter 19.
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1 I have another question for you, what we
2 also discussed previously in the, you know, previous
3 discussion on that Chapter 9. And this is the
4 operator commission actions and shutdown on the crane.
5 And these actions are obviously considered, based on
6 our previous discussion today, both in the chapter,
7 you did not identify risk-important -- nothing risk-
8 important for Chapter 19, but also you didn't get the
9 actions from the Chapter 7 and 15, right, to be
10 considered? And I was wondering why this action,
11 commission action on the crane was not considered
12 even, because if it was eliminated or reduced in the
13 likelihood that would be good example or a example of
14 human-system interface. I mean, why was not this
15 action not even mentioned in Chapter 18?
16 MR. BOWMAN: At this point, I'd have to
17 assume that, based on the information we have with the
18 automations and the evaluation of the automatic
19 functions for operation of the crane, that that would
20 then cause a reduction of risk. But, again, I think
21 we talked about this previously. It's probably better
22 off discussed in Chapter 19 how those -- how all that
23 combines to affect the risk.
24 MEMBER DIMITRIJEVIC: I do agree with your
25 previous conclusion of makeup with the charging of,
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1 you know, the CFDS. But the thing is here, when it
2 comes to the commission action, I thought you should
3 identify this action through the Chapter 7, for
4 example, 9 or something. And at least discuss it in
5 the Chapter 18. I mean, we will look at that in
6 Chapter 19, but this would be a very good example of
7 the deterministic inclusion of the actions. So,
8 that's just my comment and my personal opinion.
9 MR. BOWMAN: Thank you. I appreciate it.
10 CHAIR KIRCHNER: Doug, while we're on this
11 -- this is Walt Kirchner again. I don't know that it
12 rises to the level of a IHA, as rigorously defined,
13 but for a fire in the main control room and
14 evacuation, you do identify the alternate shutdown.
15 Since you've taken away the shutdown -- there are no
16 shutdown station kind of capability, you would have
17 operators go to an alternate location to be able to
18 effect a shutdown if they didn't do that from the main
19 control room. I know that's part of the procedure.
20 So does that rise to the level of a IHA?
21 The alternate shutdown locations and the training that
22 would be necessary to ensure that when the operators
23 went to that location, they operate the equipment to
24 effect a shutdown and actuation of the ESFAS system.
25 MR. BOWMAN: So, I'll try to answer that
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1 question. So, the alternate operator work stations,
2 first of all, really there's no change in capability
3 at the alternate operator work stations. The remote
4 shutdown station that was described in the DCA was
5 there strictly for operator monitoring. They were
6 intended to complete the shutdown and placing the
7 units in safe shutdown prior to leaving the control
8 room. We do include in that procedure set a set of
9 actions that can be taken in contingency if you are
10 unable to ensure safe shutdown from the main control
11 room once you got to the alternate operator work
12 station.
13 CHAIR KIRCHNER: Right.
14 MR. BOWMAN: And those would include
15 likely going to the MPS rooms to complete those
16 actions if you needed to do that.
17 CHAIR KIRCHNER: Right.
18 MR. BOWMAN: But this is all in -- there's
19 really no path that -- I guess, from a deterministic
20 standpoint, it's not identified as an issue because
21 it's not a -- no accidents are assumed to occur with
22 a control room evacuation. That's in the regulation.
23 CHAIR KIRCHNER: Correct.
24 MR. BOWMAN: And so if you don't have an
25 access, it's really hard to get to a situation where
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1 that would be important. So it's not a credited
2 action anywhere in Chapter 15 since we don't have an
3 accident to mitigate.
4 And from a D3 standpoint, again, D3 has a
5 little bit difference of a focus, I guess. It's
6 really surrounding the digital architecture and what's
7 potentially not diverse, and what you have to do in
8 order to address those potentially not-diverse digital
9 assets.
10 So it really doesn't show up anywhere in
11 any of those paths of important human actions. It's
12 either a deterministic thing that you had to do for
13 Chapter 15 to address an accident, or something you
14 had to do to address a lack of diversity in your
15 digital control system, or it's a risk-identified
16 action. Those are the three paths we have to get
17 there. It didn't rise up in any of those.
18 CHAIR KIRCHNER: Okay. Thank you.
19 MR. BOWMAN: All right. Next slide. And
20 then 18.7, HSI design. The HSI design element takes
21 all that other work we've done prior in terms of task
22 analysis, FA/FRA, and treatment of important human
23 actions, staffing qualification, and builds the HSI
24 design in.
25 And really what we're highlighting here is
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1 the fact that there are changes in the US460. Those
2 changes are either in response to operator feedback to
3 the ISV we performed in DCA, or design changes we had
4 to make from the US600 to get to the US460. So all
5 those design changes you're going to hear about over
6 the next month or so are all the things that we had to
7 address in the HSI design.
8 That's really the only two paths we have.
9 We either got operator feedback and we had to change
10 something to address it, or we have a design change
11 that came from the US600 which affects one of our HSI
12 designs.
13 Next slide, please. And then we also
14 added one ITAAC. This is a result of our
15 implementation plan strategy. Previously, we had a
16 single ITAAC that identified the design implementation
17 element, and you had to ensure that the final control
18 room, as built, matched the control room design that
19 you had developed.
20 In this case, we added one to ensure that
21 the integrated system validation test was completed
22 once we had completed all the other elements that led
23 up to integrated system validation. So, really, the
24 final thing is that we have a report exists and it
25 concludes that the acceptance criteria associated with
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1 the ISV has been satisfied. So that's our additional
2 ITAAC.
3 Next slide, please. And then the summary
4 of the audit. We did have 20 audit items that were
5 successfully resolved. They included a wide range of
6 questions from staff. And we also included a virtual
7 demonstration of both some of the simulated changes we
8 had made, and the staff review of those databases we
9 talked about. So we did a pretty detailed virtual
10 presentation with staff on the databases to show them
11 how they were connected.
12 Next slide, please. Are there any other
13 questions for me?
14 MEMBER BIER: I guess my question kind of
15 goes beyond what your proposing and applying for right
16 now kind of more long-range, which is regarding
17 staffing levels. If there were to be increased
18 numbers of units, would you envision that staffing
19 levels would increase kind of linearly or maybe stay
20 constant? Or you haven't really thought that far yet
21 to have a plan?
22 MR. BOWMAN: We certainly have. Again,
23 the revised staffing plan validation showed that, for
24 up to 12 units, the minimum operators required for
25 safe operation of the facility is three. That's one
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1 RO and two SROs. Obviously, there are commercial
2 considerations that would potential require you to
3 bring additional operators on staff. We can envision
4 a number of different staffing models from essentially
5 an extra shift on day shift to deal with those
6 maintenance and other commercial issues, to
7 potentially -- if you've got a 12-unit plant, you
8 might want to add an extra operator normally, just to
9 deal with work. But, again, that's all commercial
10 considerations. It's really up to the licensee on how
11 they want to staff based on that workload. Again, our
12 conclusion is that three is what's necessary for safe
13 operation of the facility.
14 MEMBER BIER: Okay. And the lack of a
15 designated SDAA is, again, based on kind of timeline
16 and remote support options? Or can you just go over
17 the rationale for that again?
18 MR. BOWMAN: So the SDAA, elimination of
19 the SDAA was identified at the time that the TMI rule
20 came out, what you had to do to do that.
21 MEMBER BIER: Yes.
22 MR. BOWMAN: And we showed in the staffing
23 -- the topical report we did for staffing, that we had
24 met those requirements, which really came down to SRO
25 training upgrades and HSI upgrades. So we believe we
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1 showed and demonstrated in the topical report that we
2 had met those requirements of the original TMI plan
3 and that we no longer needed the capability of having
4 an SDAA.
5 MEMBER BIER: So it's really based on the
6 timing issues. The timing issues are maybe an
7 additional advantage, but not necessary.
8 MR. BOWMAN: I wouldn't say the timing --
9 you're talking about the length of time for an
10 accident to progress and the need for -- I would say
11 was definitely a factor in that. It certainly helped
12 the staff with their review that they knew that we
13 could -- we had on the order of an hour before we
14 really need any supplementary people, and even longer
15 in most cases.
16 MEMBER BIER: So it's more of a kind of
17 comfort level factor at the time?
18 MR. BOWMAN: Yes.
19 MEMBER BIER: Okay. Thank you.
20 MR. BOWMAN: Any other questions?
21 MR. BROWN: Yes, I'm going to backtrack --
22 this is Charlie Brown -- on two points, but I want to
23 backtrack to the remote shutdown station issue again.
24 When we originally reviewed the HIPS
25 systems and everything else, and we went through the
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1 first versions of these, we had the remote shutdown
2 system. It was remote shutdown. That's the words
3 that were used. That was not just monitoring. Maybe
4 I misunderstood back five, six, seven years ago, but
5 we were assured, based on the other comments and the
6 notes in that earlier revision on the DCA as well,
7 that there were manual control switches that then
8 actuated and ran down -- I'm looking at the old figure
9 again -- and actuated the scram for the scram breakers
10 to get them shut down, hardwired. And that was the
11 hard-wired module that went down.
12 Now you're telling me that there is no
13 remote capability of manual shutdown through the
14 hardware systems if you have to abandon the main
15 control room under some circumstances.
16 So, I did not realize that when I did the
17 review of this, so I'm just throwing that one back on
18 the table, Walt. I'm not particularly enamored
19 without having a hardware method for shutting down the
20 reactor plant, tripping the scram breakers, if the
21 main control room is unavailable for that actuation
22 from internal.
