ML19211C771
| ML19211C771 | |
| Person / Time | |
|---|---|
| Issue date: | 06/18/2019 |
| From: | Michael Snodderly Advisory Committee on Reactor Safeguards |
| To: | |
| Snodderly, M, ACRS | |
| References | |
| NRC-0389 | |
| Download: ML19211C771 (499) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safegards NuScale Subcommittee: Open Session Docket Number: (n/a)
Location: Rockville, Maryland Date: Tuesday, June 18, 2019 Work Order No.: NRC-0389 Pages 1-316 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
1 1
2 3
4 DISCLAIMER 5
6 7 UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11 The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17 This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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1 1 UNITED STATES OF AMERICA 2 NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5 (ACRS) 6 + + + + +
7 NuSCALE SUBCOMMITTEE 8 + + + + +
9 OPEN SESSION 10 + + + + +
11 TUESDAY 12 JUNE 18, 2019 13 + + + + +
14 ROCKVILLE, MARYLAND 15 + + + + +
16 The Subcommittee met at the Nuclear 17 Regulatory Commission, Two White Flint North, Room 18 T2D10, 11545 Rockville Pike, at 8:30 a.m., Peter 19 Riccardella and Jose March-Leuba, Co-Chairs, 20 presiding.
21 COMMITTEE MEMBERS:
22 PETER RICCARDELLA, Co-Chair 23 JOSE MARCH-LEUBA, Co-Chair 24 RONALD G. BALLINGER, Member 25 DENNIS BLEY, Member NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 1 CHARLES H. BROWN, JR. Member 2 MICHAEL L. CORRADINI, Member 3 VESNA B. DIMITRIJEVIC, Member 4 JOY L. REMPE, Member 5 GORDON R. SKILLMAN, Member 6 MATTHEW W. SUNSERI, Member 7
8 ACRS CONSULTANT:
9 STEPHEN SCHULTZ 10 11 DESIGNATED FEDERAL OFFICIAL:
12 MIKE SNODDERLY 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 1 CONTENTS 2 Opening Remarks . . . . . . . . . . . . . . . . . 4 3 Chapter 3, Design of SSCs, Safety 4 Evaluation With Open Items . . . . . . . . . . 7 5 Overview of Chapter 6, Engineered Safety 6 Features, NuScale Design Certification 7 Application . . . . . . . . . . . . . . . . . 154 8 Public Comment . . . . . . . . . . . . . . . . 312 9 Adjourn . . . . . . . . . . . . . . . . . . . . 316 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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4 1 P R O C E E D I N G S 2 8:29 a.m.
3 MEMBER CORRADINI: Okay, why don't we have 4 the meeting come to order. This is a meeting of the 5 ACRS NuScale Subcommittee. My name is Mike Corradini, 6 Chairman of the NuScale Subcommittee. Members 7 currently in attendance today are Vesna Dimitrijevic, 8 Dennis Bley, Charlie Brown, Jose March-Leuba, Joy 9 Rempe, Matt Sunseri, Pete Riccardella, Gordon Skillman 10 and Ron Ballinger, and our consultant, Dr. Steve 11 Schultz. I did.
12 The Subcommittee will review the staff's 13 evaluation of Chapter 3, Design of SSCs, Chapter 6, 14 Engineered Safety Features, and Chapter 15, Transient 15 Accident Analysis of the NuScale Design Certification 16 Application and NuScale's Topical Report, TRO-17 51649417, Evaluation and Methodology of Stability 18 Analysis.
19 Today, we have members of the NRC staff 20 and NuScale to brief the Subcommittee. The ACRS was 21 established by statute and is governed by the Federal 22 Advisory Committee Act or FACA. That means that the 23 committee can only speak through its published letter 24 reports. We hold meetings to gather information to 25 support our deliberations, and interested parties who NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 1 wish to provide comments can contact our office 2 requesting time after the meeting announcement is 3 published in the Federal Register.
4 That said, we set aside ten minutes for 5 extemporaneous comments from members of the public 6 attending or listening to our meetings. Written 7 comments are also welcome. The ACRS section of the 8 U.S. NRC's public website provides our charter, 9 bylaws, letter reports and full transcripts of all 10 full and subcommittee meetings, including slides 11 presented here.
12 The rules for participation at today's 13 meeting were announced in the Federal Register on May 14 24th, 2019. The meeting was announced as an 15 open/close meeting. We may close the meeting at the 16 end of appropriate parts of the sessions to discuss 17 proprietary matters, and presenters can defer 18 questions that should not be answered in public 19 session to that time.
20 I'll go off script and note that we'll 21 make sure the NuScale and staff warn us if we're 22 straying into proprietary matters, and we can make 23 note and hold it til the closed session.
24 We received a written statement from Mr.
25 Michael Derivan. Mr. Derivan's comments will be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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6 1 attached to this meeting's transcript and they may be 2 found in the NRC's document control system under ADAMS 3 accession number ML19168A174. A transcript of the 4 meeting is being kept and will be made available as 5 stated in the Federal Register notice.
6 Therefore, we request that participants in 7 this meeting use microphones located throughout the 8 meeting room when addressing the Subcommittee.
9 Participants should first identify themselves and 10 speak with sufficient clarity and volume so they can 11 be readily heard. I'll note everybody with your 12 appliances, please turn off your phones or at least 13 put them on mute or vibrate so we don't get disturbed.
14 We have a bridge line established for the 15 public to listen to the meetings. To minimize 16 disturbances, the public line will be kept in a 17 listen-in only mode. We'll now proceed with the 18 meeting and I'll call upon Marty Bryan, Marty, of 19 NuScale to begin today's presentations. Is the green 20 light on somewhere, behind all your computers?
21 Overview of Chapter 3 22 MR. BRYAN: Yes, good morning. I'm Marty 23 Bryan. I'm the NuScale Licensing Project Manager for 24 Chapter 3, and we've reorganized it a little bit out 25 of numerical order, just to help with the flow of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 1 presentation. So this morning as you see the sections 2 here, Patrick Conley, J.J. Arthur, Josh Parker and 3 Storm Kauffman will be leading us through. So we're 4 going to start out with Patrick Conley in Section 3.2.
5 MR. CONLEY: Okay, good morning. My name 6 is Patrick Conley. I'm the programs engineer for 7 NuScale Power. I'm going to cover Section 3.2, 3.10 8 and 3.11, respectively.
9 Next slide. So 3.2 is classification of 10 SSC, System Structures and Components. Our SSCs were 11 classified according to the size and category 12 requirements, in accordance with Reg Guide 1.29 Rev.
13 5, the quality groups for Reg Guide 1.26 Rev. 4, and 14 rad waste classifications were in accordance with Reg 15 Guide 1.143 Rev 2.
16 The SSCs were classified as A1, safety-17 related and risk-significant, A2, safety-related but 18 not risk-significant, B1, non-safety but risk-19 significant, and B2, non-safety and non-risk 20 significant. There were COL items that required the 21 applicant to identify and classify any site-specific 22 SSCs according to the requirements set forth in the 23 DCA.
24 MEMBER SKILLMAN: Patrick, before you 25 change. For the record, what is the determining NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 1 characteristic for risk-significant versus not risk-2 significant?
3 MR. CONLEY: We used our PRA insights and 4 that, along with our operating experience, and we had 5 a D-RAP expert panel. We were able to determine the 6 total determination of risk significance for each 7 proponent. So there was a formula of many things that 8 went in, but PRA was part of that factoring.
9 MEMBER SKILLMAN: Thank you.
10 MR. CONLEY: Any other questions on 3.2?
11 (No response.)
12 MR. CONLEY: Next slide, please. 3.11 or 13 excuse me, 3.10, seismic and dynamic qualifications of 14 mechanical and electrical equipment. This chapter 15 addresses the dynamic and seismic qualifications of 16 Seismic Category 1, mechanical and electrical 17 equipment and supports. The methods and procedures 18 that we used were in accordance with Reg Guide 1.100, 19 Rev. 3, which invokes IEEE 344, the 2004 edition.
20 There were three COL items for the 21 applicant to establish a program, develop the records 22 and also to submit a program prior to installation of 23 any of the SSCs in plant. 3.11, environmental 24 qualification of mechanical and electrical equipment.
25 Our EQ program complies with the DSRS 311, which NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 1 deviates a little bit from NUREG 800, as you're well 2 aware.
3 The EQ program includes equipment in 10 4 C.F.R. Part 49 scope, certain PAM equipment specified 5 in Reg Guide 1.97, and active mechanical equipment 6 with safety-related functions. Our program meets the 7 requirements of the generic design criteria 124 and 8 23.
9 You could look in the GCA, you'll see that 10 Table 3.11-1 lists the harsh (phonetic) equipment that 11 is subject to environmental qualification, and 12 environmental qualification conditions considered 13 include AOOs, normal, accident and post-accident 14 conditions, and those are respectively represented in 15 Appendix 3 Charlie of the DCA.
16 There are four COL items to provide site-17 specific EQ program qualification and documentation.
18 Thanks.
19 MR. ARTHUR: My name is J.J. Arthur.
20 I'll be presenting 3.9, 3.12 and 3.13. So Section 21 3.9.1 addresses analysis methods for Seismic Category 22 1 components and supports, including ASME boiler 23 pressure vessel code Division 1, Class 1, 2 and 3 24 components, subsection NG for core support structures 25 and subsection NF for supports.
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10 1 The bulk of Section 3.9.1 is devoted to a 2 description of all of our service level A, B, C, D and 3 test events, which define the thermohydraulic 4 conditions for the NuScale power module and represent 5 bounding, representations of NPM operation that are 6 using the components for these evaluations.
7 This section also contains a catalogue of 8 the software used by NuScale in the dynamic and static 9 analysis of mechanical load stresses and definitions, 10 and also in the hydraulic transient load analysis of 11 the Seismic Category 1 components. These are listed 12 here on the slide, but I'm not going to read them to 13 you.
14 Finally, Section 3.9.2 was covered with 15 you just over a month ago, so we're not -- I'm not 16 discussing that today.
17 Safety-related pressure retaining 18 components, core support structures and component 19 supports are designed and constructed in accordance 20 with the rules of the ASME boiler and pressure vessel 21 code, Section 3, Division 1. Almost all of NuScale's 22 ASME Section 3 components are illustrated on this 23 image shown on the slide, where you can see the 24 containment vessel, reactor pressure vessel and the 25 reactor vessel internals.
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11 1 Section 3, piping, is covered in Section 2 3.12, which we'll talk about in a little bit. Can you 3 go back a moment?
4 MALE PARTICIPANT: Sure.
5 MR. ARTHUR: My last bullet there.
6 Section 3.9.3 contains descriptions of all the loads 7 considered in the design and the components, component 8 supports and core support structures, as well as the 9 load combinations and stress limits applicable for all 10 the service conditions.
11 Control rod drive system is composed of 12 the pressure-retaining housing, control rod drive 13 shaft, which attaches to the control rod assembly hub, 14 and external water cooled electromagnetic coils, which 15 provide for movement of the control rod assembly in 16 and out of the core.
17 Portions of the control rod drive system, 18 which form reactor coolant pressure boundary are 19 designed and constructed in accordance with the rules 20 of the boiler and pressure vessel code, Division 1, 21 Subsection NB for Class 1 components. While the 22 NuScale controller drive mechanisms are very similar 23 to those used in existing PWRs, the NuScale CRDMs have 24 a couple of unique features, namely a remote 25 disconnect mechanism and longer than typical drive NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 1 shafts.
2 NuScale has conducted rod drop testing and 3 confirmed that the system can perform at safety 4 function under conditions that result in misalignment 5 of the control rod drive shaft support structures.
6 Furthermore, a COL applicant is required to implement 7 a control rod drive system operability assurance 8 program that includes a series of tests including 9 performance testing, stability testing, endurance 10 testing and production testing.
11 These test are intended to demonstrate 12 acceptable performance of the CRDS with respect to 13 wear, functioning times, latching and the ability to 14 overcome a stuck rod.
15 CO-CHAIR MARCH-LEUBA: I know we're not 16 talking seismic explicitly, but this is a component of 17 concern, a very long drive shaft. Has misalignment 18 caused by long shaft been considered?
19 MR. ARTHUR: Yes. We used seismic 20 results to inform the misalignment that we tested.
21 It's supported at six locations along the upper riser, 22 and so we misaligned those in the test facility 23 informed by the seismic analysis.
24 CO-CHAIR MARCH-LEUBA: Okay, thank you.
25 MEMBER REMPE: Excuse me for a minute.
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13 1 Someone must have an open line, perhaps from NuScale 2 at this time, and we are getting feedback. So please, 3 everyone on the line mute your phones, okay? Thank 4 you.
5 DR. SCHULTZ: J.J., the prototype testing.
6 Can you amplify that a bit as to who's doing what in 7 the testing program, and what stage is it in at this 8 point?
9 MR. ARTHUR: So we don't have any further 10 testing happening right now. The COL item captures 11 that the COL applicant will establish all of that with 12 the vendor in the future. The only testing we did for 13 the design certification is the raw drop testing.
14 DR. SCHULTZ: From the tests that you've 15 done, how would you describe the scope of what you're 16 asking the COL applicant to do and what you're asking 17 us to wait for, if you will, what you're putting off 18 to the COL stage?
19 MR. ARTHUR: Yeah. I guess --
20 DR. SCHULTZ: It sounded like a pretty 21 expansive testing program that you've described so 22 far.
23 MR. ARTHUR: Yes, it is, and I can't 24 speak to a lot of the details with more than I've 25 said. If someone in Corvallis is there? That's the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 1 CRDM engineer. I don't know who's on the line.
2 MR. BRYAN: I do know in the Chapter 4 3 we've provided quite a bit of information. I was 4 looking back at those slides yesterday. So there's 5 quite a bit of detail there. But I think in summary, 6 we did cold testing, so there's a lot of hot testing 7 still to be done at the COL stage.
8 DR. SCHULTZ: I see, okay. I'll go back 9 and look at that slide set too. Thank you.
10 MR. ARTHUR: Moving on to reactor vessels 11 internals, the NuScale RVI assembly is composed of 12 several subassemblies, which are located inside the 13 reactor pressure vessel.