23 So, I don't think that's a good idea.
24 Maybe nobody likes that comment, but that's my thought
25 process on that.
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1 The second question was now on the
2 staffing. There's one licensed reactor operator and
3 two licensed senior reactor operators. But are those
4 senior reactor operators always -- at least one of
5 them, always in the main control room, along with the
6 licensed operator?
7 MR. BOWMAN: So I'll take your second
8 question first. I would like to address your first
9 statement, too. Yes, there's always -- the minimum
10 staffing inside of the main control room at all times
12 MR. BROWN: Okay, that's good.
13 MR. BOWMAN: One person could go out in
14 the field to check on things or other things, but,
15 yeah, our bare minimum staffing in the control room is
16 that.
17 MR. BROWN: So there will always be one of
18 the two -- or one of the three could leave and then
19 come back, but there would always be at least two
20 people in the main control room?
21 MR. BOWMAN: Correct.
22 MR. BROWN: Okay, you wanted to go back to
23 question one again.
24 MR. BOWMAN: Yes, I do. So we absolutely
25 have the capability outside the control room to open
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1 the reactor trip breakers in place the reactor in safe
2 a shutdown.
3 MR. BROWN: How?
4 MR. BOWMAN: We would have to go to the
5 MPS rooms and perform those tasks locally. But that's
6 no different than the current plan --
7 MR. BROWN: What's MPS again?
8 MR. BOWMAN: Module protection system
9 rooms.
10 (Simultaneous speaking.)
11 MR. BROWN: So you have to actually go to
12 the reactor instrument rooms to do that, as opposed to
13 where you had all the monitoring?
14 MR. BOWMAN: Correct. That's no different
15 than what current plants do today.
16 MR. BROWN: Not some of the plants I'm
17 familiar with. Okay.
18 MR. BOWMAN: The two plants I was licensed
19 at you had to go --
20 MR. BROWN: I'm not talking about
21 commercial.
22 (Simultaneous speaking.)
23 MR. BOWMAN: -- went down to the trip
24 breakers and opened them locally.
25 MR. BROWN: Well, I understand you can do
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1 that. I'm referring back to my naval bona fides after
2 that, which we could it at the cabinets. We could
3 also do it in the main control room, or from a local
4 station, which all you had to do is take three steps
5 to do it.
6 Let me think about that a little bit. How
7 far is the module protection system cabinets from the
8 main control room?
9 MR. BOWMAN: I would have to get you that.
10 I do not have a number on the top of my head.
11 MR. BROWN: But somebody has to exit --
12 you have to exit. You're out now, and you have to run
13 somewhere, go up and down steps, down to the room,
14 unlock -- do you have unlock a cabinet or unlock a
15 switch that's protected with a lock so it's not
16 inadvertently actuated?
17 MR. BOWMAN: There would be access control
18 requirements on the door to the room. But, beyond
19 that, the operators would have access to that so they
20 could get in and perform the actions promptly. So
21 it's nothing more than the time -- and don't forget --
22 I'll also back up. I'm going to have to state this.
23 Those operators are complete -- those plants are
24 completely safe without the operators in the main
25 control room, in our design. That's very different
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1 than what's currently in the plant designs, right?
2 MR. BROWN: I understand that.
3 MR. BOWMAN: The protection system is
4 fully operable. It will still shut the plant down
5 automatically, if needed. In the meantime -- the
6 plant is safe no matter what condition we're in,
7 because the plant is safe without operators in our
8 design.
9 MR. BROWN: I understand that. In spite
10 of all those other types -- and I understand it. I've
11 been through this for quite a few years now, and I'm
12 always uncomfortable with having delays in order to be
13 able to manually shut it down. In this case, you
14 leave the room. You have to go somewhere up and down
15 steps. If you have to have access to another room,
16 does that require a key or a special somebody else to
17 allow you in? It's just more difficult.
18 All right. I threw that back out if the
19 Committee -- I'm just a consultant now, so if the
20 Committee says this is okay, we'll proceed. I just
21 wanted to get the discussion out on the table so that
22 everybody had a common understanding of what this
23 actually looks like now.
24 So, Walt, I'm done with that, if I've
25 mouse-milked this enough.
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1 CHAIR KIRCHNER: Thank you, Charlie.
2 Members?
3 MR. BROWN: That means I'm done.
4 CHAIR KIRCHNER: No, it means you've been
5 heard as well, Charlie. It's a concern that we share,
6 I think.
7 MEMBER MARTIN: Hey, Walt. This is Bob.
8 I just wanted to follow up, maybe just fill in my own
9 lack of knowledge. You mentioned the MPS, and I
10 assume physically it's located somewhere relative to
11 the main control room and everything?
12 MR. BOWMAN: So it's on the reactor
13 building. They're essentially immediately adjacent
14 to the modules just outside of the --
15 MEMBER MARTIN: Now, should be expect
16 that, like, the design criteria for the control rooms
17 are the same for this space?
18 MR. BOWMAN: You mean for the space below
19 the operating --
20 MEMBER MARTIN: Well, for the MPS room, if
21 you have to go in. Obviously, they're not going to be
22 there very long if they're doing an action like that,
23 so that factors into it. But it's really my question
24 to fill in a gap in my knowledge. Is there any
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1 requirement that that -- I'm really thinking about
2 dose, related to that particular space and whether it
3 aligns with what's otherwise required for the main
4 control room?
5 (Simultaneous speaking.)
6 MR. BOWMAN: I can at least tell you from
7 a EQ perspective it's a mild environment, right? So
8 it should be reasonable dose rates if it were required
9 to access. Temperatures are controlled, so it's not
10 going to be hot or anything like that. And it is a
11 vital area, so it is controlled from a security
12 perspective. But, again, to kind of go back to
13 Charlie's question, the operators will have access to
14 those rooms. Normally, in my past life, you had a
15 badge that you show to the reader and got access to
16 the room so they can take the action.
17 MEMBER MARTIN: Thanks.
18 MEMBER ROBERTS: Yeah, this is Tom
19 Roberts. Can you go back to that slide on the
20 important human actions? I also have kind of an
21 ignorance question I wanted to ask while we were here.
22 The third bullet says there's a credited
23 deterministic human action from the D3 analysis. I
24 was looking through the D3 analysis just now. So what
25 are examples of deterministic human actions that are
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1 credited?
2 MR. BOWMAN: I guess that's a poorly
3 written statement. There are no -- if there were
4 deterministic human actions, they would have been
5 credited D3, but there aren't any for us.
6 MEMBER ROBERTS: Okay, yes, that's pretty
7 clear.
8 MR. BOWMAN: Sorry, that's a poorly
9 written statement, I agree.
10 MEMBER ROBERTS: All right, thank you.
11 CHAIR KIRCHNER: Doug? This is Walt
12 again. Just a follow up on Charlie's line of
13 questioning. So at the NPSs, let me just suggest one
14 scenario.
15 If you were to turn off the power to the,
16 at the chassis, would that in effect, result in the
17 scram of the system?
18 MR. BOWMAN: Yes.
19 Removing power from NPS system, all
20 actions are failed to their safe condition. So, if
21 you turn off power, everything's going to actuate.
22 So, yes, turning off the chassis would --
23 (Simultaneous speaking.)
24 CHAIR KIRCHNER: Yes, so that's what I
25 would expect, a fail-safe design.
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1 So Charlie, does that answer your
2 question? If the operator is forced, the operators
3 plural, the three-man crew, have to evacuate the main
4 control room, then they could proceed to the NPS and
5 also effect a manual scram by just turning off the
6 power to modular, module protection system.
7 MR. BROWN: Well, I would have to go back
8 and look, but the breakers that they have are if you
9 lose power to the breakers, that means the under-
10 voltage coils, whatever, they will trip the breakers.
11 It's a matter of how --
12 CHAIR KIRCHNER: Yes.
13 MR. BROWN: -- how it was delivered by the
14 NPS and do you have to turn off all, how many channels
15 do you have to turn off because they've got
16 independent power supplies.
17 And each of those power supplies is fed by
18 a separate breaker from some power panel somewhere.
19 So, it's not just one action. You have to
20 take multiple actions to get them de-energized. You
21 could get there. It's just the timing of it is a much
22 longer period.
23 And the general argument that they're
24 giving is, if everybody died in the plant, the plant
25 is safe.
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1 I'm saying that kind of, I'm maybe
2 overstating it, but just to make the point of this
3 plant.
4 CHAIR KIRCHNER: Well, this here --
5 MR. BROWN: Everybody, we could leave the
6 site and this plant is safe. You don't have to worry
7 about it. And I'm a typical skeptic just based on 35
8 years of previous experience that you ought to be
9 skeptical.
10 So, that's why I like manual switches that
11 just bypass all the other stuff that has to activate,
12 or deactivate, in order to turn them off.
13 You ought to be able to just remove, trip
14 the breakers by turning off all the under-voltage
15 coils remotely, without any access from the main
16 module protection system.
17 And back in the old days when we first
18 looked at this years ago, there were manual switches.
19 At least that's what they were advertised as being.
20 And now they've disappeared.
21 That's the major change I see. I'm not
22 saying your plant is unsafe, I'm not saying any of
23 that. It's just I'm not particularly comfortable with
24 that.
25 If I was on the committee and helping to
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1 write the letter, I would make that known. But in
2 this circumstance, I'm only presenting it as my
3 thought process.