14 Primary functions of the internals are to 15 provide structure to support, properly orient, 16 position and seat the fuel assemblies, provide support 17 and properly align the control rod drive system, and 18 provide a flow envelope to direct the natural 19 circulation primary coolant flow from the reactor core 20 to the steam generators and back to the core.
21 Core support structures and internal 22 structures are designed and constructed in accordance 23 with the rules of ASME Code, Division 1, Subsection 24 NG.
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15 1 to develop in-service testing requirements specified 2 in 3.96. In addition, NuScale has applied an 3 alternative authorization to also invoke the 4 requirements of the 2017 edition, mandatory Appendix 5 IV for the performance assessment testing of power 6 operating valves.
7 Pumps, valves and dynamic restraints that 8 are required to perform a specific function in 9 shutting down the reactor to a safe shutdown 10 condition, maintain the safe shutdown condition or 11 mitigation consequences of an accident are required to 12 be in the in-service testing program in accordance 13 with ISTE 110-0 of the ASME OM code.
14 The NuScale design has no safety-related 15 pumps, motor operated valves or dynamic restraints.
16 Our IST program contain 39 valves for NPM, which 17 includes 26 that are hydraulically operated. These 18 are the containment isolation valves and the DHRS 19 actuation valves, the five ECCS valves, two air-20 operated valves, four pressure relief valves and two 21 check valves.
22 In addition, we have an augmented valve 23 testing program that contains an additional 12 valves 24 for NPM that either provide non-safety backup to the 25 safety-related function or are non-safety related and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 1 provide augmented quality function.
2 Section 3.9 has 12 COL items that include 3 provision of final design reports for the ASME Code 4 Section 3 components, the establishment of site-5 specific pre-service and in-service inspection 6 programs, and items related to the comprehensive 7 vibration assessment program. Section 312 covers all 8 of the ASME Code Class Form 2 and 3 piping systems, 9 piping components and associated supports.
10 NuScale design has a relatively small 11 amount of -- relatively small amount of relatively 12 small diameter piping, Code Class 1, 2 and 3 piping.
13 The largest piping connected to the reactor coolant 14 system is two inch nominal pipe size, and the largest 15 Section 3 piping in the design is our 12 inch diameter 16 main steam line.
17 MEMBER CORRADINI: The largest two inch is 18 CVCS or the RVV, the safety relief valve? I can't 19 remember which one.
20 MR. ARTHUR: The piping is CVCS.
21 MEMBER CORRADINI: CVCS.
22 MR. ARTHUR: Yes.
23 MEMBER CORRADINI: And the largest the 24 ECCS valves is the RRVV or the safety relief? That's 25 what I can't remember.
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17 1 MR. ARTHUR: I believe it's the safety 2 relief. I'll look to Zack.
3 MEMBER CORRADINI: That's what I remember.
4 That's --
5 MR. ARTHUR: Yeah, I think that's 6 correct.
7 MEMBER CORRADINI: Okay, thank you.
8 MR. ARTHUR: Stress analysis in 9 accordance with the ASME Code and NRC requirements has 10 been performed for the high energy piping within the 11 NPM, in support of Section 3.6 which Storm Kauffman 12 will address later in our presentation. Finally, we 13 have screened all of our piping for thermal 14 stratification and thermal oscillations using the EPRI 15 criteria.
16 This evaluation resulted in the 17 identification of one potentially susceptible 18 location. CFD analysis was performed to demonstrate 19 that stratification does not occur, and that the 20 temperature fluctuations in the decay heat removal 21 system align, and the associated containment 22 penetration cause thermal stresses that are below the 23 endurance limit for the materials of the piping, the 24 rods and the containment vessel. This slide shows --
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18 1 for monitoring for thermal striping or any of that 2 sort of thing?
3 MR. ARTHUR: Aside from an in-service 4 inspection that we plan to do.
5 MEMBER BALLINGER: What are you defining 6 as the endurance limit?
7 MR. ARTHUR: The endurance limit is the 8 stress below which there are no concerns with fatigue, 9 for an infinite number of cycles without compromising 10 the material.
11 CO-CHAIR RICCARDELLA: We know that.
12 What's the value?
13 MEMBER BALLINGER: Yeah. For these 14 materials as soon as you put water in there, there is 15 no endurance limit. So it's defined as some number of 16 cycles, usually 10 to the 7th, maybe 10 to the 8th or 17 thereabouts.
18 MR. ARTHUR: Yeah. I don't recall what 19 we used in this case.
20 CO-CHAIR RICCARDELLA: For stainless, I 21 think the ASME Code goes out to 10 to the 9th.
22 MALE PARTICIPANT: 10 to the 9th.
23 CO-CHAIR RICCARDELLA: Yeah. They just, 24 a new rev of the ASME Code. Okay.
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19 1 has been identified as potentially susceptible, what 2 consideration is given to having amend flow orifice or 3 a triple flow, so that there's just a constant, almost 4 inconsequential heat loss that nevertheless 5 stratification stopped.
6 MR. ARTHUR: I'm not sure we could do 7 that. This is the DHR condensate line, where it 8 attaches to the steam generator from the feed water 9 line. I guess I'll say the analysis we've done has 10 shown us that while it was potentially susceptible, 11 we're not concerned about it any longer.
12 MEMBER SKILLMAN: Thank you.
13 CO-CHAIR RICCARDELLA: Under what mode of 14 operation is the concern? Under normal operation or 15 under DHRS operation or --
16 MR. ARTHUR: Yes. The potentially 17 susceptible situation is during normal operation.
18 MEMBER CORRADINI: So we're talking at the 19 bottom of the condensate line connecting to the steam 20 generator?
21 MR. ARTHUR: Correct.
22 MEMBER CORRADINI: Okay, thank you.
23 MR. ARTHUR: This image shows all of our 24 -- almost all of our Section 3 piping. The middle 25 image shows the piping inside the containment vessel.
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20 1 Most of what you see there is main steam feedwater and 2 CVCS, and then the image on the right shows the piping 3 on the containment vessel head, and then also attached 4 to the side of the containment.
5 CO-CHAIR MARCH-LEUBA: One concern I've 6 always had with this small pipes connecting the vessel 7 to the containment is the seismic problem when you 8 have two humongous masses moving at different 9 frequencies, and clearly having some displacement. I 10 see that you are going with the straight shots and 11 everything is curved. Has it been considered that it 12 has enough elasticity not to break in a seismic event?
13 MR. ARTHUR: Yes. All of our analysis 14 includes consideration of the seismic anchor motions 15 --
16 CO-CHAIR MARCH-LEUBA: I would think that 17 that is the limiting by far.
18 MR. ARTHUR: Yes, yes, that's correct.
19 CO-CHAIR MARCH-LEUBA: You have two 20 humongous masses going like this.
21 MR. ARTHUR: Right.
22 CO-CHAIR RICCARDELLA: So the ASME Code 23 for -- permits socket welded piping for I think two 24 inch diameter and less. At a prior meeting I asked 25 the question about your small bore piping and I got NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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21 1 the response that you're not using any socket welds?
2 MR. ARTHUR: That's correct. We don't 3 have socket welds.
4 CO-CHAIR RICCARDELLA: You're using full 5 penetration welds?
6 MR. ARTHUR: That's right.
7 CO-CHAIR RICCARDELLA: Thank you. Just 8 wanted to confirm.
9 MR. ARTHUR: The design analysis of 10 pressure boundary threaded fasteners complies with the 11 ASME code for Class 1, 2 and 3. Not applying any code 12 cases, but following the code as written. We have one 13 COL item in this section for a site-specific in-14 service inspection program for threaded fasteners.
15 (Off microphone comment.)
16 CO-CHAIR RICCARDELLA: Thanks, J.J. Now 17 Josh Parker will take us through the 37386.
18 MR. PARKER: Good morning. My name is 19 Josh Parker and I'll be going through, as Marty said, 20 3738 and 33 through 35. Before I get started through 21 the bulk of my presentation, I was going to present 22 this cross section through some of our major 23 structures. So what you see here is the -- a cross 24 section through the reactor building, which is the 25 building in the middle with the red roof on it.
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22 1 To the right of that or to the east is the 2 control building. Both of these structures are 3 Seismic Category 1 buildings on the site, and then to 4 the left or to the west of the reactor building is our 5 rad waste building, and that is a Seismic Category 2 6 structure and also a Rad Waste Category RW2A, per the 7 requirements in Reg Guide 1.13.
8 So we'll go into Section 33 now on the 9 wind and tornado loadings. This was discussed as a 10 part of Chapter 2 of the FSAR back in December. But 11 our design basis wind load is 145 mile an hour, three 12 second gusts and used Exposure Category C. The 13 reactor building, control building and rad waste 14 building are all assessed using an importance factor 15 of 1.15.
16 For tornado loads, we have conservatively 17 assumed the tornado characteristics of Region I that 18 are defined in Reg Guide 1.76, and this is the highest 19 wind speed and this is all the characteristics that 20 are defined in the Reg Guide. Similarly, the design 21 basis hurricane wind speed NSO (phonetic) criteria 22 have been taken directly from Reg Guide 1.221. All 23 this results in wind analysis complying with the 24 requirements of GDC 2 and 4.
25 Then in this section we have a COL item NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 1 that confirms that site-specific adjacent structures 2 will not collapse or have an adverse effect on the 3 Seismic Category 1 structures. As for 3.4 --
4 MEMBER CORRADINI: So let me ask the 5 question before we move on. So the limiting missiles 6 are not from wind and tornado?
7 MR. PARKER: The limiting missiles?
8 MEMBER CORRADINI: The loadings and 9 missiles that would be generated by them are not from 10 wind and tornado, the way the design is set up. Am I 11 understanding correctly?
12 MR. PARKER: We have assessed for the --
13 we have assessed wind and tornado missiles, and we've 14 assessed for turbine missiles, and that will be 15 discussed in 3.5.
16 MEMBER CORRADINI: Okay, fine. Thank you.
17 MR. PARKER: For flood design, first 18 internal flooding. Our internal flooding is done by 19 a level by level and room by room area analysis for 20 both the reactor building and the control building for 21 postulated flooding events. So we do that by first 22 assessing the water systems that might be in an area, 23 and the volume of water and the flow rate and then the 24 time to isolate that system if there was a postulated 25 break.
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24 1 Then we identify the -- any equipment that 2 need protecting in those rooms, and then can determine 3 what the mitigation, mitigating features are to 4 protect that particular system. So that could be 5 watertight doors, it could be elevating the system, 6 what have you, and that's all been described in 3.4 of 7 the FSAR. Next slide. So the previous slide was on 8 internal flooding --
9 MEMBER CORRADINI: So let me ask a 10 different question related to that. You guys in the 11 design already know what's on what floor?
12 MR. PARKER: Yes.
13 MEMBER CORRADINI: So what's at the bottom 14 floor?
15 MR. PARKER: Rad waste systems.
16 MEMBER CORRADINI: Okay, and then?
18 MEMBER CORRADINI: And then?
19 MR. PARKER: Electrical and I&C.
20 MEMBER CORRADINI: So where's the 21 batteries?
22 MR. PARKER: That's on the 75 foot 23 elevation of the reactor building, one floor below 24 grade.
25 MEMBER CORRADINI: Okay, thank you.
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25 1 MR. PARKER: And our favorite system, the 2 boron system, EAS?
3 MEMBER CORRADINI: That's the 50 foot 4 elevation. It's two floors below grade.
5 MR. PARKER: Two floors below the water 6 level, is that -- or below grade?
7 MEMBER CORRADINI: Below grade.
8 MR. PARKER: Water level is higher?
9 MEMBER CORRADINI: Well, if we're talking 10 water level in the pool, that water level is higher.
11 But the water level we look at is for any --
12 MR. PARKER: You're talking about 13 flooding?
14 MEMBER CORRADINI: Yes, right.
15 MR. PARKER: Yeah. So any piping systems 16 that are running through the building, we postulate 17 grades for all of them, and then assess the equivalent 18 that's around it.
19 CO-CHAIR MARCH-LEUBA: So I'm just a 20 little -- I only had one cup of coffee this morning.
21 The BAS is then a 55, which is like ten feet below 22 grade?
23 MR. PARKER: So we establish grade at 100 24 feet at 100 feet as nominally -- that's what we call 25 our grade level --
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26 1 (Simultaneous speaking.)
2 CO-CHAIR MARCH-LEUBA: 45 feet?
3 MR. PARKER: So the 50 foot elevation is 4 actually two floors below grade. So nominally each 5 floor is about 25 feet in elevation.
6 CO-CHAIR MARCH-LEUBA: Two floors, two big 7 floors?
8 MR. PARKER: Two big floors, right. Okay.
9 So that was internal flooding. Now with respect to 10 external flooding, we defined our maximum flood 11 elevation for potential flooding as one foot below 12 grade, and the maximum water elevation of two feet 13 below grade. So these all become -- as a result, 14 there is no dynamic flood loads on these structures.
15 But these saturated soil pressures and 16 water loads are used in the static and dynamic 17 analysis of the structures that's a part of 3.7 and 18 3.8. Similarly for precipitation, the rates are given 19 there, and those become input to the structural 20 analysis loads that we used for the design of the 21 structures.
22 Lastly on 3.4, we're satisfying GDC 4 and 23 2. We also look at the interaction between non-24 Seismic Category 1 with Seismic Category 1 structures 25 to assess for any potential credible flooding source.
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27 1 As we talked about previously, all of the equipment 2 that needs protection is in the reactor building or 3 control building. So those buildings have been 4 assessed and we provide those mitigation features 5 there.
6 Then we have seven COL items to confirm 7 site-specific conditions, and the -- and materials for 8 the structure used.
9 MEMBER BALLINGER: I had a curiosity 10 question. Could you back up one slide? On the 11 extreme snow load, 75 PS, pounds per square foot?
12 MR. PARKER: Yes.
13 MEMBER BALLINGER: Water is 62.4 pounds 14 per cubic foot. So that would be 65 pounds per square 15 foot. Where did the 75 pounds come from?
16 MR. PARKER: So these are --
17 MEMBER BALLINGER: I mean how do you get 18 75 pounds?
19 MR. PARKER: Yeah. Well, these come from 20 ASME 7.
21 MEMBER BALLINGER: It just struck me as 22 weird.
23 MEMBER SKILLMAN: Well, for 14 feet of 24 snow.
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28 1 14 feet of snow.
2 CO-CHAIR RICCARDELLA: No, it's ice. It's 3 ice.
4 MEMBER BALLINGER: So divide by ten 5 roughly, right, for the snow?
6 MEMBER SKILLMAN: It's ice.
7 CO-CHAIR RICCARDELLA: Depends on where 8 you are.
9 MALE PARTICIPANT: Well, ice would be pure 10 water, yeah.
11 CO-CHAIR RICCARDELLA: If you're in New 12 England, you divide by 2.
13 CO-CHAIR MARCH-LEUBA: The roof is very 14 flat. I mean it just accumulates the snow and that's 15 --
16 MR. PARKER: The reactor building roof is 17 relatively flat and yes, I mean there may be some 18 slight sloping to it. There likely would be for 19 drainage, so there wouldn't be any ponding.
20 CO-CHAIR MARCH-LEUBA: So you're in 21 Northern Minnesota. You will accumulate a lot of snow 22 there, right?
23 MR. PARKER: Potentially.
24 MALE PARTICIPANT: With global warming, 25 no.
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29 1 CO-CHAIR MARCH-LEUBA: I know it's higher.
2 (Simultaneous speaking.)
3 MR. PARKER: All right. Moving on to 4 missile analysis now. So here again we're meeting GDC 5 2 and 4. As I mentioned before, our rad waste 6 building is classified as RW2A, and Reg Guide 1.143 7 indicates that that would have us also use the same 8 missiles as a Seismic Category 1 structure. So our 9 rad waste building is designed for the same missiles 10 as the reactor building and the control building.
11 The missiles considered are the five kind 12 of categories that are listed here on the slide, 13 internally generated turbine missile, tornado and 14 hurricane missiles. They also list site for aircraft 15 hazards, but we don't postulate any missiles from 16 them, giving the siting criteria we've established in 17 Chapter 2.
18 Then there's -- we have a beyond design 19 basis aircraft impact assessment that was covered in 20 19.5 the last time.
21 MEMBER SKILLMAN: When you say you've 22 excluded what I think you said is the wind-driven 23 missiles --
24 MR. PARKER: We have not excluded those.
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30 1 a Buick coming through the roof or a Silverado coming 2 through the truck bay at 35 and 50 miles an hour.
3 MR. PARKER: So the wind-driven missiles 4 are an automobile, a sphere and essentially a wooden 5 shaft like a hole. All those have been considered.
6 MEMBER SKILLMAN: Thank you.
7 MR. PARKER: Next slide. So for the 8 NuScale design, we had no credible missiles inside 9 containment, and as we just talked about, the wind and 10 hurricane missiles have been defined by the criteria 11 in Reg Guide 1.76 and 1.221 respectively.
12 All of our safety-related and risk 13 significant components are located inside the reactor 14 building or the Seismic Category 1 portion of the 15 control building, and a function of those structures 16 is to act as a barrier to protect safety-related 17 systems. We have -- our analyses have shown they can 18 serve that function against all postulated missiles.
19 We have four COL items to confirm the 20 site-specific missile analysis in the missile 21 criteria. So next on turbine missiles, a little more 22 detail here.
23 MEMBER BLEY: Can I stop you before you 24 get started on this one?
25 MR. PARKER: Sure.
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31 1 MEMBER BLEY: I have a lot of concerns on 2 this one, and let me just put a few things out there 3 before you start. Some of this may apply more to the 4 staff than to you, but I think that it applies to 5 both. Early on we had a lot of plants designed with 6 unfavorable positioning of the turbine, and then once 7 we started talking about turbines, we mostly tried not 8 to point them at anything we care about.
9 You've decided to claim that's something 10 you care about and to use the -- not a probabilistic 11 argument, which might be pretty good given the size of 12 your rotors. I think you have monoblocks, right.
13 But you decided to use the barrier 14 approach. I'm going to ask the staff. I don't know 15 if anybody's used that before in the turbine missile 16 licensing process. If we have -- that gives us a 17 little more basis. If we haven't, going through your 18 report, you throw out equations for penetration depth 19 and other things. I didn't see a source for those in 20 your report.
21 If we go to NRC's reg guide, they point to 22 a 19, I think it's 87 or 97 reference. When you go 23 through that, it points out that most of the data they 24 used to come up with -- and they came up with five 25 different formulations or they found five different NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 1 formulations, I think it was five, from the munitions 2 area looking at penetration of -- intentional 3 penetration of missiles.
4 So I suspect everything from that kind of 5 analysis, that when they think of being conservative, 6 they want to be sure no missiles get through. For our 7 case, we want to make sure the missiles don't get 8 through.
9 So whatever was conservative over there is 10 probably non-conservative over here. One, where did 11 your formulations come from, what's the basis for 12 them? Two, do you think they're conservative?
13 I was involved in a probabilistic analysis 14 of this problem some years ago. It's very complex if 15 you look at all of the different angels you can come 16 from and all the different masses and momentums, the 17 key thing. The reg guide cites a penetration velocity 18 that echoes just a wealth of physics built into it.
19 It's not just a number.
20 So if you can tell me why you have really 21 good confidence, I'd appreciate it. It strikes me 22 that if this is unique, as I expect it is and even if 23 it isn't, I would have expected in a technical report 24 or something to really lay out how you did this 25 analysis, because proving that you can't get through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 1 the barrier is crucial, and from what I've read, it's 2 hard to be convinced that you really thought of 3 everything and looked at it thoroughly. Go ahead.
4 MR. PARKER: All right. So with respect 5 to our turbine missile approach or our barrier 6 approach, you're right. I don't believe there's been 7 another vendor that's done it this way. We reviewed 8 the paper you referenced. There are a couple of 9 things about that paper.
10 One is that one of the out, the going-in 11 assumptions is that both the missile and the target 12 are infinitely rigid, and the paper makes the point 13 that if there is any deformation, that those equations 14 were drastically over-predict the amount of 15 penetration. They're also based on a --
16 MEMBER BLEY: But we don't know what 17 drastically means, but go ahead.
18 MR. PARKER: They're also based on a 19 certain penetration depth to barrier thickness values.
20 So we opted to use a finite element analysis to 21 predict the amount of penetration distance that our 22 missile would go into our barriers, and we validated 23 --
24 MEMBER BLEY: That is from like the basic 25 physics approach?
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34 1 MR. PARKER: Yeah, we validated --
2 MEMBER BLEY: Other than these 3 correlations that's in another place?
4 MR. PARKER: Correct, that's right. We 5 validated that, what the software was producing with 6 this test data.
7 MEMBER CORRADINI: What test data?
8 MR. PARKER: I don't have that, those 9 references in front of me.
10 MEMBER CORRADINI: So to follow Dennis' 11 question, I think, is there a topical or technical 12 report that we can look at? I think that's where --
13 MR. PARKER: No.
14 MALE PARTICIPANT: The answer is no.
15 MR. PARKER: There's not. No, this is all 16 in our FSAR and our supporting calculations.
17 MEMBER CORRADINI: So the supporting 18 calculational document?
19 MR. PARKER: Yes.
20 MR. BRYAN: Yeah. I would note for the 21 reason you stated, the staff had quite a few of the 22 same questions. We had a number of interactions, and 23 then in June we submitted an extensive document with 24 that background or basis for how we got there.
25 MEMBER BLEY: Is that on the docket?
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35 1 MR. BRYAN: Yes sir. So it's pretty 2 comprehensive.
3 MEMBER BLEY: We'd like to see that Mike, 4 if you can find that for us. Not today, you know, 5 some time later. We'll look at it before they come 6 back.
7 MS Is that the enclosure to that RAI 8 response or is it --
9 MR. BRYAN: Yes, it's part of the RAI 10 responses.
11 MEMBER BLEY: Oh okay. So maybe we found 12 it yesterday.
13 (Simultaneous speaking.)
14 MEMBER CORRADINI: Is that what you 15 requested?
16 MEMBER BLEY: Well, yes I did.
17 MEMBER CORRADINI: Okay.
18 MEMBER BLEY: The backup document, I 19 haven't looked at it yet. But is that the document 20 that really provides the full engineering analysis?
21 MR. BRYAN: Yes, it is. We extracted from 22 all the supporting documents, and then the staff has 23 audited those. But it's quite a bit of detail in that 24 RAI response in June.
25 MEMBER BLEY: I hope so, because this is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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36 1 -- I mean this is kind of new ground for the NRC, and 2 it's a place we've got to make sure we've done it 3 right.
4 MR. BRYAN: The staff had the same 5 questions, yeah.
6 MEMBER BLEY: Including your analysis must 7 have looked at -- okay. I'm assuming you didn't have 8 any missiles go all the way through?
9 MR. PARKER: Not for the reactor building, 10 and for the control building, it didn't go through the 11 exterior wall. For the control building our barriers, 12 the exterior wall and the grade floor elevation. So 13 we didn't see missiles going through our credited 14 barriers.
15 MEMBER BLEY: Okay, that's good. I really 16 want to look at that. From the study I was involved 17 in some years ago, the really interesting thing that 18 came out of that was, and this was going through heavy 19 reinforced concrete, if you had enough energy to get 20 through the wall, even though your velocity is lower 21 inside, some of the cases we looked at you didn't have 22 enough energy to get out the opposite wall, and then 23 it looks like a pinball machine in there.
24 It just spins all around, crashing into 25 wall and cutting everything to pieces. I look forward NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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37 1 to looking at that detailed analysis. If you saw --
2 did you do a probabilistic analysis?
3 MR. PARKER: No. We defined these three 4 missiles here, the blade, a blade with a portion of 5 the rotor --
6 MEMBER BLEY: We don't care about a blade.
7 MR. PARKER: The blade, the portion of the 8 rotor, and then half of the last stage of the rotor.
9 MEMBER BLEY: Okay, and that -- half of 10 the last stage is new. You hadn't looked at that 11 before?
12 MR. PARKER: That's right.
13 MEMBER BLEY: Okay. Well half is pretty 14 big, and you know most of the stuff I've seen, you 15 usually get two to four or something like that big 16 pieces out. So half is -- I mean three to four. So 17 half is pretty big, and you assumed the velocity --
18 MR. PARKER: So the velocity was based on 19 a 3600 RPM turbine and then we also looked at varying 20 overspeed conditions.
21 MEMBER BLEY: Okay, and did you assume 22 overspeed trucks would keep it from going too fast?
23 MR. PARKER: We varied the overspeed. I 24 mean we defined our destructive overspeed at 160 25 percent. But we looked at it all the way up to 210 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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38 1 percent.
2 MEMBER BLEY: Okay, okay. Well, I look 3 forward to seeing that detail. Go ahead.
4 MEMBER SKILLMAN: I'd like to ask a basic 5 question. Given all the effort in industry over the 6 last probably 30 years on this topic, as I imagine our 7 footprint east-west, with modules 1 through 12 8 adjacent to each other kind of in the center and going 9 west, the control building unit to the east, rad waste 10 on the left, here you have these two turbine buildings 11 that are basically on the same footprint as the 12 reactor building.
13 Why wouldn't you just have oriented the 14 turbines to have shafts north and south, so that you 15 would not be dealing with this?
16 MR. PARKER: It really just kind of became 17 to an optimization of piping layout at the time when 18 the site was arranged.
19 MEMBER BLEY: So minimizing piping or 20 something along those lines?
21 MR. PARKER: Right.
22 MEMBER BLEY: Because we asked about this 23 in an earlier meeting. You put it off until now, and 24 the arguments for orienting them this way weren't 25 really clear from that meeting. I guess I'm not 100 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 1 percent clear here, but a little extra piping to avoid 2 this problem would have been a nice idea.
3 MEMBER BALLINGER: You use a monoblock 4 rotor, right? So what is the definition of half of 5 last stage turbine rotor disc? It's a monoblock 6 rotor. There's no disc.
7 MR. PARKER: I'll defer that to --
8 MEMBER BALLINGER: Have I got the design 9 wrong?
10 MR. HOUGHTON: This is Zack Houghton, the 11 mechanical design engineering manager, and I was also 12 here presenting Chapter 10, where we discussed this 13 topic a bit as well.
14 MEMBER BLEY: Yeah.
15 MR. HOUGHTON: For the monoblock rotor, it 16 looks at a section that would be where the roof 17 section of the blade is. So that's where I think it's 18 referring to disc, when it looks at a chunk of the 19 rotor from that section.
20 MEMBER BLEY: Okay.
21 MR. HOUGHTON: As far as some of the other 22 questions that we heard, the reason for the 23 orientation the way it is, one of the things that we 24 looked at is not just length of piping but also 25 balancing, right, so that we get equivalent operation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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40 1 between all the turbine generators.
2 If we stack them all width-wise, they 3 would be further out than the reactors are themselves, 4 right? The width of the turbine island would be 5 larger than the reactors. So your modules for 6 Reactors 1 and 6 would be much further away than for 7 Modules 3 and 4. So we looked at all the impacts of 8 going with the barrier approach per Reg Guide 1.115.
9 I'll note that even in the favorable 10 orientation, we would still have to account for high 11 trajectory blades. That's a requirement of the 12 regulation, and it also --
13 MEMBER BLEY: I know it is, but you know 14 that goes away on probability basis.
15 MR. HOUGHTON: Right. So the probability 16 basis --
17 (Simultaneous speaking.)
18 MEMBER BLEY: Just the solid angles, then 19 you drop one back down to get it.
20 MR. HOUGHTON: And even for a favorable 21 orientation, we still have to do the probability 22 analysis to look at, you know, to ensure that the 23 missile generation frequency would be acceptably low.
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41 1 like, and they confirmed that they could meet the 2 requirements of the probability analysis, whether it 3 was in the favorably or unfavorably oriented position.
4 MEMBER BLEY: I would -- that's what I 5 would expect with a monoblock rotor. But I haven't 6 seen any data on those. I don't know what our 7 experience has been with these.
8 (Simultaneous speaking.)