4 But I don't like back stepping like this.
5 So, I don't know if that answers Walt, but that's,
6 that's just long time and a lot of plants being
7 responsible for 35 years. So, in real time.
8 MEMBER ROBERTS: Well, let's ask the
9 person who is responsible for writing this letter. I
10 guess I wanted to respond to Walt's hypothesis with a
11 question for the applicant.
12 If you really turned off the MPS power
13 supply from the control room, does that deactivate the
14 UPSs that, or holding the power for some time after
15 loss of input power, or would you really be able to
16 completely de-energize the MPS from the control room?
17 MR. BROWN: That's a good point, Tom.
18 CHAIR KIRCHNER: Well, I wasn't suggesting
19 from the control room. I was suggesting de-energizing
20 from the MPS panel. Presuming --
21 (Simultaneous speaking.)
22 MEMBER ROBERTS: Right. I was thinking if
23 you can get to that room, I think you would just trip
24 the breakers.
25 CHAIR KIRCHNER: Probably.
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1 MR. BROWN: It would be much faster to
2 manually trip the breakers, wherever the switchboard
3 is for the scram breakers.
4 They are manually, I presume the scram
5 breakers are manually operable. Is that correct,
6 NuScale?
7 MR. BOWMAN: Yes, we can manually open the
8 reactor trip breakers.
9 MR. BROWN: Just like going to your power
10 panel in your house and tripping the breaker?
11 MR. BOWMAN: Well, little more involved
12 than that because it's --
13 MR. BROWN: I got it.
14 MR. BOWMAN: -- a three-phase, it's a
15 three-phase directable breaker that you have to go
16 push a button on it, have springs popping open.
17 MR. BROWN: Yes.
18 MR. BOWMAN: But in function, yes, it's
19 the same thing.
20 MR. BROWN: Okay.
21 CHAIR KIRCHNER: But we're talking about
22 a scenario first of all, that requires evacuation of
23 the control room.
24 So if they follow their procedures, and
25 we're not reviewing procedures at this juncture, it's
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1 reasonable to expect that they can write a procedure
2 that directs the operators to the MPSs, and/or the
3 breakers to disable, disable isn't the right word, but
4 to trip the reactor protection system and the S-class.
5 I don't think that's an unreasonable or
6 unrealistic length of time to accomplish that task.
7 Is that a good assumption, Doug?
8 MR. BOWMAN: Yes. I would find it no
9 more, I mean, I am aware of an operatin g plant
10 where an operator from the control room had to
11 walk down stairs, and a short walk to get to the trip
12 breakers to trip them open. So to me, this is a
13 similar action; similar timeframe; similar
14 capability.
15 We're not asking, we're not trying to do
16 anything different than what's been done in the
17 industry in my 24 years of experience, for a very long
18 time.
19 MR. BROWN: Walt, can I make another
20 comment relative to that?
21 CHAIR KIRCHNER: Yes, you may.
22 MR. BROWN: I've gone back and looked at
23 the older diagram. In addition to the remote station,
24 at least in that diagram and I couldn't since I
25 obviously messed it up with my review somehow on this
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1 one.
2 But there are theoretically, manual,
3 safety manual control switches in the main control
4 room. So you can bypass the entire main, at least you
5 could.
6 And you could manually trip the scram
7 breakers literally from the main control room, with
8 the manual safety switches.
9 That's what was called out in the legend
10 and everything in the original design of this stuff.
11 And that part I couldn't, right now I was unable to
12 get the rev 1 of this latest version.
13 Somehow that was not in the documents in
14 my SharePoint when I went to SharePoint to get it. So
15 I don't, are those manual safety control switches
16 still available in the main control room?
17 MR. BOWMAN: Yes.
18 MR. BROWN: To bypass everything?
19 MR. BOWMAN: Yes, absolutely.
20 MR. BROWN: Okay.
21 MR. BOWMAN: That's what we would go,
22 that's how we would take the action prior to leaving
23 the control room.
24 MR. BROWN: Yes, if, that's, I guess you
25 can argue if the fire is, somebody's got to leave the
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1 room and on the way out they turn out the lights.
2 That's the way I would think about it. Is
3 that correct?
4 MR. BOWMAN: Yes, absolutely.
5 MR. BROWN: Okay.
6 So, there still is some manual capability
7 with if you do it, execute it from the main control
8 room before you evacuate?
9 MR. BOWMAN: Correct, and that's the way
10 the procedure is written and that's our intended path
11 to get to safe shutdown.
12 Those actions in the local are only there
13 as a contingency. If you haven't achieved safe
14 shutdown when you get to the remote monitoring
15 station, then you would dispatch operators to go do
16 that.
17 MR. BROWN: Yes.
18 CHAIR KIRCHNER: So, thank you, Doug.
19 MR. BROWN: I thought that might help as
20 long as we know. Because I asked that question
21 earlier and it didn't come out very clear back earlier
22 in the meeting, so I apologize for that.
23 CHAIR KIRCHNER: I think we got the
24 answer, Charlie.
25 MR. BROWN: Okay, thanks, Walt.
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1 CHAIR KIRCHNER: Thank you. Okay, members
2 and consultants, any further questions on the NuScale
3 presentation for Chapter 18?
4 If not, let's transition to the staff. I
5 think we can hear the staff's evaluation of Chapter 18
6 and then take stock and see if we'll take the lunch
7 break, and then return for the rest of the agenda.
8 (Pause.)
9 CHAIR KIRCHNER: Okay, if you're ready to
10 proceed, go ahead. Is this Tom Hayden speaking?
11 Presenting?
12 MR. HAYDEN: Yes, that's correct.
13 CHAIR KIRCHNER: Yes, go ahead, Tom, if
14 you're ready.
15 MR. HAYDEN: Thanks.
16 Yes, this is Tommy Hayden. I'll be
17 presenting part of the Chapter 18 Human Factors
18 Engineering staff review slides.
19 Reviewers for this effort from the NRC
20 side, Amy D'Agostino, who is sitting to my left;
21 Maurin Sheetz, who is remote and online today;
22 Kamishan Martin, Brian Green, and then myself as the
23 project manager for the chapter; And Getachew Tesfaye
24 is the lead PM for the NuScale SDAA review.
25 All right, so these sections NuScale
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1 covered the sections in Chapter 18. They matched the
2 HFE, Human Factors Engineering, elements that you see
3 in NuReg-1711 and 12 there, and then 18.0 as the
4 overview for both in the submittal of the chapter, and
5 then in our SDA as well.
6 So for an overview, NuScale submitted
7 Chapter 18, Human Factors Engineering, Rev. 0 of the
8 NuScale SDAA FSAR on December 31, 2022, and Revision
9 1 on October 31, 2023.
10 The NRC audit of Chapter 18 was performed
11 from March 2023 to August 2023, generating 20 audit
12 issues, all of which were resolved in the audit.
13 And 12 of which resulted in NuScale
14 submitting supplemental information to address those
15 questions raised during the audit.
16 As a result, no RAIs were issued and the
17 staff completed their Chapter 18 review, and issued an
18 advance safety evaluation to support today's ACRS
19 Subcommittee meeting.
20 At this point, I'll turn it over to Maurin
21 Sheetz, to discuss the significant changes in the DCA
22 to the SDA. Maurin?
23 MS. SHEETZ: Thanks, Tommy. Hi, I'm
24 Maurin Sheetz, an NRC technical reviewer for Chapter
25 18. I was also one of the techn ical reviewers on
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1 chapter 18 for the NuScale design certification.
2 This table summarizes significant changes
3 in the area of human factors engineering from the
4 design certification, to the standard design
5 application.
6 So, in the area of main control room, the
7 design certification application included a design-
8 specific staffing requirement, in which six licensed
9 operators operate up to 12 reactor modules from a
10 single control room.
11 SDA for three operators to operate up to
12 six modules from a single control room. And that
13 includes the use of the NuScale control room staffing
14 plan, which the staff approved in 2021.
15 The design certification application
16 included two risk important human actions. For the
17 SDA, no human actions met thresholds for risk
18 significance, and none are credited in the Chapter 15
19 accident analysis.
20 During our audit, we learned that changes
21 made to the US 460 design altered the risk
22 significance of those two risk important human actions
23 from the DCA, and they are no longer risk significant
24 for the US 460 design.
25 The staff's review of important human
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1 actions for HFE interfaces with the reviews of
2 Chapters 7, 15, and 19, the latter two are still
3 ongoing reviews.
4 So, the staff will verify the status of
5 important human actions when these reviews are
6 complete.
7 As we already talked at length, here the
8 DCA included a remote shutdown station. The SCA does
9 not have a remote shutdown station, however, as was
10 the case for the DCA, the capability for remote
11 shutdown exists locally at the module protection
12 system cabinets.
13 And operators can monitor plant status at
14 alternate locations on site, alternate operator work
15 stations, which have very similar HSI design from the
16 main control room to, compared to the main control
17 room. Very similar.
18 Note that a designated facility for remote
19 shutdown is not required by general design criteria
20 19, and in the event of a main control room evacuate,
21 operators are expected to shut down each module from
22 the main control room before evacuating.
23 I'll add that during our review of the
24 Chapter 18 for the DCA, we observed an ISV scenario in
25 which operators conducted a main control room
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1 evacuation, and shut down 12 units from the main
2 control room before evacuating.
3 And we found that to be a successful path
4 for 12 units, so the SDA is for six units, so even
5 less to do there before evacuating.