9 MEMBER CORRADINI: I was going to say, 10 with the monoblock rotor I would expect --
11 MALE PARTICIPANT: Zero.
12 MEMBER CORRADINI: Zero.
13 MEMBER BLEY: Yeah, or something close.
14 Anyway --
15 MR. HOUGHTON: And we would still have to 16 look at blade failure. So even in the monoblock 17 rotor, you're still looking at the root section, 18 right, and it's -- even in a monoblock rotor --
19 MEMBER BLEY: What do the blades go 20 through? Come on.
21 MR. HOUGHTON: What's that?
22 MEMBER BLEY: The blades don't go through 23 much.
24 MR. HOUGHTON: Sure.
25 MEMBER BLEY: They might get out of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 1 machine, but they don't have much energy.
2 MR. HOUGHTON: The rotor fragment is the 3 more limiting missile. Yes, we agree.
4 MR. PARKER: Okay. I think we've covered 5 everything in this slide through their discussions, so 6 let me go to the next slide. So -- kind of continuing 7 on. Our barriers are designed through the SRP. We do 8 look at both the local effects of the missile as well 9 as the global effects from the missile, which is 10 really where we see more of the last stage of the 11 rotor, since it's much larger than the others.
12 So we determined the required thickness 13 for the penetration, perforation and scabbing. Given 14 the wall thickness of the reactor building and the 15 wall thickness and slab thickness of the control 16 building, we have shown adequate protection. I mean 17 there is a number of conservatisms in the analysis 18 that we haven't considered.
19 For example, we don't take credit for any 20 of the reinforcement of the walls. We don't take 21 credit for the thickness of the turbine casing. None 22 of that is included in the analysis.
23 MEMBER BLEY: You just left me confused.
24 The barrier wall you did, of course, that's what 25 you're hitting.
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43 1 MR. PARKER: But yeah. We assumed --
2 MEMBER BLEY: You're assuming there was no 3 steel in it?
4 MR. PARKER: We assumed that the missile 5 hit between layers of reinforcement.
6 MEMBER BLEY: What kind of spacing is 7 there in your reinforcement?
8 MR. PARKER: I mean it's six to 12 inches, 9 depending on the wall. So again, that's a 10 conservative --
11 MEMBER BLEY: All you analyzed was the 12 concrete wall?
13 MR. PARKER: That's right.
14 MEMBER BLEY: Hmmm.
15 MR. PARKER: And still showed that the 16 missile didn't penetrate.
17 MEMBER BLEY: We'll look forward to 18 looking at that.
19 (Off microphone comments.)
20 MR. PARKER: Next slide. So at this point 21 we'll transition into seismic design. First of all, 22 our input parameters. So this is our certified 23 seismic design response vector, our CSDRS, this figure 24 on the lower left. This is our input specter for all 25 the Category 1 SSCs in our design, and that's on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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44 1 lower left.
2 In addition, we have a CSDRS that we've 3 also added some high frequency content to. So that is 4 our CSDRS HF or that's the figure that's on the lower 5 right. These are both our input specters for our 6 Category 1 structures.
7 MEMBER CORRADINI: So, you're looking at 8 the maximum of those two?
9 MR. PARKER: We look at both and, then, we 10 take the bounding case.
11 MEMBER CORRADINI: Okay.
12 MR. PARKER: Yes.
13 MEMBER CORRADINI: Thank you.
14 MR. PARKER: Both these spectra are 15 developed based on industry data and they meet the 16 requirements of GDC 2 and 10 CFR 50, Appendix S.
17 Next slide, yes.
18 So, the figure on the screen here shows 19 our spectra in orange and, then, also, the spectra 20 from Reg Guide 1.60, anchored at 0.3g. We can see 21 that the spectra that we've used is both higher in its 22 ZPA -- we have a 0.5 ZPA -- and is more broad than the 23 typical spectras used for design certifications.
24 Next slide.
25 Moving on with our other seismic inputs, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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45 1 we have five seed time histories that are compatible 2 with their CSDRS and an additional seed time history 3 compatible with the CSDRS HF.
4 We have four generic soil profiles that 5 we've used in the analysis, and they vary from soft, 6 which is a soil type 11, to very hard rock, or soil 7 type 9.
8 The damping values for the analysis models 9 follows the guidance of Reg Guide 1.61. And as 10 previously stated, we only have two Seismic Category 11 I structures. They're the reactor building and the 12 control building. The rad waste building is Seismic 13 Category II, given its proximity to the reactor 14 building. And so, it's included in our multi-building 15 analyses models that are developed in SAP2000 and 16 SASSI. And I'll discuss those in the new few slides.
17 So, first, SAP2000, we use that to model 18 the static analyses for the structures. These models 19 have both material properties for both uncracked and 20 fully cracked concretes. In the case of the reactor 21 building, we also have a submodel of the NPM that's 22 based on the work that's performed in ANSYS. And that 23 was discussed as a part of 392 and the separate NPM 24 Seismic Technical Report.
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46 1 models of these buildings that we call single building 2 models. And then, we also create a model that has all 3 three of the reactor building, control building, and 4 rad waste building. And we call that triple building 5 model. We create individual models and triple 6 building models for it in SAP2000.
7 MEMBER CORRADINI: The purpose is 8 interaction between the buildings?
9 MR. PARKER: That's correct.
10 MEMBER CORRADINI: Okay. Thank you.
11 MR. PARKER: Yes. Given their proximity 12 to each other.
13 Next is SASSI. We use SASSI2010 for the 14 dynamic soil structure interaction analysis. The 15 SAP2000 models that we talked about in the previous 16 slide are converted to SASSI. In doing so, the models 17 have exactly the same node numbers, coordinates, 18 element types, and material properties. And the 19 figure on the upper right shows the triple building 20 SASSI model for our design, and we have individual 21 SASSI models as well.
22 We perform our SSI analysis using the 23 extended subtraction method. In this approach, the 24 interaction nodes are taken at the skin of the 25 excavated soil model, which is the six sides, and we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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47 1 also have a plane in the middle. So, that's our 2 extended subtraction method. Or we also call that the 3 7P method. And that's done for development of ISRS 4 and for the dynamic forces and moments in the 5 structures.
6 Next slide.
7 Next, I'll talk about our use of ANSYS in 8 the structural analysis. We have three main purposes 9 for using ANSYS in structural design.
10 The first is the study of Fluid-Structure 11 Interaction. So, given that neither SAP2000 nor SASSI 12 has the ability to model fluid elements, we use ANSYS 13 to determine the hydrodynamic pressure in the reactor 14 pool.
15 MEMBER CORRADINI: So, just educate me.
16 MR. PARKER: Yes.
17 MEMBER CORRADINI: So, that's the reason 18 you go with ANSYS inside, because it has fluid 19 elements?
20 MR. PARKER: That's right.
21 MEMBER CORRADINI: But the reason you 22 don't use ANSYS outside is because the other two have 23 an advantage in some manner?
24 MR. PARKER: Well, so SASSI is what we use 25 for the dynamic analysis. So, that models the soil NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 1 structure interaction.
2 MEMBER CORRADINI: Okay. And that's 3 unique to that tool?
4 MR. PARKER: That's unique to that tool.
5 MEMBER CORRADINI: Okay.
6 MR. PARKER: And SAP2000 was used mainly 7 just out of convenience. It's a common tool for 8 structural analysis --
9 MEMBER CORRADINI: Okay.
10 MR. PARKER: -- building structural 11 analysis.
12 MEMBER CORRADINI: But the reason to 13 switch is strictly the fluid model?
14 MR. PARKER: Well, so we have three main 15 purposes for ANSYS. One is for the fluid analysis, so 16 to get those dynamic pressures in the pool and, also, 17 to determine the sloshing wave height.
18 MEMBER CORRADINI: Right.
19 MR. PARKER: We use it for that FSI 20 approach. We also use it for non-linear stability 21 analysis. And I'll talk about that a little bit when 22 we talk about 385.
23 MEMBER CORRADINI: Okay. Thank you.
24 MR. PARKER: And then, the third 25 approach --
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49 1 MEMBER BLEY: Well, I'm sorry. On the 2 sloshing --
3 MR. PARKER: Yes?
4 MEMBER BLEY: -- how much can you actually 5 slosh out and where does it go? Does it come right 6 back in?
7 MR. PARKER: So, the sloshing height is 8 about 2 feet. It's about 24 inches.
9 MEMBER BLEY: Uh-hum.
10 MR. PARKER: And our freeboard is about 6 11 feet. So, there's no water that sloshes out.
12 MEMBER BLEY: None should come out?
13 MEMBER CORRADINI: Nothing comes out?
14 MR. PARKER: Nothing comes out.
15 MEMBER CORRADINI: I guess just thinking 16 out loud, the fact that you have a fluid element and 17 it can take zero shear, and its load, its unusual load 18 on structures would be probably the most important 19 thing versus the sloshing, assuming you've got a high 20 enough wall.
21 MR. PARKER: Sure. Yes, in that case.
22 MEMBER CORRADINI: Okay. All right.
23 Thank you.
24 MR. PARKER: Then, our last use of ANSYS 25 is determining the thermal analysis and hydrogen line NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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50 1 break analysis. So, we actually use 3D solid concrete 2 elements and explicit rebar and pool elements. And 3 then, we do cases for operating thermal and accident 4 thermal to determine the strains in the rebar and pool 5 liner during the thermal and hydrogen line break.
6 So, on 373, our subsystem analysis, we 7 have four subsystems that we've considered in the DCA.
8 The NPM, which is described in the NPM Seismic 9 Technical Reports; the crane; and the fuel racks, 10 which are discussed as a part of Chapter 9, and then 11 the bioshield.
12 So, the bioshield is the main subsystem we 13 talk about in 373. It's a non-safety-related, not-14 risk-significant Seismic Category II component. Its 15 major functions are for fire protection, radiation 16 protection, ventilation, and to support personnel 17 access, as they're located over on top of the modules.
18 MEMBER CORRADINI: So, just so I 19 understand, the bioshield is the wall facing the pool.
20 There's two concrete partitions or walls that separate 21 the modules, and then, there's a concrete wall which 22 is your missile protection on the outside. So, the 23 bioshield is on the inside facing the other modules?
24 MR. PARKER: Correct. And it has a 25 concrete portion that's horizontal.
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51 1 MEMBER CORRADINI: Right, but it lays on 2 top.
3 MR. PARKER: And it's this steel portion 4 that we kind of see here, which is on the lower right 5 of this screen.
6 MEMBER CORRADINI: Is this the new design 7 that we --
8 MR. PARKER: This is the new design. The 9 previous design had hinged panels.
10 MEMBER CORRADINI: Right, right, right.
11 Okay.
12 MR. PARKER: This is the latest design 13 that is --
14 MEMBER CORRADINI: So, it's louvered?
15 MR. PARKER: Yes, essentially, it's 16 louvered. It has these kind of alternating HDPE 17 panels with a steel frame that allows for continuous 18 passive airflow.
19 MEMBER CORRADINI: Since you've mentioned 20 airflow, are you allowed to say in open session what 21 the height of these things are?
22 MR. PARKER: Yes, I think so. It's about 23 25-ish feet.
24 MEMBER CORRADINI: No, no, each of the 25 louvered panels.
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52 1 MR. PARKER: Oh, each of the louvered 2 panels? Yes, that probably would be getting into the 3 proprietary --
4 MEMBER CORRADINI: Okay, fine. We'll come 5 back to it later.
6 MR. PARKER: Yes.
7 MEMBER CORRADINI: Thank you.
8 MR. PARKER: Okay. Next slide.
9 Lastly, with respect to --
10 MEMBER SKILLMAN: Yes, let's go back just 11 for a second.
12 MR. PARKER: Okay.
13 MEMBER SKILLMAN: You're clear that it's 14 non-seismic, not safety. So, let's take module 2, 15 which is one east from one.
16 MR. PARKER: Correct.
17 MEMBER SKILLMAN: And we're going to 18 refuel two. This is an approximately 75-ton, 150,000-19 pound load, right? As I read your book, that's what 20 I remembered, 75 --
21 MR. PARKER: That's right.
22 MEMBER SKILLMAN: -- 145,000 pounds.
23 MR. PARKER: Okay.
24 MEMBER SKILLMAN: And so, there might be 25 some adjustment for the new shield, for the new NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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53 1 curtain. So, the No. 2 bioshield will be lifted and 2 placed either on one or three, correct?
3 MR. PARKER: Or on the other side.
4 MEMBER SKILLMAN: Oh, okay. What ensures 5 that in that configuration this device placed on 6 probably an identical device does not become either an 7 unacceptable load or under seismic acceleration does 8 not become dislodged or dislodge the one that is 9 holding it?
10 MR. PARKER: So, that is a design basis 11 consideration, that one bioshield will be placed on 12 top of another. So, we've performed an analysis for 13 that very case.
14 MEMBER SKILLMAN: Is the maximum stacking 15 two --
16 MR. PARKER: Yes.
17 MEMBER SKILLMAN: -- to only one?
18 MR. PARKER: Right, you would just stack 19 one on top of --
20 MEMBER SKILLMAN: Only plus one?
21 MR. PARKER: Correct. And they're bolted 22 down. One bolts down on top of the other.
23 MEMBER SKILLMAN: So, it's fixed?
24 MR. PARKER: Correct.
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54 1 configuration?
2 MR. PARKER: That's right.
3 MEMBER SKILLMAN: Oh. Thank you.
4 MEMBER SUNSERI: And the airflow was 5 considered?
6 MR. PARKER: Yes. Yes. There's --
7 MEMBER SUNSERI: There's two of them right 8 now.
9 MR. PARKER: Right. Yes. And so, it's 10 not flush up against it. There is some space between 11 one and the other.
12 MEMBER SUNSERI: Okay.
13 MEMBER CORRADINI: So, remind us, since we 14 were discussing privately, without the bioshield, I'm 15 worried about radiation exposure to personnel in the 16 refueling area?
17 MR. PARKER: Right.
18 MEMBER CORRADINI: That's the issue, due 19 to scattering of radiation coming out of a module, 20 bouncing around, and --
21 MR. PARKER: And someone being across the 22 pool.
23 MEMBER CORRADINI: Right.
24 MR. PARKER: Yes.
25 MEMBER CORRADINI: But someone wouldn't be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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55 1 across the pool under operation. They would be down 2 in the refueling area. There would be nobody across 3 the pool per se, right?