6 And then, with regard to the applicant's
7 strategy for using HFE during the design process,
8 during the DCA, NuScale submitted what are called
9 results summary reports for each of the HFE program
10 elements.
11 Results summary reports summarize the
12 results of using NUREG-0711, which is the HFE program
13 review model.
14 So, they summarized using that model for
15 each HFE element, include a brief description of the
16 method used to achieve those results.
17 For the SDA, NuScale submitted
18 implementation plans for five of the review elements,
19 and results summary reports for two of the elements.
20 Those two are treatment of important human
21 actions, and staffing and qualifications.
22 Implementation plans describe the
23 methodology for conforming criteria, for conforming
24 two criteria NuReg-0711, and they're submitted for
25 work that's not complete at the time of the SDA.
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1 However, and NuScale has committed to
2 providing results summary reports before fuel load,
3 excuse me.
4 Okay, I will talk more about that on the
5 next slide, about the HFE strategy.
6 MR. BROWN: Can I ask a question?
7 MS. SHEETZ: Yes.
8 MR. BROWN: This is Charlie Brown again.
9 You talked about, I'm trying to go back to
10 your statement about remote location. You can monitor
11 all these remote like a technical support center, et
12 cetera, et cetera, et cetera.
13 And I presume, it looks like from the
14 other older diagrams, there's several locations where
15 you can do this remote monitoring.
16 In the versions we saw before and I only
17 had the non-proprietary diagram, architecture, to look
18 at, all of the data, if you go backwards in the DCA
19 and the earlier ones, all the data that went out to
20 these remote monitoring stations, was transmitted via
21 hardware-based non-unit directional-type data
22 transmission services.
23 And I just wanted to confirm that that is
24 still the methodology. Because that was back in the
25 earlier HIPS design itself, when it was determined
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1 that we would at least that design was approved with
2 the idea that any data going out of the main
3 processing, main control system, MPS, everything else,
4 would, any data, just data transmissions without
5 requiring control functions, would be handled with
6 unit directional-type data transmission.
7 Is that still the case?
8 MS. SHEETZ: So, this sounds like a
9 Chapter 7 type question, but what I can say, what I do
10 know is that these alternate operator work stations,
11 there's no control there.
12 There wasn't control there for safety
13 related systems in the DCA, and same thing, same exact
14 situation for the SDAA.
15 It's a non-safety systems control in the
16 main control room, but as far as --
17 (Simultaneous speaking.)
18 MR. TANEJA: Yes, this is Dinesh, Charlie.
19 MR. BROWN: Yes?
20 MR. TANEJA: So, there is no change to
21 that scheme of data diode. So there's still, it's the
22 same as what we had in the DCA.
23 MR. BROWN: Okay.
24 MR. TANEJA: So the data going out of that
25 network, plant network, is all going through one-way
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1 data communication using a one-way.
2 MR. BROWN: That's fine.
3 MR. TANEJA: Yes, so that scheme is the
4 same.
5 MR. BROWN: Okay, I was just trying to
6 make sure since I didn't have rev 1 of the proprietary
7 version to look at, and the other stuff was
8 abbreviated somewhat.
9 MR. TANEJA: Yes.
10 MR. BROWN: So I went back and looked at
11 it while we were doing this. So, I just wanted to
12 confirm we were still in the same data transmission
13 mode, that we were years ago.
14 MR. TANEJA: Correct, correct.
15 MR. BROWN: And your answer is yes.
16 MR. TANEJA: Yes, no change there.
17 MR. BROWN: Okay, I'm happy then. Thank
18 you.
19 MR. TANEJA: Yes.
20 MR. BROWN: Sorry about that, Walt.
21 CHAIR KIRCHNER: No, that's fine. Please
22 proceed.
23 MS. SHEETZ: Okay, Tommy, next slide.
24 Okay, this is Maurin again, and thank you Dinesh, for
25 taking that last question.
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1 So, this slide explains the overall
2 NuScale HFE program and their strategy. The staff
3 conducted an in-depth review of the NuScale HFE
4 program, as described in Chapter 18 of the SDA.
5 Six implementation plans, or IPs and two
6 results in reports, or RSRs. Those were submitted to
7 support Chapter 18 information. And we used the
8 guidance and criteria in Revision 3 of NUREG-0711 to
9 conduct our review.
10 Implementation plans describe a
11 methodology for completing a Human Factors Engineering
12 program element. They're submitted for work that's
13 not complete at the time of the SDA.
14 To determine whether an IP is acceptable,
15 we evaluate whether the implementation plan is
16 complete, detailed, and verifiable. And then
17 implementation plans must be followed by submittal of
18 a results summary report to show that the associated
19 activities are complete.
20 As I said before, the applicant has
21 committed to submitting results summary reports before
22 fuel load. And the staff can review them, if
23 necessary. For example, during the verification of
24 ITAAC closure.
25 Results summary reports summarize the
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1 results of the applicant's conformance to a particular
2 NUREG-0711 element, and include a brief description of
3 the methodology used to derive the results.
4 And we reviewed two results summary
5 reports for staffing and qualifications, and the
6 treatment of important human actions.
7 Procedure development, training program
8 development, and human performance monitoring are
9 designated as COL items.
10 And then finally, the HFE inspections,
11 tests, analysis, and acceptance criteria, or ITAAC,
12 ensures that all remaining HFE activities are complete
13 and produce adequate results.
14 So, for this, for the SDA, there are HFE
15 ITAAC. One is a requirement for verification and
16 validation of the main control room design, through
17 the performance of an inspection of the as-built
18 configuration of the main control room HSI.
19 And the second one is for a test called an
20 integrative system validation, which is used to ensure
21 that the final control room design culminating from
22 the combined results of various HFE activities,
23 supports the conclusion that operators can maintain
24 plant safety.
25 Acceptance criteria for the ISV tests is
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1 discussed in the applicant's verification, and
2 validation implement plan.
3 Together, these two HFE ITAAC confirm that
4 the final design has incorporated human factors
5 engineering principles, and minimizes the potential
6 for operator error.
7 Next slide, please. In conclusion, while
8 there are some differences between the DCA and the
9 SDA, the staff found that that applicant provided
10 sufficient information to support staff safety
11 finding.
12 And we found that all applicable
13 regulatory requirements were adequately addressed.
14 And that's all I have for the staff's review of
15 Chapter 18. I can take any questions.
16 CHAIR KIRCHNER: Members, consultants, any
17 questions?
18 MEMBER BIER: Yes, this is Vicki Bier. A
19 couple of questions. One, do you want to comment on
20 the same question I asked the applicant on the
21 staffing levels, and the justification for that based
22 on test results, et cetera?
23 MS. SHEETZ: Could you repeat the question
24 just so I?
25 MEMBER BIER: Sorry. When the applicant
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1 was presenting earlier, I asked about the
2 justification for the staffing level based on observed
3 performance on tasks, et cetera.
4 And do you just want to comment on that,
5 or you're satisfied with what they presented? Are
6 there any issues that you may want to think about?
7 MS. SHEETZ: Well, I will as far as
8 staffing goes, the staff, the NRC staff, this is
9 Maurin, we reviewed the staffing plan, which is two
10 SROs and one RO for up to, to operate up to 12
11 modules, as part of our review of a topical report,
12 NuScale control room staffing plan.
13 So that was done in 2021, so our safety
14 evaluation still stands there. And like I said,
15 that's three operators for up to 12 units.
16 So here we are for the SDA, three
17 operators for up to six units. That certainly falls
18 within the scope of the applicability of that topical
19 report.
20 MEMBER BIER: Okay, and kind of asking you
21 to go beyond what the applicant is requesting now, if
22 somebody someday wanted to operate a plant with I
23 don't know, 24 units or whatever, I assume there would
24 need to be a new justification and analysis presented
25 for what the staffing levels would need to be?
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1 Would it follow a similar process to
2 what's been done already?
3 MS. SHEETZ: Yes, this is Maurin, NRC.
4 Yes, for that hypothetical situation of how, what's
5 the staffing needed for 24 units, that would have to
6 go through another review.
7 Staffing applies, 10 CFR 50.54(m) applies
8 to a COL. So the staff is going to look at this again
9 when the COL comes in, and has another opportunity
10 through the exemption process.
11 Because nobody can meet 50.54(m) as its
12 written for this type of scheme of more than three
13 units.
14 So, that review would be done again and
15 the staff would most likely follow the process in
16 NUREG, I think it's 1781 for how to process an
17 exemption from, for staffing. How to look at
18 alternate staffing models.
19 MEMBER BIER: And again, any comment on
20 the lack of a remote shutdown station, that there's
21 adequate ways for people to perform any tasks that
22 are needed?
23 Or there's not anticipated to be any
24 needed tasks so?
25 MS. SHEETZ: I'll just say, this is
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1 Maurin. So on the topic of remote shutdown station,
2 obviously we noticed that there was one for the DCA;
3 there's not one for the SDA.
4 However, there was never a capability even
5 in the DCA, to shut down the reactors from a remote
6 shutdown station. It was just for monitoring.
7 MEMBER BIER: Got it.
8 MS. SHEETZ: So, GDC 19 requires the
9 capability for remote shutdown. That has always been
10 both for the DCA and the SDA, locally at the module
11 protection systems.
12 But the expectation is operators shut down
13 the units before they evacuate.
14 MEMBER BIER: Okay, thank you for the
15 clarifications. I don't know if other people have
16 questions or comments.
17 MEMBER DIMITRIJEVIC: I have one general
18 question.
19 Just so I was wondering after we, the
20 Chapter 15 and 19, the conclusion about important
21 human actions change.