4 MR. PARKER: Well, unless we're refueling.
5 MEMBER CORRADINI: Okay, but that's down 6 -- the crane does not have personnel on the crane as 7 it's moving the NMPs around.
8 MR. PARKER: But we move -- there's going 9 to be people on top of the module taking off piping 10 and disconnecting the module --
11 MEMBER CORRADINI: Oh, oh.
12 MR. PARKER: -- during the refueling 13 process.
14 MEMBER CORRADINI: Excuse me. I'm sorry.
15 Got it. Okay.
16 And then, the final question is, is there 17 a louver at the top panel for some sort of air 18 movement at the very top?
19 MR. PARKER: I believe there is some gap 20 up at the top to allow for ventilation.
21 MEMBER CORRADINI: okay. Right. We'll 22 come back to seismic. Thank you.
23 MR. PARKER: Sure.
24 MEMBER CORRADINI: Thank you.
25 MR. PARKER: So, lastly, with respect to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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56 1 3.7, we did conduct a number of sensitivity studies, 2 given the uniqueness of our design. The cases are 3 listed here, but we looked at a case where the dry 4 dock is empty, looked at a case where we had less than 5 12 modules. We looked at soil separation. We looked 6 at -- I talked before about our extended subtraction 7 use in SASSI. So, we looked at that as a comparison 8 of the direct method and, also, what we call a 9P.
9 So, we're adding planes and assessing the adequacy of 10 the 7P as well as non-vertically propagating shear 11 waves.
12 As to the results, in all cases the design 13 basis reinforcement was adequate for the sensitivity 14 studies considered. There was some slight ISRS 15 modifications in a couple of points, but those were 16 also very minor.
17 And then, in the 3.7 area, we have 15 COL 18 items, mainly to confirm the site-specific adequacy of 19 the inputs and comparison of the site-specific inputs 20 at a particular site.
21 DR. SCHULTZ: So, Josh, this is a site 22 seismic input data characterization, principally 15 23 different elements that you're going to address --
24 MR. PARKER: That's the majority, yes.
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57 1 be needed or required also?
2 MR. PARKER: There will be, yes. Yes.
3 So, we'll look at the site-specific input motion as 4 well as the swell properties, and then, confirm that 5 ISRS at a variety of locations is still bounded as 6 well as the design of the critical sections in the 7 structure.
8 DR. SCHULTZ: Based on what you've done 9 for sensitivity studies, does the COL applicant need 10 to redo sensitivity studies, as you see it?
11 MR. PARKER: I don't believe so. I mean.
12 we do have COL items that confirm that. So, there are 13 COL items to confirm, for example, soil separation.
14 DR. SCHULTZ: I was looking at that one, 15 too.
16 MR. PARKER: Yes. So, there's a number of 17 COL items that confirm these as well.
18 DR. SCHULTZ: But things like less than 12 19 modules, for example --
20 MR. PARKER: There's a COL item for that 21 as well. The COL will need to determine the module 22 loading sequence.
23 CO-CHAIR RICCARDELLA: But if the site-24 specific in-structure response spectra were below your 25 design spectra --
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58 1 MR. PARKER: Yes.
2 CO-CHAIR RICCARDELLA: -- why do you have 3 to do additional COL analysis?
4 MR. PARKER: Well, you need to -- in that 5 case, you wouldn't do anything.
6 CO-CHAIR RICCARDELLA: So, would you call 7 it a tiered evaluation --
8 MR. PARKER: Sure, yes.
9 CO-CHAIR RICCARDELLA: -- that's being 10 done?
11 MR. PARKER: Sure, yes. I mean, you would 12 only need to make modifications if you weren't 13 bounded.
14 CO-CHAIR RICCARDELLA: Okay. Thank you.
15 MR. PARKER: All right. Next, we'll move 16 on to 3.8. And I'll talk briefly about the 17 containment, and that will be talked in more detail 18 later on when we discuss Chapter 6.
19 But the containment in the NuScale design 20 is a steel vessel designed per SME Section 3, Division 21 1, Subsection NB. It's approximately 75-feet tall and 22 nominally 15 feet in diameter. It's obviously 23 slightly larger at the bolted flanges. It's operated 24 in the reactor pool in the reactor building, and the 25 internal design pressure is 1,050 psia and design NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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59 1 temperature, 550 degrees F.
2 MEMBER CORRADINI: Maybe this just is 3 Code. The outside design pressure is 60? Is that 4 just a Code requirement? I'm trying to figure out how 5 I get to 60.
6 MR. ARTHUR: A Code requirement to 7 establish an external design pressure?
8 MEMBER CORRADINI: Yes.
9 MR. ARTHUR: Yes.
10 MEMBER CORRADINI: Okay. All right. And 11 then, there were some RAIs and staff questioning about 12 internal pressure of 1050. Has that been settled or 13 are you still in discussions with staff about that?
14 I'm going to ask the staff the same question because 15 I got confused with the RAIs. No offense, but there's 16 a lot of RAIs flying back and forth that I can't 17 figure out.
18 MR. ARTHUR: The original internal design 19 pressure was 1,000.
20 MEMBER CORRADINI: So, it's up 50?
21 MR. ARTHUR: Correct.
22 MEMBER CORRADINI: The staff analyzed and 23 confirmed? Or they're in the middle of reviewing 24 that?
25 MR. ARTHUR: You'll have to ask them. I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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60 1 don't know about that.
2 MEMBER CORRADINI: Okay, fine. That's 3 good. That's fine. Thank you.
4 MR. PARKER: The last thing I'll say here 5 is that the CNV Ultimate Pressure Integrity Technical 6 Report documents are in conformance to the guidance of 7 NUREG-6909 and determined the ultimate pressure 8 capacity of 1,240 psia.
9 MEMBER CORRADINI: Maybe this is not the 10 right place, but we had somewhere -- I can't remember 11 where it was. Maybe it was in one of the chapters.
12 You're asking for an exemption for containment leak 13 rate on the whole containment, is that correct?
14 MR. PARKER: I think I'll defer that to 15 the Chapter 6 conversations. Is that --
16 MEMBER CORRADINI: Oh, good. All right.
17 Let's do that. That's where I remember reading it.
18 Thank you. Sorry.
19 MR. PARKER: That's all right.
20 All right. Next slide.
21 So, transitioning back to the structures, 22 so in 3.7, we talked about the design loads. And in 23 3.8, we discussed the design of the structures. The 24 structures are Seismic Category I and, therefore, 25 designed to ACI-349 and AISC N690. All the normal and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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61 1 extreme loads have been included for the Code-required 2 load combinations. And this allows us to determine 3 the maximum demand load for each of the structural 4 elements in the structure.
5 We have six COL items to confirm here in 6 3.8 on the site-specific acceptability of the 7 structural design as well.
8 Next slide.
9 So, the detailed information on the design 10 of the Category I structures is discussed and provided 11 in Appendix 3B of the FSAR. And that's where we 12 define the critical sections for each of the 13 structures. The critical sections are the portions of 14 the building that perform a safety-critical function 15 or are subject to large stresses or demand forces.
16 They might be difficult to construct or just generally 17 representative of the building. So, we wanted to 18 include at least one slab, one wall, a beam, a 19 buttress, a pilaster, and then, also, wanted to ensure 20 that a bay wall or a pool wall in the pool was 21 considered as well as the supports for the NPM.
22 As referenced in the previous slide, the 23 load combinations determine the maximum demands for 24 each section. And then, after determining a 25 reinforcement pattern, we can substantively determine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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62 1 the capacity. And so, we go section by section within 2 the structure to ensure that the capacity is always 3 greater than demand, so that the structure is 4 adequate.
5 Lastly, in 3.8, is 3.8.5, which is on the 6 foundations. And I'll focus here on stability, which 7 is the ability of the structures to resist sliding and 8 overturning.
9 So, given the high water table that we 10 assumed and the high input spectra, we ended up using 11 a non-linear analysis for sliding for both the reactor 12 building and the control building. And in doing so --
13 that was one of the uses of ANSYS -- we saw sliding of 14 less than an eighth of an inch for either structure.
15 So, very, very small values. We also used a non-16 linear analysis for overturning for the control 17 building and, again, saw even smaller numbers here of 18 about a 64th of an inch of total uplift in the 19 structure. So, very, very small numbers. And this 20 allowed us to conclude that the structures adequately 21 resist sliding and overturning.
22 So, with that, I'll --
23 MEMBER CORRADINI: Can you remind me, 24 sliding or overturning due to seismic events?
25 MR. PARKER: Correct.
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63 1 MEMBER CORRADINI: Oh, okay, fine.
2 MR. PARKER: Yes.
3 MEMBER CORRADINI: Thank you.
4 CO-CHAIR MARCH-LEUBA: I don't know if 5 this question belongs to you, but have you guys 6 considered expansion of the concrete over time? I'm 7 sure you're going to use the correct concrete, but, 8 even good concrete moves.
9 MR. PARKER: Sure.
10 CO-CHAIR MARCH-LEUBA: Is that part of 11 this design?
12 MR. PARKER: At this point, we were just 13 looking at nominal dimensions for the sections. So, 14 our wall thickness if 5 feet. We assume it's 5 feet 15 in the analysis. And there's ACI guidances for 16 tolerances.
17 CO-CHAIR MARCH-LEUBA: You are aware that 18 there are some operating plants that have issues with 19 expanding concrete?
20 MR. PARKER: Sure.
21 CO-CHAIR MARCH-LEUBA: And one thing that 22 they were worried about is where two buildings get 23 close together, they can start touching. Have you 24 brought enough clearances for just in case?
25 MR. PARKER: Yes. So, the buildings, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 1 between the control building and the reactor building 2 is about -- I think it's 30 or 35 feet. And then, 3 between the rad waste building and the reactor 4 building is another 20 feet. So, there's quite a bit 5 of space between --
6 CO-CHAIR MARCH-LEUBA: Sometimes there are 7 structures that are placed from a different wall.
8 Some of the issues shield, things like that. Just 9 make sure you consider that because it's been a big 10 problem somewhere else.
11 MR. PARKER: Sure.
12 MR. BRYAN: Before you get started here --
13 thank you, Josh.
14 Before we get started, I want to make one 15 clarification on the socket weld. J.J., do you want 16 to address that?
17 MR. ARTHUR: Yes. We got a note back from 18 our office. So, there are no socket welds in the 2-19 inch-diameter piping, but we do allow socket welds for 20 less than three-quarter-inch. There's some tubing 21 connecting to the ECCS valves that we do allow socket 22 welds.
23 MR. BRYAN: Thank you.
24 Okay. With that, Storm Kauffman will take 25 us through 3.6.
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65 1 MR. KAUFFMAN: All right. The remaining 2 section in Chapter 3 is 3.6, "Protection Against Pipe 3 Rupture Effects".
4 I'm Storm Kauffman. I will take you 5 through that for a few minutes.
6 While we followed the Standard Review 7 Plan, there are a number of considerations in the 8 NuScale design that require some interpretation 9 because of the differences in how NuScale's plant is 10 put together. Examples listed on this slide include 11 the smaller-diameter and shorter pipe lengths involved 12 in the NuScale plant; the fact that high-energy line 13 breaks inside the containment vessel are limited to 2-14 inch-diameter piping; the containment being designed, 15 as Josh just described, to ASME Code Section -- sorry 16 -- Section III, Class 1. The containment operates in 17 a vacuum, and it's immersed in a pool of water.
18 High-energy line break and moderate-energy 19 line break response is passive. It does not require 20 electrical power or motor-operated valves.
21 There are no concerns from a GSI-191 22 standpoint with stripping of piping insulation. In 23 fact, there's no piping insulation inside containment 24 where the ECCS is located.
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66 1 steam piping and the feedwater piping inside the 2 containment vessel that's designed to leak before 3 break criteria. In the area under the bioshield, it's 4 designed to Branch Technical Position 3-4 break 5 exclusion criteria, which I'll discuss in a minute.
6 The shorter lengths of piping and the 7 small footprint of the plant result in more congestive 8 piping arrangements that have to be considered.
9 All of the containment isolation valves 10 and decay heat removal system valves are located 11 outside containment.
12 And, of course, we have to deal with the 13 fact that the nuclear power modules are moved and the 14 F12 modules in relatively-close proximity.
15 Next.
16 We divided our high-energy line break or 17 piping break analysis into three regions of the plant.
18 The reason for this is that those three regions have 19 different environments. The systems that are higher 20 energy, high or moderate energy, are different, and 21 there are potential target SSCs that are different and 22 have different characteristics.
23 The three particular regions are inside 24 the containment vessel where we looked at specific 25 locations and arrangements, and how those affected the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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67 1 possible pipe break locations and their consequences.
2 As I said, main steam and feedwater are designed to 3 satisfy leak before break. The remaining piping is 4 relatively short length and it's all small bore.
5 There is no containment penetration area inside the 6 containment vessel.
7 We followed Branch Technical Position 3-4, 8 "Guidance for Determining Break Locations".
9 Basically, we account for small-diameter tourmaline 10 breaks, and we design the remaining piping to avoid 11 intermediate breaks.
12 In the area under the bioshield, there are 13 specific locations and arrangements considered and 14 possibility of break locations is excluded by 15 designing to Branch Technical Position 3-4 in the 16 containment penetration area to prevent terminal end 17 breaks, and throughout the area under the bioshield to 18 Branch Technical Position 3-4 to prevent intermediate-19 sized breaks. However, BTP 3-3 requires that we 20 consider non-mechanistic breaks of the main steam 21 system and feedwater system, which I'll discuss in a 22 minute.
23 Finally, the third area --
24 MEMBER CORRADINI: This is inside the hut?
25 MR. KAUFFMAN: Yes.
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68 1 MEMBER CORRADINI: Outside the 2 containment, outside the hut?
3 MR. KAUFFMAN: Correct.
4 MEMBER CORRADINI: Okay.
5 MR. KAUFFMAN: Under the bioshield that 6 you were asking about.
7 The third area of the plant that we 8 evaluate is the reactor building. There's piping 9 present in the pipe galleries and in lower floors for 10 the CVCS system. The exact piping arrangements are a 11 COL item, and therefore, we took an approach of doing 12 bounding analyses to assure that a pipe break in any 13 location, given that the arrangement is not yet 14 determined, would, in fact, have acceptable 15 consequences. And I'll describe how we do that next.
16 Non-mechanistic breaks. The largest pipe 17 in the main steam system or the largest pipe in the 18 NuScale plant is the main steam piping, which is 12-19 inch NPS. BTP 3.3 requires that a non-mechanistic 20 break in the containment penetration area of 12 --
21 sorry -- of 1 square foot be considered. That's 22 larger than the flow area of a NuScale main steam 23 pipe. So, it's a little bit illogical to apply that.
24 We looked at the likelihood of main steam 25 and feedwater system piping breaks and concluded that, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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69 1 because our piping was made of stainless steel, which 2 is more corrosion- and failure-resistant, and also 3 because it's smaller, we should still consider main 4 steam and feedwater system non-mechanistic breaks, but 5 we should scale the size.
6 Basically, what we did is we took the main 7 steam piping size, compared that to main steam piping 8 in large PWR plants, and scaled down to about from 1 9 square foot to 12 square inches for the non-10 mechanistic main steam break in the NuScale plant. We 11 also have a scaled-down feedwater piping break.
12 So, we analyzed for those piping breaks 13 under the bioshield. We don't have piping breaks, 14 double-ended circumferential piping breaks because of 15 the design to Branch Technical Position 3-4.
16 MEMBER CORRADINI: Can you say that all 17 over again? All right. Let me say it back to you.
18 So, you basically scaled off of piping sizes that are 19 larger and got down from 144 square inches to 12?
20 MR. KAUFFMAN: Correct.
21 MEMBER CORRADINI: Based on what?
22 MR. KAUFFMAN: I was agreeing -- we did 23 that because we considered our piping is less 24 susceptible to corrosion-induced failure than the 25 large PWR piping to which the non-mechanistic break NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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70 1 normally applies.
2 MEMBER CORRADINI: Yes, qualitatively, 3 I've got that part. I'm trying to understand how you 4 got down by a factor of 12.
5 MR. KAUFFMAN: I tried to show it 6 graphically here. The AP1000 main steam pipe maximum 7 diameter I believe is 38 inches.
8 MEMBER CORRADINI: Right.
9 MR. KAUFFMAN: So, 38 inches. If you look 10 at 1 square foot flow area in relation to the flow 11 area of 38-inch pipe --
12 MEMBER CORRADINI: You ratioed it off of 13 flow area?
14 MR. KAUFFMAN: Yes.
15 MEMBER CORRADINI: And then, one last 16 question. And then, let's just leave that for the 17 moment. And so now, this 12-square-inch pipe just 18 comes apart and that's what is your non-mechanistic 19 break?
20 MR. KAUFFMAN: It doesn't come apart in a 21 circumferential break. It just --
22 MEMBER CORRADINI: But in an area?
23 MR. KAUFFMAN: Yes, you just open up a 24 hole.
25 MEMBER CORRADINI: Got it. Okay, fine.
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71 1 Thank you.
2 CO-CHAIR RICCARDELLA: You don't consider 3 a guillotine break of the pipe?
4 MR. KAUFFMAN: Not under the bioshield 5 because we design to meet the criteria for break 6 exclusion of Branch Technical Position 3-3.
7 CO-CHAIR RICCARDELLA: Got it. And again, 8 you did leak before break for the dynamic loading to 9 eliminate having to consider the dynamic loading?
10 This is just to consider the LOCA transient, 11 basically, right?
12 MR. KAUFFMAN: Leak before breaks only 13 applied inside containment. We're talking outside 14 containment here.
15 CO-CHAIR RICCARDELLA: Ah, okay. Okay.
16 MR. KAUFFMAN: And I get to re-explain 17 this again using this slide.
18 So, starting inside containment, we've 19 excluded main steam and feedwater piping breaks by 20 meeting leak-before-break criteria. We consider the 21 terminal end breaks in accordance with Branch 22 Technical Position 3-3, and we exclude intermediate 23 breaks, in accordance with Branch Technical Position 24 3-3.
25 In the area under the bioshield, we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 1 consider leakage cracks and non-mechanistic breaks.
2 And in the rest of the reactor building, we basically 3 consider any break in any location that are high-4 energy or where medium-energy piping exists.
5 CO-CHAIR MARCH-LEUBA: If you are moving 6 one element to refueling position, one module, and it 7 drops and it falls right against another one, will it 8 be a possibility of breaking those lines? It's a lot 9 of mass heatings on the side.
10 MR. KAUFFMAN: I am not knowledgeable 11 about what our assumptions are on drops during 12 movement. I have to defer to --
13 CO-CHAIR MARCH-LEUBA: We assume they 14 happen with very low probability, but it's, 15 technically, they can happen --
16 MR. KAUFFMAN: Correct.
17 CO-CHAIR MARCH-LEUBA: -- by 10 to the 18 minus 11.
19 MR. KAUFFMAN: I mean, in the limit, if 20 you have to consider it, then you might possibly have 21 a break, but I did not look at that.
22 CO-CHAIR MARCH-LEUBA: That's one thing I 23 would consider to be -- either that or the seismic 24 loads -- okay.
25 MR. KAUFFMAN: I understand the question.
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73 1 CO-CHAIR MARCH-LEUBA: I'll leave it to 2 the guys who know more about this than I do.
3 MR. KAUFFMAN: Okay.
4 Next, please.
5 All right. I'm going to step through the 6 various phenomena that result from a high-energy break 7 from blast waves through subcompartment 8 pressurization, and there are a number of stops in 9 between.
10 So, starting with blast waves, when you 11 assume an instantaneous opening of a break, you get a 12 very rapid mass and energy dump into the surrounding 13 ambient environment. Because of that assumption, you 14 can form a blast or shockwave. We've concluded that 15 only steam-filled lines can generate a blast. The 16 fluid in a two-phase blowdown can't get out fast 17 enough. There is no blast if the break opening time 18 is more than a few milliseconds. So, in a way, it's 19 an artifact of the assumption that we analyze for.
20 And the NuScale smaller-diameter piping reduces the 21 mass and energy output, and therefore, the severity of 22 any blast wave that does form.
23 MEMBER CORRADINI: So, just let me ask, is 24 this set of assumptions common?
25 MR. KAUFFMAN: It's --
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74 1 MEMBER CORRADINI: In other words, steam 2 only, two-phase not, smaller because of smaller pipe 3 size, is that a set of assumptions for other blast 4 wave analyses?
5 MR. KAUFFMAN: Well, we had the small 6 piping. So, other people do not look at the smaller-7 diameter piping.
8 MEMBER CORRADINI: That part I get, but 9 your first two assumptions, if only steam-filled lines 10 generate a shockwave, two-phase discharges do not, and 11 then, the break or the opening times.
12 MR. KAUFFMAN: The opening time is an NRC 13 guidance item that you have to assume break opening 14 time of no more than 1 millisecond unless you provide 15 a technical justification. So, most break analyses 16 assume 1 millisecond or instantaneous.
17 MEMBER CORRADINI: Okay, fine.
18 MR. KAUFFMAN: And as far as the steam-19 filled lines generating blast, I don't recall if other 20 applicants have made that assertion.
21 MEMBER CORRADINI: Okay, fine. Thank you.
22 MR. KAUFFMAN: To do the blast wave 23 analysis, we concluded that we needed to do 3-24 dimensional computational fluid dynamics. We used the 25 CFX code. We qualified or verified and validated the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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75 1 use of the code using eight test problems. We did 2 have a simplified plant geometry, but one that was 3 representative of the important structures.
4 We developed a force-time history for 5 three degasification line break cases inside 6 containment and three main steam line break cases 7 inside the reactor building. In all cases, the forces 8 from the blast were very brief and acceptably low.
9 They pass very quickly and die off fairly quickly with 10 distance.
11 MEMBER CORRADINI: And probably you didn't 12 worry about obstructions that would break apart the 13 shockwave energy?
14 MR. KAUFFMAN: That's why we did 3-15 dimensional --
16 MEMBER CORRADINI: Oh, so you did?
17 MR. KAUFFMAN: -- CFD.
18 MEMBER CORRADINI: Okay, fine.
19 MR. KAUFFMAN: Because without the 3-20 dimensional CFD, if you do 2Ds --
21 MEMBER CORRADINI: Sure.
22 MR. KAUFFMAN: -- CFD --
23 MEMBER CORRADINI: It's symmetric.
24 MR. KAUFFMAN: Well, you get a shadow, and 25 you don't get the wraparound that an actual blast wave NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 1 does.
2 MEMBER CORRADINI: Okay. Thank you.
3 MR. KAUFFMAN: Next, please.
4 The next phenomena is pipe whip. In this 5 case, the smaller-diameter and shorter-length piping 6 in the NuScale plant is important again, because the 7 impact energy depends on the fluid conditions inside 8 the pipe and the length of piping. Smaller pipes have 9 lower energy content. And the congestion of the pipes 10 results in a shorter length that can whip, and 11 therefore, basically, there's less of a windup and 12 less kinetic energy on impact.
13 So, our process was to determine, first, 14 if the fluid energy within the pipe was sufficient to 15 form a plastic hinge and cause the pipe to whip. If 16 it was, we looked at motion of the pipe whip in the 17 plane, the pipe geometry. If any essential SSCs were 18 too far away to be struck, we were done; they were out 19 of range. If a large, robust SSC nearby served as a 20 barrier and prevented the pipe from striking anything 21 more vulnerable, we were also done. The best example 22 of that is inside containment. If it runs into the 23 reactor pressure vessel wall, a 2-inch-diameter pipe, 24 which is the largest pipe that we have to consider a 25 rupture for, is not going to continue. It's going to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 1 basically come to rest against the reactor pressure 2 vessel. And finally, if an impact was possible, we 3 looked at the impact force and would show that it is, 4 in fact, acceptable.
5 Next.
6 One special case that we looked at was the 7 piping gallery where there were a large number of 8 parallel piping runs of main steam piping and 9 feedwater piping. The Branch Technical Position 10 362 -- sorry -- Standard Review Plan 362 says that you 11 do not have to consider that a pipe of a given size 12 and pipe thickness causes a pipe rupture in a pipe the 13 same size of pipe thickness or larger. In other 14 words, a main steam pipe, if it rips, would not cause 15 a secondary failure in another main steam pipe because 16 they're the same size and thickness.
17 However, in the pipe gallery, we do have 18 smaller-diameter piping in the main steam piping, like 19 the feedwater system. So, we wanted to be sure that 20 we bounded the effects of possible secondary ruptures.
21 Therefore, when we did the analysis for blast waves, 22 when we did the analysis for subcompartment 23 pressurization, we looked at a main steam pipe failure 24 resulting in a whip, causing a failure of a feedwater 25 pipe or an adjoining main steam bypass line, which is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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78 1 also smaller, and used the blowdown characteristics 2 from that secondary rupture added to the first, to 3 make sure that, again, we're bounded for any future 4 analysis that the COL applicant may do, once the 5 piping arrangements are finalized.
6 Pipe whip summary. As I said, the impact 7 force depends on the thrust force of the break exit, 8 which is dependent on the size of the piping. The 9 NuScale force, just because of the large --
10 MEMBER CORRADINI: Can you go back?
11 MR. KAUFFMAN: Yes.
12 MEMBER CORRADINI: We're just trying to 13 understand the picture.
14 So, the thick pink is the missile shield 15 or is the wall that acts as a missile shield? Or is 16 the missile shield outside of that?
17 MR. KAUFFMAN: The missile shield is the 18 outside wall of the building, correct?
19 MR. PARKER: Yes. That's the, well, 20 that's the --
21 MR. KAUFFMAN: That's the wall of the 22 pool, what you see there. So, that's in the center of 23 the building.
24 MR. PARKER: So, farther down the page 25 would be the exterior --
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79 1 MEMBER CORRADINI: Okay. So, the exterior 2 wall is outside of that? That's what I was --
3 MR. KAUFFMAN: The exterior wall is down 4 here.
5 MEMBER CORRADINI: Okay. That's what we 6 were trying to understand.
7 CO-CHAIR RICCARDELLA: This is a plant 8 view here?
9 CO-CHAIR RICCARDELLA: Looking down.
10 MR. KAUFFMAN: Yes, looking down.
11 CO-CHAIR RICCARDELLA: Okay.
12 MEMBER CORRADINI: Okay. That answers the 13 question. Thank you.
14 MR. KAUFFMAN: You're welcome, yes.
15 Pipe whip, as I said, the impact force 16 depends on the thrust force of the break exit.
17 Because of the smaller-diameter pipe, NuScale's thrust 18 force is about 5 percent of that of the large main 19 steam pipe breaks in large PWRs. Again, the short 20 length of the pipe limits the impact energy. We know 21 the piping arrangements in the containment vessel.
22 So, we analyzed those for explicit pipe with motions.
23 We made conservative assumptions in the reactor 24 building.
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80 1 targets. We've got metallic components in pipes in 2 the containment and under the bioshield and we have 3 concrete structures under the bioshield and in the 4 reactor building. For the concrete structures, we 5 looked at penetration of the walls due to pipe whip, 6 using a formula from Sandia by Young, and we also 7 looked at spalling, and concluded that the concrete 8 surfaces would not be penetrated by pipe with impact.
9 MEMBER SKILLMAN: Hey, Storm, on your 10 first caret there, thrust force at break exit, 5 11 percent. Isn't the steam pressure here at about 550 12 psi?
13 MR. KAUFFMAN: Yes.
14 MEMBER SKILLMAN: And let's take a normal 15 PWR. It's 1050?
16 MR. KAUFFMAN: Yes, that also plays into 17 it.
18 MEMBER SKILLMAN: So, this is not really 19 5 percent. It's 50 percent for the same diameter 20 pipe?
21 MR. KAUFFMAN: No. No, because the thrust 22 force is integrated over the cross-sectional area of 23 the pipe. So, it's the pressure times the area.
24 MEMBER SKILLMAN: If I have a 2-inch pipe 25 at 1,000 psi, a 2-inch at 500 psi, one's 50 percent of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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81 1 the other.
2 MR. KAUFFMAN: Right, but we --
3 MEMBER SKILLMAN: You're trying to tell me 4 this is not a 36-inch line; it's a 10-4. Got it.
5 MR. KAUFFMAN: Okay.
6 MEMBER SKILLMAN: I think that's a nice 7 statistic, but it's a stretch, just saying as a steam 8 plant operator of many years.