22 In your opinion, would that just require
23 minimum change to this section like just the part on
24 the important human actions, or require some
25 additional viewing of the staffing procedures, the
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1 training, and things like that?
2 I mean, if this change, if important
3 actions are going to, will be added back to it, in
4 your opinion, is this section ready for it?
5 MS. SHEETZ: So, this is Maurin. That's
6 a good question.
7 So, our review of the SDA is set up so we
8 have an opportunity to alter our safety evaluation if
9 there's any important human actions that come out of
10 the other chapter reviews.
11 So, but that's like confirmatory items.
12 So we would go back into that. If an operator action
13 whether it's risk important or deterministic, is
14 determined, we would expect the applicant would
15 validate, would update their documents here.
16 For example, the treatment of important
17 human actions, or at some point, show us how the
18 operators can safely perform those important actions
19 with their, the human system interfaces in the main
20 control room. Or wherever they exist.
21 So, and we would also expect that during
22 the integrated system validation test, that those
23 would be tested. And they have to be tested during
24 ISV.
25 So, there are other, there's other work
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1 that would have to be done on both NuScale and the
2 staff's, and the NRC staff's review of those.
3 MEMBER DIMITRIJEVIC: All right, thanks.
4 CHAIR KIRCHNER: Members, any further
5 questions?
6 (No response.)
7 CHAIR KIRCHNER: Okay.
8 MR. BOWMAN: Oh, Walt, can you hold on for
9 a minute?
10 CHAIR KIRCHNER: Yes.
11 MR. BOWMAN: This is Doug Bowman from
12 NuScale. I just wanted to make one correction to a
13 statement that I believe was made during our
14 presentation.
15 And I'm not sure who it was, but somebody
16 had mentioned that there was an uninterruptible power
17 supply as part of MPS.
18 There is no uninterruptible power supply
19 as part of MPS. It is solely powered from EDAS, those
20 highly reliable batteries.
21 And so, to turn off the chassis, to get
22 direct trip well, we have to simply open two breakers,
23 one from two trains of EDAS and we could accomplish a
24 trip that way.
25 So, just to make sure that's clear on the
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1 record. Thank you.
2 CHAIR KIRCHNER: Okay, thank you, Doug.
3 Okay, this completes the chapter presentations and
4 evaluations that were scheduled for today.
5 Mike, I need to confer with you. We are
6 at the noon hour in East Coast time. The open
7 presentation on the high impact technical items is
8 relatively short, isn't it?
9 MR. SNODDERLY: One slide. Well, it's one
10 slide but it depends on how you want to discuss it.
11 I'm glad you brought this up. If I could just --
12 (Simultaneous speaking.)
13 MR. SNODDERLY: Okay, go ahead.
14 CHAIR KIRCHNER: Let me just say where I'm
15 going. I would like to do the following. I would
16 like to have, complete our open presentations and
17 allow opportunity for public comment. And then take
18 our lunch break and go to our closed session with the
19 expectation we're not going to return to another open
20 session, and ask the public to just stand by for
21 several hours.
22 So, if we can conclude our open portion of
23 the meeting and any deliberation by the members, and
24 opportunity for public comment, then we could take a
25 lunch break and go to our closed session.
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1 MR. SNODDERLY: So, I'm on the same page
2 but I would only add this. And I'd like us to stay in
3 open session until after lunch because I'd like the
4 members and you, to consider during lunch this, this,
5 where we're at as far as next steps in the September
6 full committee.
7 Right now, there are no planned
8 presentations on this matter, meaning chapters, the
9 review of Chapters 7, 9, 12, and 18.
10 CHAIR KIRCHNER: Right.
11 MR. SNODDERLY: Okay, based on that this
12 was a delta review and any issues identified.
13 Now, if there is a need for further
14 discussion at the September full committee, that needs
15 to be identified now and in open session.
16 So, I'd like, so I guess what I'm saying
17 is I would like some member discussion after lunch to
18 clarify what if any, support, or what would be needed
19 to support the September full committee meeting.
20 Because right now, the idea was that these
21 letters would be fairly clean and would not require
22 formal presentations.
23 Of course, the staff of NuScale will be
24 there to support the memo writing, but no
25 presentation.
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1 So if that needs to change, meaning
2 further presentations upon issues identified, then
3 that needs to be discussed and put into place now.
4 So I guess what I'm saying is, think about
5 it at lunch and if there's an issue that needs to be
6 delved into further --
7 CHAIR KIRCHNER: All right, so let me
8 outline my expectations and then if any member has
9 additional requirements. My expectation is we will
10 have the chapter write-ups from each lead member for
11 our P&P session in September. Is that correct, Mike?
12 MR. SNODDERLY: Yes, sir, that's the
13 current plan unless.
14 CHAIR KIRCHNER: Right, so, members and
15 consultants, if that, if you feel the need, you can
16 think about it over our lunch break, for further
17 presentations, we need to make that decision and
18 recommendation coming out of this open session, and
19 then go from there to the closed session.
20 MR. MOORE: That's correct, Walt. This is
21 Scott.
22 CHAIR KIRCHNER: Yes. So, I'll ask --
23 I'll have to speak on behalf of Dave Heddy, who is not
24 with us today, unfortunately. But for the other three
25 leads, do you see at this juncture, a need for further
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1 information or presentations from the applicant,
2 and/or the staff for your particular chapters?
3 I'll start with you, Tom.
4 MEMBER ROBERTS: Yes, I think I'd echo
5 something that Vesna said, which is I think there's a
6 potential for some of the lay reviews to influence
7 this write up.
8 And so, whatever we present or have
9 available in two weeks, I assume the intent is that to
10 be the current status that could change when for
11 example, Chapters 15 and 19 get reviewed.
12 CHAIR KIRCHNER: Exactly. We always have
13 that, that challenge when we do a serial review, and
14 we don't have some of the more important chapters
15 still to come, in particular 15. And we haven't had
16 4, 5, or 6, which are key chapters.
17 So, if you, my recommendation for our
18 letters is that if there's an issue that you think may
19 be impacted by a downstream review, be it Chapter 15
20 or the other core chapters, flag that not necessarily
21 as a concern, but just as an item to pick up in a
22 subsequent review.
23 That would be my recommendation for the
24 chapter leads, in terms of their letter report at this
25 juncture.
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1 MEMBER ROBERTS: Okay, that, yes.
2 CHAIR KIRCHNER: We had the same challenge
3 when we did this for the DCA, and did chapter write-
4 ups. We did them serially. Chapter 15, Chapter 19,
5 they came much later in the process.
6 So, we, if you will, we left pointers in
7 our letters to potentially matters that should be
8 looked at in conjunction with that subsequent review.
9 MEMBER ROBERTS: Okay, good. So with that
10 caveat, I don't see a need for anymore presentations
11 in September.
12 CHAIR KIRCHNER: Matt, from your
13 perspective?
14 MEMBER SUNSERI: So, Walt, the only thing
15 I see open right now is our understanding of the
16 impacts of this ultimate heat sink water level, on the
17 integrity of the containment vessel.
18 I think we're going to talk about that in
19 the closed session, and I anticipate that we'll have
20 a better understanding then.
21 The applicant appears to be convinced that
22 it's okay, and they just are unable to talk about it
23 because of the proprietary information.
24 So, outside of that issue, I don't see
25 anything else that we will want presented at the full
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1 committee meeting, and I'll have a memo prepared to
2 discuss for that meeting.
3 MEMBER DIMITRIJEVIC: Matt?
4 MEMBER SUNSERI: Yes, Vesna?
5 MEMBER DIMITRIJEVIC: Matt, you know in
6 this charging thing, in this flow reduction, things
7 like that can change in this containment is a
8 relation.
9 And maybe that could be part of the,
10 something which we will learn more later what's
11 happening.
12 MEMBER SUNSERI: Yes, and that's a good
13 point, Vesna.
14 I didn't bring that up because they're
15 talking, I guess my understanding is, is that's an
16 operational matter, not a system configuration issue
17 at this point.
18 So we're going to talk about it in
19 chapter, some other Chapter 15, I think is where
20 that's probably going to be --
21 (Simultaneous speaking.)
22 CHAIR KIRCHNER: Well, I think the
23 hardware aspects will be addressed in Chapter 6.
24 MEMBER SUNSERI: 6.
25 CHAIR KIRCHNER: And then, the safety
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1 analysis will be addressed in 15. And that should be
2 reflected, Vesna, I think in Chapter 19, as well.
3 MEMBER DIMITRIJEVIC: Right, right. But
4 I wasn't sure what part, will any part go in Chapter
5 9 versus Chapter 6.
6 Yes, I'm just bringing this up because it
7 was introduced in those flow reductions was introduced
8 as a part of Chapter 9.
9 MEMBER SUNSERI: I think it's no doubt
10 that's something that we want to know more about.
11 It's just not related to Chapter 9 is all I'm saying,
12 yes.
13 MEMBER DIMITRIJEVIC: Okay.
14 CHAIR KIRCHNER: Okay.
15 I addressed what was the major issue that
16 we had with Chapter 12. It was the combustible gas
17 monitoring. So, I don't want to repeat what the
18 statement I made earlier.
19 And then, Vicki, on Chapter 18, what's
20 your position?
21 MEMBER BIER: Yes, I don't see a need for
22 further presentations in September. Seems pretty
23 straightforward at this point.