9 MR. KAUFFMAN: Well, it's important 10 because the image you can have is taking a chopstick 11 and hitting something with it and taking a log and 12 hitting something with it. The NuScale pipe is more 13 like the chopstick; it carries less energy.
14 MEMBER SKILLMAN: I understand what you're 15 communicating.
16 MR. KAUFFMAN: Okay.
17 MEMBER SKILLMAN: I'll grant you that.
18 I'm just saying that there's another way to look at 19 it, in which case it's about half, not 5 percent, not 20 1/20th.
21 MR. KAUFFMAN: I'll move on.
22 MEMBER SKILLMAN: Thank you.
23 MR. KAUFFMAN: Thank you.
24 Next slide.
25 Okay. The next one I wanted to discuss is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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82 1 jet impingement. Impingement pressure energy depends 2 on the fluid conditions, the degree of spreading of 3 the jet, once it exits the pipe break, and the 4 distance from the break exit to the target.
5 There are different behaviors, whether the 6 jet is issuing a steam or it's two-phase liquid or, 7 for that matter, as subcooled, highly subcooled 8 liquid. In the case of steam, I'll talk about that in 9 a minute. In the case of two-phase liquid, we use the 10 approach of NUREG-2913 to estimate the pressure-11 versus-distance relationship and the spread or zone of 12 influence of the jet.
13 For the overall process, the steps we went 14 through were to determine the pressure threshold for 15 damage to essential SSCs. The big difference that the 16 NuScale plant has is that we don't have insulation 17 stripping as a concern. And therefore, we don't have 18 to worry about a jet delivering a few psi to some 19 component and stripping insulation from it. Instead, 20 we have hard targets, and those have a lot more robust 21 capability to withstand the pressure forces of a jet.
22 In looking at the zone of influence, we 23 took a conservative approach inside containment. We 24 assumed that the zone of influence is the hemisphere 25 in front of the original location of the pipe. So, if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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83 1 the pipe was pointing straight forward, we considered 2 a zone of influence anywhere in the hemisphere 3 forward-pointing.
4 The jet spreading half-angle is something 5 that the Standard Review Plan requires be addressed.
6 There's concerns that the ANS standard that is 7 referenced, or has been referenced for years, is non-8 conservative. It specifies, roughly, a 45-degree 9 spreading angle. We used 30 inside containment. That 10 was more conservative from the standpoint of 11 maintaining a higher pressure versus distance 12 downstream than if you used the 45. It was also 13 supported by the CFD calculations we did for blast 14 effects, where we found that the spread of the jet as 15 it formed after the blast departed was on the order of 16 60 degrees.
17 When the jet issues from the pipe exit, if 18 it impinges upon a large structure such as the 19 containment vessel wall, then, basically, it stops.
20 It can't go through the wall. And if there any 21 essential SSCs that are close enough to be within the 22 range of the jet, then we have to consider whether or 23 not the impingement pressure is acceptable.
24 Next, please.
25 MEMBER BALLINGER: By the way, it's not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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84 1 2913. It's NUREG-CR-2913.
2 MR. KAUFFMAN: Yes. Thank you.
3 Jet shape and pressure I'll review then.
4 I described the zone of influence for inside 5 containment. For pressure, I also mentioned 30 6 degrees. I forgot to mention that, after five 7 diameters downstream, we take that back to a 10-degree 8 spread, so it stays more concentrated.
9 In the reactor building, the arrangements 10 for the piping do not get finalized. So, a jet could 11 end up pointing anywhere, and that's what we assumed.
12 We also assumed that the target SSC is only 2 L over D 13 from the break exit. So, there's basically no 14 reduction in pressure with distance.
15 Next.
16 There's one specific portion of 362 that 17 requires evaluation of the potential for amplification 18 of the pressure forces due to a resonance. There's 19 been a lot of research on this subject. It's pretty 20 much concluded that to have a resonance, you must have 21 a symmetric or axisymmetric jet and you have to 22 develop what one set of researchers called a phased 23 lock, which basically is a stable condition that you 24 form a standing wave between the exit of the jet and 25 its target.
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85 1 In a high-energy line break in a plant 2 like NuScale's, there are multiple reasons why 3 resonance will not occur. Those include the fact that 4 the fluid is not a single-phase gas. It's condensable 5 fluid, and the proximity to condensation or 6 evaporation results in a damping effect that basically 7 takes energy out of the jet. The distorted exit 8 geometry violates the axisymmetric condition and 9 prevents developing a phase lock. The absence of a 10 large, flat impingement surface limits the ability to 11 return the energy to the break exit nozzle area, and 12 also eliminates the possibility of a resonance.
13 The instability of the jet, the fact that 14 the pipe is whipped and is moving again disrupts the 15 stability necessary to form a resonance. There are 16 intersecting obstacles that disrupt the jet, and there 17 also are fluid spray conditions that reenter the jet 18 to disrupt it. And finally, there's a frequency 19 mismatch with the structures. The jet has a frequency 20 on several kilohertz, the tens of kilohertz range, 21 which is not going to excite the structures which have 22 natural frequencies on the order of hertz or tens of 23 hertz.
24 Next, please.
25 DR. SCHULTZ: Storm?
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86 1 MR. KAUFFMAN: Yes?
2 DR. SCHULTZ: This last evaluation with 3 the bullets, is that a rationale that demonstrates 4 that the research papers don't apply or have you done 5 an analysis that demonstrates it because of these 6 elements?
7 MR. KAUFFMAN: The research papers apply 8 to single-phase gas jets. There is no research paper 9 that I could find that ever reported finding a 10 resonance or amplification condition for a steam jet.
11 MEMBER CORRADINI: For a what?
12 MR. KAUFFMAN: A steam jet.
13 MEMBER CORRADINI: A steam jet?
14 MR. KAUFFMAN: Anything that involves 15 being close to saturation conditions. And, in fact, 16 there's been a lot of work because the gas jets cause 17 problems called screech and are something that the 18 industry has been trying to avoid in a number of 19 applications. There's been a lot of research on how 20 to avoid the amplification, and they do things such as 21 put little tabs on the exits of the nozzle to disrupt 22 the axisymmetric conditions. They also inject fluid 23 or liquid into a single-phase gas jet to help damp the 24 oscillations. So, the things that the researchers 25 have done to prevent or reduce the amplification in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 1 single-phase gas jets are the sort of things that 2 exist naturally in a high-energy line break.
3 DR. SCHULTZ: And so, here are a number of 4 reasons/rationale why it's not a concern for the 5 application here?
6 MR. KAUFFMAN: Yes.
7 DR. SCHULTZ: Thank you.
8 CO-CHAIR RICCARDELLA: Excuse me. Could 9 you go back a couple of slides to slide 46, please?
10 MR. KAUFFMAN: Okay.
11 CO-CHAIR RICCARDELLA: That top bullet, 12 would you explain the meaning of NPS 2 only; 13 therefore, no longitudinal breaks?
14 MR. KAUFFMAN: Okay. Inside the 15 containment vessel, there is main steam and feedwater 16 piping that is leak before break --
17 CO-CHAIR RICCARDELLA: Okay.
18 MR. KAUFFMAN: -- leaving only 2-inch NPS 19 piping.
20 CO-CHAIR RICCARDELLA: Yes, yes.
21 MR. KAUFFMAN: By the guidance for when 22 you have to consider longitudinal breaks, 2-inch or 23 NPS 2 piping does not have to be considered for a 24 longitudinal break. So, we considered circumferential 25 offset breaks, but not longitudinal.
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88 1 CO-CHAIR RICCARDELLA: That's NRC 2 guidance?
3 MR. KAUFFMAN: Yes.
4 CO-CHAIR RICCARDELLA: Okay. What is it, 5 a Reg Guide or an SRP?
6 MR. KAUFFMAN: I think it's in Standard 7 Review Plan 362. It may be in the Branch Technical 8 Position 3-4. Right now, I'm not sure which.
9 CO-CHAIR RICCARDELLA: Okay. Thank you.
10 MR. KAUFFMAN: Okay. So, in summary, to 11 go back over jet impingement, the jet impingement 12 depends on conditions that are less severe for the 13 NuScale plant design. We've looked at spreading angle 14 for the underexpanded jet and the distance to the 15 target SSCs. Concluded that there is acceptable 16 performance, in other words, no potential for 17 unacceptable damage to essential SSCs inside 18 containment. There are no jets under the bioshield 19 because non-mechanistic breaks don't have to consider 20 jets. And there's no dynamic amplification of jets.
21 Finally, in the reactor building, we've 22 done a bounding analysis to account for whatever final 23 piping arrangements the COL applicant comes up with.
24 The last area of effects of subcompartment 25 pressurization. For this, we performed GOTHIC NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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89 1 analyses for the various high-energy line breaks in 2 different areas of the plant. Inside the containment 3 vessel, any high-energy line break transient is 4 bounded by the ECCS initiation transient, and 5 therefore, did not require further evaluation.
6 Under the bioshield, the high-energy line 7 breaks are limited to the non-mechanistic breaks of 8 the main steam and feedwater system. We provided 9 passive venting. That was what was being shown on 10 slide 31 with the louvers on the bioshield. So, that 11 passive vent path limits the temperature and pressure 12 buildup under the bioshield. There's a structural 13 limit of 1 psi that's met, and the temperature 14 envelope is used, then, to determine the environmental 15 qualification envelope for equipment under the 16 bioshield.
17 In the reactor building, we also have vent 18 paths to limit overpressurization or differential 19 pressures across walls and floors. We looked at 20 various breaks for the piping systems in different 21 locations in the building, and have shown that the 22 structural limit of 3 psid is met throughout the 23 building.
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90 1 in the NuScale plant are limited by application of 2 break exclusion criteria from Branch Technical 3 Position 3-4 and application of leak before break.
4 We evaluated blast effects using 3-5 dimensional computational fluid dynamics. We 6 evaluated pipe whip, jet behavior, including jet 7 spreading and jet impingement. We've looked at 8 subcompartment pressurization using GOTHIC.
9 And finally, we have three COL items that 10 require that the final pipe rupture hazards assessment 11 confirm our conclusions and that the final 12 arrangements determined by the COL applicant also 13 satisfy our criteria.
14 One remaining slide to complete Chapter 3 15 is leak before break. NuScale, as I said, applied 16 leak before break to the 12- and 4-inch, 4- and 5-inch 17 large-diameter piping inside the containment vessel.
18 This piping is all, basically, stainless steel 304 19 dual-certified, which is corrosion, erosion, and 20 fatigue, and, in general, failure-resistant.
21 In accordance with NRC criteria, we 22 applied a margin of 10 on detectable leak rate and a 23 margin of 2 on flaw size. We detect leakage by 24 changes in containment vessel pressure and by changes, 25 accumulation of condensate in the containment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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91 1 evacuation system sample vessel that's constantly 2 drawing a suction on the containment vessel.
3 And at that point, we're into the 4 acronyms.
5 CO-CHAIR RICCARDELLA: Storm, compared to 6 a normal PWR or BWR, your leak detection capability is 7 very small, right?
8 MR. KAUFFMAN: Is capable of detecting 9 very small leaks, yes.
10 CO-CHAIR RICCARDELLA: Very small leaks, 11 yes.
12 MR. KAUFFMAN: The containment evacuation 13 system sample vessel is kind of an integrating 14 measure. It's constantly accumulating condensate.
15 So, you can measure the rate of rise. And the 16 containment vessel pressure, because we're at a 17 vacuum, it's very sensitive to anything that puts more 18 mass in the containment.
19 CO-CHAIR RICCARDELLA: So, refresh my 20 memory, what is the gpm detectability that you claim?
21 MEMBER CORRADINI: If it's public.
22 MR. KAUFFMAN: That I don't know.
23 CO-CHAIR RICCARDELLA: Okay. It's not 24 public?
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92 1 remember what it is?
2 MR. KAUFFMAN: That's right.
3 MEMBER SKILLMAN: I think it is public, 4 and it's in, actually, 5.2, and this is what is 5 written: "The containment pressure monitoring 6 instrumentation is capable of detecting a minimum leak 7 rate from the RCS of .007 gpm and the containment 8 pressure will increase" -- and this is the kicker --
9 "containment pressure will increase .1 psi in less 10 than one minute with a 1-gpm leak from the RCS."
11 This is where I was digging. This is 12 5.2.5.1, Revision 2. So, it's really, really tight, 13 remarkably tight. And it's that vessel that Storm 14 just discussed, the CES, the evacuation system vessel, 15 which is a constant draw into the -- I guess it's a 16 flask of some sort. It's measured --
17 CO-CHAIR RICCARDELLA: So, that's the more 18 sensitive of the two?
19 MEMBER CORRADINI: You're going to turn to 20 somebody else to tell you the answer to that?
21 MR. KAUFFMAN: Yes. Yes, I'm not the guru 22 on the containment evacuation system.
23 CO-CHAIR MARCH-LEUBA: I would expect the 24 pressure to be the most sensitive of all because 25 you're under vacuum. You put a gram of water and, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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93 1 poof, the pressure goes up.
2 The problem is if you develop a second 3 leak, then the pressure is already high. You have 4 your vacuum system going and you cannot detect it.
5 So, you have to rely on the drain. One leak, 6 pressure; two leak, drain.
7 MEMBER SKILLMAN: Yes, which one is it, is 8 the question. You know, what's going on in there?
9 You can get one or two.
10 CO-CHAIR MARCH-LEUBA: Once you have one, 11 essentially, you're blind. You don't know if you have 12 two or three or four.
13 MEMBER SKILLMAN: Or a simmering relief 14 valve.
15 CO-CHAIR MARCH-LEUBA: Yes, but you're in 16 a blind condition.
17 MEMBER BLEY: You're not so blind that you 18 don't know there's a leak.
19 MEMBER SKILLMAN: Pardon?
20 MEMBER BLEY: I said, you're not so blind 21 that you don't know there is a leak.
22 MEMBER SKILLMAN: That's correct, yes.
23 MEMBER CORRADINI: Other questions for 24 NuScale?
25 Oh, excuse me.
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94 1 MR. HOUGHTON: I didn't hear a specific 2 question there. I was just stepping up in case we had 3 a point-specific on the containment evacuation.