24 CHAIR KIRCHNER: Okay.
25 So then, my expectation is we might would
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1 then have letters that have pointers in them, if there
2 are concerns, or open matters.
3 But we would have those letters for the
4 September full committee meeting.
5 MR. SNODDERLY: Thank you, Walt, that was
6 very helpful and I think the staff at NuScale will
7 appreciate that, too.
8 CHAIR KIRCHNER: Yes.
9 MR. SNODDERLY: Just to know what the
10 expectations are for September. And thank you for
11 having this discussion.
12 I still would suggest that, well, now the
13 question is, do you want to go through the high impact
14 technical issue open session, or do we take a break
15 and maybe let the members think a little bit more
16 about everything they took from this morning?
17 CHAIR KIRCHNER: Yes, let's take a break.
18 We've gone for over two hours at this point. We'll
19 take perhaps a shortened lunch break.
20 Is reconvening at 1300, at 1:00 o'clock
21 Eastern Time acceptable?
22 MR. SNODDERLY: Yes, I think that would
23 be, and we'll be in open session. We'll let NuScale
24 do their open presentation on the HITIs, and then take
25 public comment.
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1 CHAIR KIRCHNER: Right.
2 MR. SNODDERLY: Close the session and go
3 into closed session, and not go back into open
4 session. Is that?
5 CHAIR KIRCHNER: Excellent, yes. That's
6 the plan.
7 Okay, so with that, we will recess until
8 1:00 o'clock Eastern Time. And I thank all the
9 presenters, and I once again thank the NuScale people
10 for such an early start.
11 Thank you for your presentations. Thank
12 you also to the staff.
13 We are in recess until 1:00 o'clock
14 Eastern Time.
15 (Whereupon, the above-entitled matter went
16 off the record at 12:13 p.m. and resumed at 1:01 p.m.)
17 CHAIR KIRCHNER: Okay, we are back in
18 session. This is a meeting of the NuScale
19 Subcommittee, and we are going to turn to NuScale for
20 a presentation and discussion on their high impact
21 technical issues.
22 And I'll turn to Thomas Griffith.
23 MR. GRIFFITH: Thank you, I'm Thomas
24 Griffith, Licensing Manager at NuScale. Pleased to
25 have the opportunity to provide an update on the US-
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1 460 standard design approval application, high impact
2 technical issues.
3 You may hear me use the phrase HITI.
4 That's the shorthand way that we've been referencing
5 the high impact technical issues, just for some
6 clarity.
7 I do have with me Kris Cummings, do you
8 want to introduce yourself?
9 MR. CUMMINGS: Yes, Kris Cummings,
10 licensing engineer for NuScale. I've been with
11 NuScale for about four and a half years now. Prior to
12 that, I've had several, I've had roles with
13 Westinghouse, Holtec, and NEI. My focus is generally
14 nuclear-related stuff, but they seem to come to me
15 with it whenever they get stuck on things.
16 So, I'm here to support Tom.
17 MR. GRIFFITH: Thanks, Kris.
18 Next slide, please. Appreciate the award
19 from the Department of Energy, and appreciate their
20 support in helping with NuScale with our mission.
21 Next slide, please. So, the HITIs are
22 identified as specific topic areas that NRC and
23 NuScale management have agreed, require elevated
24 management attention.
25 The use of the term HITI does not imply a
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1 shortfall in the application, rather that there is a
2 need to prioritize the resolution of the issue to
3 support overall review schedule.
4 The HITI list in NS's opinion, is close to
5 resolution because the NRC has provided NuScale audit
6 items and RAIs, related to the HITIs.
7 Regulatory basis and acceptance criteria
8 have been established, and due dates for final
9 products to provide during the review, is largely
10 understood.
11 To date, 10 high impact technical issues
12 have been established. Three were considered resolved
13 at the least quarterly management meeting between
14 NuScale, and the NRC. That took place at the end of
15 July.
16 Due to the hard work of the NRC and
17 NuScale, two additional HITIs, HITI-1, the design and
18 classification of the augmented DC power system, as
19 well as HITI-3, are now considered resolved, as well.
20 I would like to defer to the closed
21 session for a detailed discussion of each of the
22 HITIs, as much of the detail could be considered
23 proprietary.
24 CHAIR KIRCHNER: Okay, thank you, Thomas.
25 At this point, I think members and
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 151
1 consultants, we can hold detailed questions. Is there
2 anything, Thomas, that you can say?
3 You indicated you have now five of these
4 resolved. Is there anything that you want to say on
5 NuScale's behalf about items 4, 5, and 9 in an open
6 session?
7 MR. GRIFFITH: I think for items 4, 5, and
8 9, they're all related to material changes that
9 NuScale has implemented with the containment vessel,
10 the RPV vessel, respectively.
11 And item number 9 is a result of the upper
12 and lower RVB material being different. The NRC as a
13 result of some of the quarterly management meetings
14 that we had, identified the need to review some of the
15 detailed calculations from NuScale regarding the shear
16 loading of results from differential thermal
17 expansion.
18 And to that end, we believe that items 4,
19 5, and 9 are largely understood, and we do anticipate
20 being able to provide detailed presentations on each
21 of those items when we get to the chapters for ACRS.
22 CHAIR KIRCHNER: Okay. And there are no
23 open ASME code issues with 4 and 5?
24 MR. GRIFFITH: Not that I'm aware of.
25 CHAIR KIRCHNER: Yes, just wanted to put
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 152
1 that on the record.
2 Okay, all right with that then, I think
3 you only had one slide, Thomas, is that correct?
4 MR. GRIFFITH: That is correct.
5 CHAIR KIRCHNER: Just the list, yes.
6 So, Mike at this point, I think we can
7 turn to the public and ask for any comments from the
8 public, and proceed to that.
9 MR. SNODDERLY: Yes, Walt.
10 So, are there any members of the public
11 that would like to make a comment on today's meeting?
12 MEMBER MARTIN: There appears to be one.
13 Tim Polich, go ahead and unmute yourself and make your
14 comment.
15 MR. POLICH: Yes, can you hear me?
16 MEMBER MARTIN: Yes.
17 MR. POLICH: Yes, I was encouraged to hear
18 that one of the carve-outs was closed today. Still
19 concerned with the other, the carve-outs that were
20 left from the last approved design.
21 And I see a lot of work that looks like it
22 needs to be done, because I've been monitoring not
23 only the ADAMS, but also the SDA review dashboard.
24 I see responses seem to come in slowly.
25 The average response time seems to be rather high. It
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 153
1 appear the REIs, they seem to either need multiple
2 responses or something because they go from awaiting
3 response to reviewing, back to awaiting response.
4 And I'm just concerned that maybe there
5 needs to be some more testing that needs to get done
6 at this, and that just seems a little late in the game
7 to be doing that.
8 But I'm still concerned, and it's probably
9 some of the same concerns that I believe it's Dr.
10 Dimitrijevic had in 2020 about the steam generators,
11 and still haven't seen anything that resolves there.
12 So, that's my comment. Thank you.
13 MEMBER MARTIN: Okay, are there any
14 others?
15 Hearing none, or seeing no hand raised,
16 Walt, I'll return it to you to close.
17 CHAIR KIRCHNER: Okay, thank you for
18 monitoring that, Bob.
19 Thank you for the public comment, Tim, and
20 with that, we are at a juncture where we can adjourn
21 the open session and for those people authorized to
22 attend, go to our Teams link for the closed session.
23 So once again, I want to thank NuScale for
24 agreeing to such an early start, and I thank you for
25 your presentations as well. And thank you to the
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 154
1 staff, likewise.
2 So with that, we are adjourned and for
3 those who have permission, please go to the Teams link
4 for the closed meeting. Thank you.
5 (Whereupon, the above-entitled matter went
6 off the record at 1:09 p.m.)
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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 LO-172559
August 19, 2024 Docket No. 052-050
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) US460 Standard Design Approval Application Chapters 7, 9, 12, and 18, PM-172558, Revision 0
The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on August 22, 2024. The materials support NuScales presentation of the subject chapters of the US460 Standard Design Approval Application.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) US460 Standard Design Approval Application Chapters 7, 9, 12, and 18, PM-172558, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Chelsea Lockwood at 541-452-7171 or at clockwood@nuscalepower.com.
Sincerely,
Thomas Griffith Manager, Licensing NuScale Power, LLC
Distribution: Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC
- ACRS Subcommittee Meeting (Open Session) US460 Standard Design Approval Application Chapters 7, 9, 12, and 18, PM-172558, Revision 0
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com LO-172559
ACRS Subcommittee Meeting (Open Session) US460 Standard Design Approval Application Chapters 7, 9, 12, and 18, PM-172558, Revision 0
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com
Presentation to the Advisory Committee on Reactor Safeguards Subcommittee Staff Review of NuScales US460 Standard Design Approval Application Final Safety Analysis Report, Revision 1
Chapters 7, 9, 12, and 18
August 22nd, 2024 (Open Session)
1 Non-Proprietary Acronyms and Definitions
- ACRS - Advisory Committee on Reactor Safeguards* EMIB - Mechanical Engineering and Inservice
- APLB - PRA Licensing Branch B Testing Branch
- APLC - PRA Licensing Branch C
- ESEB - Structural Civil Geotech Engineering Branch
- ARCB - Radiation Protection and Consequences
- FSAR - Final Safety Analysis Report Branch
- GDC - General Design Criteria
- CFR - Code of Federal Regulations
- NCSG - Corrosion and Steam Generator Branch
- COL - Combined License
- NLIB - Licensing and Regulatory Infrastructure
- COLA - Combined License Application Branch
- DC - Design Certification (refers to NuScale US600
- NPM - NuScale Power Module design)
- NRLB - New Reactor Licensing Branch
- DCA - Design Certification Application (refers to
- NRR - Office of Nuclear Reactor Regulation NuScale US600 design)
- DEX - Division of Engineering and External Hazards* SCPB - Containment and Plant Systems Branch
- DNRL - Division of New and Renewed Licenses
- SDA - Standard Design Approval (refers to NuScale
- DRA - Division of Risk Assessment US460 design)
- DSS - Division of Safety Systems
- SDAA - Standard Design Approval Application
- EEEB - Electrical Engineering Branch (refers to NuScale US460 design)
- EICB - Instrumentation and Controls Branch
- SNRB - Nuclear Methods Systems and New
- ELTB - Long Term Operations and Modernization Reactors Branch Branch
2 Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1
Chapter 7, Instrumentation and Controls"
August 22, 2024 (Open Session)
3 Non-Proprietary NuScale SDAA FSAR Chapter 7 Review Overview
- NuScale submitted Chapter 7, Instrumentation and Controls, Revision 0 of the SDAA FSAR on December 31, 2022, and Revision 1 on October 31, 2023
- NRC regulatory audit of Chapter 7 was performed from March 2023 to August 2023
- One audit issue was issued and resolved in the audit
- The audit issue resulted in NuScale submitting supplemental information to address questions raised during the audit
- No RAIs issued
- Staff completed Chapter 7 review and issued an advanced safety evaluation to support today s ACRS Subcommittee meeting
4 Non-Proprietary NuScale SDAA FSAR Chapter 7 Review Contributors
- Technical Reviewers
- Joseph Ashcraft, NRR/DEX/EICB
- Dinesh Taneja, NRR/DEX/ELTB
- Project Managers
- Ricky Vivanco, PM, NRR/DNRL/NRLB
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB
5 Non-Proprietary NuScale SDAA FSAR Chapter 7 Review Sections
- Section 7.0 - Introduction and Review Process
- Section 7.1 - Instrumentation and Controls - Fundamental Design Principles
- Section 7.2 - Instrumentation and Controls - System Characteristics
6 Non-Proprietary NuScale SDAA FSAR Chapter 7 Review Significant Changes from DCA to SDA
- NPM power uprate and safety analysis resulted in changes to Reactor Trip Setpoints & Engineered Safety Feature Actuation System actuation logic and analytical limits
- Setpoint methodology TR-122844 -P documents the new analytical limits and setpoints
- Common Cause Failure coping analysis revised due to reductions in digital sensors
- Added Type F Post-Accident Monitoring variables
7 Non-Proprietary NuScale SDAA FSAR Chapter 7 Review Conclusion
- While there are some differences between the DCA and SDAA, the staff found that the applicant provided sufficient information to support the staff s safety finding
- The staff found that all applicable regulatory requirements were adequately addressed
8 Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1
Chapter 9, Auxiliary Systems"
August 22, 2024 (Open Session)
9 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Overview
- NuScale submitted Chapter 9, Auxiliary Systems, Revision 0 of the NuScale S DA A FSAR on December 31, 2022, and Revision 1 on October 31, 2023
- NRC regulatory audit of Chapter 9 was performed from March 2023 to August 2023, generating 33 audit issues
- 23 audit issues were resolved in the audit
- 10 audit issues resulted in NuScale submitting supplemental information to address questions raised during the audit
- 13 RAIs were issued
- Staff completed Chapter 9 review and issued an advanced safety evaluation to support today's ACRS Subcommittee meeting
10 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Contributors
- Technical Reviewers - Angelo Stubbs, NRR/DSS/SCPB
- Thinh Dinh, NRR/DRA/APLB - Brian Lee, NRR/DSS/SCPB
- Daniel Ju, NRR/DRA/APLB - David Nold, NRR/DSS/SCPB
- Naeem Iqbal, NRR/DRA/APLB - Gordon Curran, NRR/DSS/SCPB
- Marie Pohida, NRR/DRA/APLC - Nageswara Karipineni, NRR/DSS/SCPB
- Steven Alfernik, NRR/DRA/APLC - Raul Hernandez, NRR/DSS/SCPB
- Sunwoo Park, NRR/DRA/APLC - Joshua Miller, NRR/DSS/SNRB
- Edward Stutzcage, NRR/DRA/ARCB - Ryan Nolan, NRR/DSS/SNRB
- Adakou Foli, NRR/DEX/EEEB - Zhian Li, NRR/DSS/SNRB
- Sheila Ray, NRR/DEX/EEEB
- Project Managers
- Dinesh Taneja, NRR/DEX/ELTB - Greg Cranston, PM, NRR/DNRL/NLIB
- Joseph Ashcraft, NRR/DEX/EICB - Getachew Tesfaye, Lead PM,
- Nicholas Hansing, NRR/DEX/EMIB NRR/DNRL/NRLB
- Thomas Scarbrough, NRR/DEX/EMIB
- Ata Istar, NRR/DEX/ESEB
- George Wang, NRR/DEX/ESEB
- Matthew Yoder, NRR/DNRL/NCSG
11 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Sections
- 9.1 Fuel Storage and Handling Systems
- 9.2 Water Systems
- 9.3 Process Auxiliaries
- 9.4 Air Conditioning, Heating, Cooling, and Ventilation Systems
- 9.5 Other Auxiliary Syste m s
- Appendix 9A Fire Hazards Analysis
12 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Significant Changes from DCA to SDA
- Fire barrier rating
- In the DCA steel-composite material was limited to staircase walls with a required 2-hour rated fire barriers.
- In the SDA the reactor building, control building, and radioactive waste building floors, walls and ceilings to be made almost entirely of reinforced concrete or steel composite walls to provide a 3 -hour fire barrier rating.
- The SDA only houses six NPM and therefore reactor building size reduced compared to that of the 12 NPM DCA design.
- Related parameters used in the analysis of the spent fuel pool (SFP) and ultimate heat sink (UHS) cooling analysis changed accordingly.
- Reactor building crane rated capacity increased; modular interface incorporated into crane design; dry dock jib crane added for refueling.
13 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Significant Changes from DCA to SDA
- SDA design
- Consolidated SFP cooling and cleanup system and combined the UHS cooling systems into a single system that cools both the SFP and UHS.
- UHS pool water level is lower
- UHS operating temperatures increased
- Fuel storage rack design and analysis assigned to COL applicant
- Increased core thermal power which will impact SFP and UHS cooling due to increased heat loads associated with the NPMs and the spent fuel assemblies
14 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Significant Changes from DCA to SDA
- SDA Chemical and Volume Control System (CVCS) reconfiguration of certain valves outside containment required compensating design changes. Multiple CVCS valves moved out of NPM bay. Valves included internal restrictions credited for CVCS line breaks outboard of containment isolation valves. Flow restricting venturis added into containment vessel nozzles inside containment for the injection and discharge lines. Flow restricting venturis credited in both Chapter 15 and Chapter 19 events.
15 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Significant Changes from DCA to SDA
- SDA reactor building crane - risk significant design change from DCA to SDA to include a Reactor Building Crane Control System. Reliability of programmable logic controller and associated components results in operator errors being negligible contributors to module drop.
- In DCA, over 95% of the core damage frequency driven by module drop. Calculated drop probability dominated by operator errors of commission (e. g., overspeed, over-raise, overtravel and failure of instrumentation (interlocks/limit switches).
- The application does not include specific SFP criticality safety design information and corresponding criticality safety analyses for the SDA. This is addressed with two COL items requiring COL applicants to perform criticality safety analyses for the new fuel and spent fuel pool.
16 Non-Proprietary NuScale SDAA FSAR Chapter 9 Review Conclusion
- While there are some differences between the DCA and the SDA, the staff found that the applicant provided sufficient information to support the staff s safety finding.
- The staff found that all applicable regulatory requirements were adequately addressed.
17 Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1
Chapter 12, Radiation Protection"
August 22, 2024 (Open Session)
18 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Overview
- NuScale submitted Chapter 12, Radiation Protection, Revision 0 of the NuScale SDAA FSAR on December 28, 2022, and Revision 1 on October 31, 2023
- NRC regulatory audit of Chapter 12 performed March 2023 to August 2023, generating 13 audit issues
- 11 audit issues were resolved in the audit
- 9 audit issues resulted in NuScale submitting supplemental information to address questions raised during the audit
- 2 RAIs were issued and resolved
- Staff completed Chapter 12 review and issued an advanced safety evaluation to support today's ACRS Subcommittee meeting
19 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Contributors
- Technical Reviewers
- Edward Stutzcage, NRR/DRA/ARCB
- Project Managers
- Alina Schiller, PM, NRR/DNRL/NRLB
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB
20 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Sections
- Section 12.1 - Assuring that Occupational Radiation Exposures Are As Low As Is Reasonably Achievable
- Section 12.2 - Radiation Sources
- Section 12.3 - Radiation Protection Design Features
- Section 12.4 - Dose Assessment
- Section 12.5 - Operational Radiation Protection Program
21 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Significant Changes from DCA to SDA
- The DCA rule includes an open item to be addressed by the COLA related to the radiological dose consequences to workers performing activities associated with hydrogen and oxygen monitoring following a core damage accident and potential dose consequences to the control room workers and the public resulting from activities associated with hydrogen and oxygen monitoring
- Compliance with 10 CFR 50.34(f )(2)(xxviii), 10 CFR 50.34(f )(2)(vii), and 10 CFR 52.47(a)(2)(iv)
- The SDAA includes a passive autocatalytic recombiner and NuScale has requested an exemption from the combustible gas monitoring requirements of 10 CFR 50.44(c)(4)10 CFR 50.34(f )(2)(xvii)(C)
- The staff is reviewing the exemption under FSAR Chapters 6 and 19
- If the exemption is approved, hydrogen and oxygen monitoring is unnecessary, and the associated radiological implications need not be accessed
22 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Significant Changes from DCA to SDA
- Source terms
- Source terms throughout Chapter 12 were updated to account for changes in reactor power, the number of units, and other less significant design changes
- Methodology for developing source terms remained mostly unchanged
- Shielding and dose rates throughout the facility were adjusted accordingly
- Staff reviewed the updated source terms, shielding, and zoning for significant radiation sources and found them to be acceptable
23 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Significant Changes from DCA to SDA
- The reactor coolant travels through the RCS significantly faster in the SDAA, impacting the N-16 source term
- Results in significant N-16 concentration remaining in CVCS line exiting the reactor module bays during operation
- NuScale adequately accounted for N-16 in shielding and dose calculations until N-16 reaches 10 half -lives, at which time N-16 doses are insignificant
- Staff reviewed the N-16 concentrations and CVCS system shielding and found them to be acceptable
24 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Significant Changes from DCA to SDA
- Concrete Equivalent Attenuation Shielding
- Radiation shielding is specified in terms of nominal concrete attenuation thicknesses.
- FSAR specifies that materials used in place of the specified concrete provides the equivalent attenuation as the prescribed concrete shielding.
- Alternative radiation shielding must meet the prescribed radiation zoning provided in FSAR Chapter 12 and must be verified to comply with all regulatory requirements.
- Staff reviewed the shielding specifics provided in the FSAR and in audited shielding calculations and found them to be acceptable. Staff found the approach to be acceptable because it ensures that the zoning provided in Chapter 12 and applicable regulatory requirements continue to be met.
25 Non-Proprietary NuScale SDAA FSAR Chapter 12 Review Conclusion
- While there are some differences between the DCA and SDAA, the staff found that the applicant provided sufficient information to support the staff s safety findings.
- The staff found that all applicable regulatory requirements were adequately addressed for the design.
26 Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1
Chapter 18, Human Factors Engineering"
August 22, 2024 (Open Session)
27 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review Contributors
- Technical Reviewers
- Amy D'Agostino, NRR/DRO/IOLB
- Maurin Scheetz, NRR/DRO/IOLB
- Kamishan Martin, NRR/DRO/IOLB
- Brian Green, NRR/DRO/IOLB
- Project Managers
- Thomas Hayden, PM, NRR/DNRL/NRLB
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB
28 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review Sections
- Section 18.0 - Human Factors Engineering Overview
- Section 18.1 - Human Factors Engineering Program Management
- Section 18.2 - Operating Experience Overview
- Section 18.3 - Functional Requirements Analysis and Function Allocation
- Section 18.4 - Task Analysis
- Section 18.5 - Staffing and Qualifications
- Section 18.6 - Treatment of Important Human Actions
- Section 18.7 - Human System Interface Design
- Section 18.8 - Procedure Development
- Section 18.9 - Training Program Development
- Section 18.10 - Human Factors Verification and Validation
- Section 18.11 - Design Implementation
- Section 18.12 - Human Performance Monitoring
29 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review Overview
- NuScale submitted Chapter 18, Human Factors Engineering, Revision 0 of the NuScale SDAA FSAR on December 31, 2022, and Revision 1 on October 31, 2023
- NRC regulatory audit of Chapter 18 performed March 2023 to August 2023, generating 20 audit issues
- All issues were resolved in the audit.
- 12 issues resulted in NuScale submitting supplemental information to address questions raised during the audit
- No RAIs were issued
- Staff completed Chapter 18 review and issued an advanced safety evaluation to support today's ACRS Subcommittee meeting
30 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review Significant Changes from DCA to SDA Area DCA S DA A Main Control Room 12 units 6 units S taffing 6 operators 3 operators Important Human Actions 2 risk important actions No important human actions Other design changes Remote Shutdown Station No remote shutdown station HFE strategy Results Summary Reports and Implementation Plans for:
ITAAC for Design Operating Experience Review Implementation Functional Requirements Analysis & Function Allocation Task Analysis HSI Design Verification & Validation Design Implementation ITAAC for Integrated System Validation and Design Implementation
31 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review HFE Strategy
- The staff reviewed Implementation Plans for:
- Operating Experience Review
- Functional Requirement Analysis and Function Allocation
- Task Analysis RSRs will be available
- HSI Design before fuel load
- Human Factors Verification and Validation
- Design Implementation
- The staff reviewed Result Summary Reports for:
- Staffing and Qualifications
- Treatment of Important Human Actions
- Procedure development, training program development and human performance monitoring
- No. 03.15.01: the main control room HSI is consistent with design verified and validated by the integrated system validation including any changes reconciled during design implementation
- No. 03.15.02: for integrated system validation of the main control room design
32 Non-Proprietary NuScale SDAA FSAR Chapter 18 Review Conclusion
- While there are some differences between the DCA and SDAA, the staff found that the applicant provided sufficient information to support the staff s safety finding.
- The staff found that all applicable regulatory requirements were adequately addressed.
33 Non-Proprietary LO-172563
August 19, 2024 Docket No. 052-050
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session), August 22, 2024, High Impact Technical Issues Discussion, PM-173236, Revision 0
The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on August 22, 2024. The materials support NuScales presentation of the high impact technical issues identified during the US460 Standard Design Approval Application review.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) High Impact Technical Issues Discussion, PM-173236, Revision 0.
The proprietary version is provided in a separate submittal, under letter number LO-173238.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Chelsea Lockwood at 541-452-7171 or at clockwood@nuscalepower.com.
Sincerely,
Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC
Distribution: Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC
- ACRS Subcommittee Meet ing (Open Session) High Impact Technical Issues Discussion, PM-173236, Revision 0
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com LO-172563
ACRS Subcommittee Meeting (Open Session) High Impact Technical Issues Discussion, PM-173236, Revision 0
NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com
Meeting title Open Session NuScale Subcommittee Meeting on Staff's Evaluation of Chapters 7, 9, 12 and 18
Participants
Michael Snodderly Shandeth W alton Alissa Neuhausen James Cordes Court Reporter Thomas Dashiell Jack Zhao Tammy Skov Sandra W alker Tim Polich Chelsea Lockwood NuScale Tyesha Bush Charlie Brown Ralph Costello Ata Istar Amanda Bode NuScale Thomas Scarbrough W endy Reid NuScale Seth Robison NuScale Ron Ballinger Scott Palmtag Justin Kramer W alt Kirchner Gordon Curran Steven Alferink Christina Antonescu David Nold Gregory Halnon Jim Schneider NuScale Rob Meyer NuScale P Leary Josh Bouma NuScale Chris Noack NuScale PJ Evans NuScale Bill Acton NuScale Kevin Lynn NuScale Derek W idmayer David Rickenbach NuScale Ricky Vivanco David Jackson Cindy W illiams NuScale Thomas Roberts Larry Hu NuScale Behnke, Donald H.
Rose Charoensombud NuScale Tania Martinez Navedo George W ang Matt Sunseri Jon Bristol NuScale Mahmoud -MJ-Jardaneh Mara Swanson NuScale Derek Noel NuScale River Rohrman Myron J Hecht Gene Eckholt Naeem Iqbal Troy Von Rentzell NuScale Prosanta Chowdhury Rob Krsek Levi Schimelfining NuScale Jay Robinson Spencer Toohill The Breakthrough Institute Vesna B Dimitrijevic Maurin Scheetz Tamela Cohen NuScale Amy D'Agostino Tyler Beck NuScale Robert Martin Joe Remic NuScale Russell Felts Adakou Foli David Drucker Greg Cranston Norbert Carte Joy Jiang Nicholas Hansing Andrea Mota NuScale Dan Serbanescu Thomas Hayden Elisa Fairbanks NuScale Augi Cardillo NuScale Daniel Ju James Anderson Raul Hernandez Jeff Correll Getachew Tesfaye Emily Larsen NuScale Leigh Lloveras The Breakthrough Institute Eric Lantz NuScale Justin Donohoe Brian Green Andrea Torres Sunwoo Park DaRo Amanda Bode NuScale Ryan Nolan Angelo Stubbs Dinesh Taneja Kevin Hsueh Joseph Ashcraft Mara Swanson NuScale W ren Fowler NuScale Trudy Mitchell NuScale Sarah Bristol NuScale Evan Davidson Ben Bristol NuScale Erin Blumsack NuScale Edward Stutzcage Emily Larsen NuScale Tony Nakanishi Vesna B Dimitrijevic Mark Shaver Daniel Diefendorf NuScale Emily Larsen NuScale Erin Blumsack NuScale Allyson Callaway NuScale Carlen Donahue NuScale J.J. Arthur W endell Morton Dennis Galvin Kris Cummings NuScale Tom Case NuScale Sarah Turmero NuScale Freda Ahmed NuScale Jim Osborn NuScale Doug Bowman NuScale Jordan Green NuScale Thomas Griffith NuScale Erik Slobe NuScale Barry Reichelderfer NuScale Milton Valentin Brian Lee Rebecca Patton Thinh Dinh Zhian Li Joshua Miller Marie Pohida