4 (Laughter.)
5 MEMBER CORRADINI: Yes, stand at the 6 ready.
7 Any questions? Any further questions by 8 the members?
9 DR. SCHULTZ: Just one question. Storm, 10 you talk about the three COL items. The way you 11 describe the analyses and the evaluation you've done, 12 you really don't expect that it will be difficult --
13 that the arrangements would be any different than it 14 would be standard practice for the COL applicant to 15 check the boxes, if you will, to demonstrate that the 16 evaluations are met?
17 MR. KAUFFMAN: Correct. We did as much as 18 possible at this stage to assure that the COL 19 applicant would have a clean path to show acceptance.
20 DR. SCHULTZ: Good. Thank you.
21 MEMBER CORRADINI: Mike?
22 MR. SNODDERLY: I just wanted to 23 apologize; we didn't have the sign-in sheets out the 24 first thing this morning. So, they're out now. And 25 so, if people could remember in the audience from the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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95 1 public and from staff to sign the sign-in sheets, we'd 2 appreciate it.
3 MEMBER CORRADINI: Okay, okay. All right.
4 Other members, questions?
5 Otherwise, we'll take a break. 10:30.
6 Thank you.
7 (Whereupon, the above-entitled matter went 8 off the record at 10:14 a.m. and resumed at 10:30 9 a.m.)
10 MEMBER CORRADINI: Do you want to start us 11 off?
12 MS. VERA: Sure, sure.
13 Good morning, everyone. My name is 14 Marieliz Vera. I'm Project Manager for Chapter 3 for 15 this application.
16 Today, we're going to present a Chapter 3, 17 "Design of Structural Components, Equipment, and 18 Systems".
19 The first couple of slides is the list of 20 technical reviewers. Chapter 3 is a long one. So, we 21 have a couple of reviewers here. We have most of them 22 in the audience.
23 We're going to focus our presentation 24 today in sections that have open items or that we know 25 that the ACRS has shown some interest. So, here we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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96 1 have the sections that are not going to be presented 2 because they don't have any open items. But if you 3 have any questions, here we have the staff to respond 4 to any questions you have on any of these sections.
5 So, we're going to start our presentation 6 with Section 3.6.3, and we're going to have a couple 7 of groups presenting. So, you're going to see a 8 little bit of movement here. We're trying to make it 9 as smooth as possible.
10 And the first section we're going to 11 present is leak before break, evaluation procedures, 12 with Eric Reichelt.
13 MR. REICHELT: Good morning.
14 My name is Eric Reichelt, and I'm a Senior 15 Materials Engineer in the Office of New Reactors.
16 This week I'm in the Division of Engineering, Safety 17 Systems and Risk Assessment, the Materials and 18 Chemical Engineering Branch. I am the technical 19 reviewer for Section 3.6.3, "Leak Before Break 20 Design," for the NuScale DCD.
21 I would like to give you a brief 22 presentation on the work that has and will be 23 performed for this particular section. The LBB 24 approach for new reactors to use bounding limits 25 established during the design certification phase, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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97 1 to verify the final as-built design during the 2 construction phase using ITAAC, was approved by the 3 Commission in its SRM for SECY-93087.
4 The use of LBB applications has been 5 utilized in previous design certification 6 applications. However, the leak-before-break 7 evaluation for the NuScale design posed some unique 8 challenges because of the piping diameters and complex 9 nature of the piping layouts involved. In addition, 10 the sensitivity of the NuScale leakage detection 11 system, which you heard about in Section 5.2.5, is 12 also significantly different from traditional large 13 lightwater reactor designs and will play a major role 14 in a NuScale design being able to meet established LBB 15 criteria.
16 Because of these complexities, the 17 technical review is being performed by Engineering 18 Mechanics Corporation of Columbus, otherwise known as 19 EMC2, our contract for this review, who as an 20 organization has decades worth of experience with the 21 NRC-endorsed approach to leak before break.
22 And at this time, I would like to 23 introduce the EMC staff who has assisted the NRC for 24 the review of NuScale LBB analysis, and in my opinion, 25 are the leading experts in the field of leak before NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 1 break and fracture mechanics.
2 To my right I have Dr. Gery Wilkowski, 3 who's the President of EMC2. Next is Dr. Prabhat 4 Krishnaswamy, Senior Research Leader and principal 5 investigator for the NuScale project. And finally, 6 Dr. Mohammed Uddin, who is the principal engineer.
7 Next slide.
8 The NuScale design requested to apply 9 leak-before-break methodology to main steam piping and 10 the feedwater piping systems. The main steam piping 11 is NPS 8 and NPS 12. The feedwater piping is NPS 4 12 and NPS 5.
13 Unique aspects for NuScale is the curved 14 piping system and making sure fabrication, i.e., pipe-15 bending, limits cold-working to an acceptable limit, 16 and the methods and criteria to evaluate LBB are 17 consistent with the guidance in SRP-363 and 18 NUREG-1061, Volume 3.
19 Next slide, please.
20 The staff has reviewed the applicable 21 NuScale DCD subsections in Section 3.6.3. We're 22 reviewed the DCD references for applicability and use.
23 We've held public meetings with the NuScale staff 24 about technical issues and RAIs leading to proposed 25 DCD markups. The technical issues and response by NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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99 1 NuScale to RAIs were acceptable and were, therefore, 2 closed.
3 The staff is currently reviewing the 4 methodology and will perform a confirmatory analysis 5 on the feedwater system proprietary information and 6 data which has been provided by NuScale, and EMC2 is 7 in the early stages of reviewing the information and 8 eventually performing the confirmatory analysis.
9 The staff will also review the methodology 10 and will perform a confirmatory analysis on the main 11 steam lines when NuScale provides the proprietary 12 information and data, which is tentatively scheduled 13 for July 15th, 2019. Based on the confirmatory 14 analysis, this is being tracked as a confirmatory 15 item.
16 This concludes our presentation at this 17 time.
18 MEMBER CORRADINI: Questions by the 19 Committee?
20 Okay. Keep on going. I love your name.
21 Sorry.
22 MS. VERA: The structural people, please 23 come.
24 Okay. Now we're going to continue with 25 mostly structural, Section 3.5.1.3, "Turbine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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100 1 Missiles," and Section 3.5.3, "Barrier Design 2 Procedures," that John Honcharik and BP Jain will 3 start.
4 MR. HONCHARIK: Good morning.
5 My name is John Honcharik, and I'll talk 6 about turbine missiles. Basically, the regulatory 7 basis for it is GDC 4 requires the SCCs, important 8 safety be protected against environmental effects, 9 including missiles.
10 The safety-related, risk-significant SCCs 11 in the NuScale design are located in the rack building 12 and the control room building. The turbine generators 13 are unfavorably oriented with respect to the reactor 14 building and control room building.
15 To meet GDC 4, NuScale proposes to use the 16 installed or existing structures to protect these 17 essential SCCs.
18 MEMBER BLEY: John, is this, indeed, the 19 first time anyone has used that approach for a turbine 20 missile protection?
21 MR. HONCHARIK: Yes. To my knowledge, 22 yes, that is correct.
23 MEMBER BLEY: Okay.
24 MR. HONCHARIK: It's the first time.
25 Next slide.
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101 1 And right now, basically, I'm looking at 2 the turbine missile parameters, and BP, to my right, 3 is going to be looking at assessing how those missile 4 parameters will affect the structures to protect the 5 essential SCCs.
6 And for my section, there's one open item.
7 Basically, NuScale didn't use the full spectrum of 8 turbine missiles, including the size, weight, and 9 speed. They said they were using a turbine blade, 10 which basically isn't a missile because you could 11 throw blades most of the time.
12 We're concerned with the rotor. And we 13 basically said up to half of the last stage of the 14 rotor.
15 NuScale --
16 MEMBER CORRADINI: What is that based on?
17 Half of the last stage of the rotor, and what's the 18 weight of that?
19 MR. HONCHARIK: Yes, basically, from past 20 experiences of turbine shafts failing, there are 21 pretty big sizes, including up to a half of the rotor, 22 half of the last stage of the rotor.
23 MEMBER CORRADINI: Right, but these are 24 monoblock rotors?
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102 1 time, they are. When we first proposed this, we 2 weren't sure what they were.
3 MEMBER CORRADINI: All right. Okay, but 4 what I'm trying to get at is, I'm trying to unwrap.
5 So, the open item is what they've assumed so far is 6 not considered bounding or it's not an appropriate 7 spectrum but it is bounding? That's what I'm not sure 8 about.
9 MR. HONCHARIK: It was not an appropriate 10 spectrum.
11 MEMBER CORRADINI: Okay.
12 MEMBER BLEY: And therefore, was not 13 bounding.
14 MR. HONCHARIK: Right.
15 MEMBER CORRADINI: So, their 3,000-pound 16 portion of the rotor is not big enough?
17 MR. HONCHARIK: Well, they just submitted 18 that a couple of weeks ago.
19 MEMBER CORRADINI: Oh, okay. Excuse me.
20 Okay, fine.
21 MR. HONCHARIK: Yes.
22 MEMBER BLEY: You have not yet had a 23 chance to really review that, have you?
24 MR. HONCHARIK: No, we have not.
25 MEMBER CORRADINI: Okay. That helps me.
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103 1 I didn't put the 2 and 2 together. Thank you.
2 MR. HONCHARIK: Yes, initially, they came 3 in with an analysis. We did an audit, and they only 4 had an analysis of the blade, and a blade with a part 5 of the rotor, but it was only like 30 pounds.
6 MEMBER CORRADINI: Sure. Got it.
7 MR. HONCHARIK: So, we said, no, that's 8 not a turbine missile.
9 MEMBER CORRADINI: Okay.
10 MR. HONCHARIK: You've got to include 11 something -- you have to come up with -- you know, 12 because part of this turbine missile for barriers, you 13 have to come up with an actual missile. So, what is 14 the bounding missile. Okay? Based on some operating 15 experience, you could see that it could be up to at 16 least the last stage of the rotor. And the last stage 17 is usually the larger piece.
18 MEMBER SKILLMAN: John, are you convinced 19 that that disc portion, which is the last piece of the 20 rotor, is bounding?
21 MR. HONCHARIK: Yes, pretty much. I mean, 22 based on what we've seen and, also, the size of their 23 rotor -- you know, it's only a 50-megawatt. And 24 usually, the last stages are a lot bigger than the 25 rest of it.
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104 1 MEMBER SKILLMAN: All right. But you're 2 comfortable that that piece is bounding? That's where 3 I'm going with my question.
4 MR. HONCHARIK: Yes.
5 MEMBER SKILLMAN: Okay. So, the path 6 forward is demonstrating that that piece not 7 penetrating is sufficient for this adverse design?
8 MR. HONCHARIK: Yes, not penetrating or 9 backscabbing, which BP will talk about that.
10 MEMBER SKILLMAN: Okay. Thank you, John.
11 MEMBER CORRADINI: Does that consider the 12 generator, too?
13 MR. HONCHARIK: No, this is just the 14 turbine, not the generator.
15 MEMBER CORRADINI: Okay.
16 MEMBER BALLINGER: Has there ever been a 17 failure of a monoblock rotor in any way?
18 MR. HONCHARIK: Well, monoblock rotors 19 have only been, I guess -- what? -- maybe 15 years or 20 so, 20. So, right now, there has not been any that we 21 know of. I mean, for nuclear applications.
22 MEMBER CORRADINI: But nothing makes this 23 unique to nuclear in terms of being a monoblock rotor.
24 They're used in other -- in fossil facilities, I would 25 assume, I thought.
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105 1 MR. HONCHARIK: They've been using --
2 MEMBER CORRADINI: Monoblock rotors in 3 small fossil plants? No? I thought so.
4 MR. HONCHARIK: Yes, they have.
5 MEMBER CORRADINI: I'm looking at Ron 6 because --
7 MR. HONCHARIK: Yes, they have.
8 MEMBER BALLINGER: Yes, I just don't know 9 that there's ever been a failure.
10 MEMBER BLEY: I don't, either, but the 11 Applicant chose to not look at the probabilistic side 12 of this and to look at the missile protection by the 13 barriers. So, I think --
14 MEMBER BALLINGER: Because there have been 15 breaks -- breaks generally, one in like 1960, one in, 16 say, the 1970-something, where the composition of the 17 material and the cleanliness of the material underwent 18 big changes.
19 MR. HONCHARIK: Right.
20 MEMBER BALLINGER: And so, if you look at 21 failures of rotors as a function of time, and you put 22 those in those bins, you'll find out that failure of 23 rotors after 1975 for a manufacturer, very rare.
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106 1 integrity, making sure the processes that you use to 2 make the rotor --
3 MEMBER BALLINGER: Yes.
4 MR. HONCHARIK: -- make sure you didn't 5 have impurities, right? Because a lot of them, you 6 know, Hinkley and stuff had mainly sulfide inclusions, 7 right? And that's what basically -- and that was a 8 solid rotor, right?
9 MEMBER BALLINGER: That was lousy 10 material.
11 (Laughter.)
12 MR. HONCHARIK: Correct. But, right now, 13 NuScale does not credit anything.
14 MEMBER BALLINGER: Okay.
15 MR. HONCHARIK: They have no overspeed 16 protection that is credited.
17 MEMBER BALLINGER: Yes.
18 MR. HONCHARIK: There is no material 19 properties that are credited, inspections, or 20 anything.
21 MEMBER BALLINGER: Well, it's extremely 22 conservative. So, their analysis would be extremely 23 conservative.
24 MR. HONCHARIK: Right. Yes. So, that's 25 why, basically, they're assuming it's going to fail, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 1 and now you have to have that missile to see whether 2 or not it could be prevented to penetrate or backscab, 3 and so on, on the barrier.
4 And like I said, we just received this.
5 So, right now, we're currently reviewing this 6 information. And basically, it is that RAI responses.
7 There's no Technical Report attached to it. So, if 8 you look for a Technical Report, it's probably not 9 going to be there.
10 And I don't know if you had any other 11 questions on the turbine missile size.
12 MEMBER BLEY: Well, you're looking at 13 me --
14 MR. HONCHARIK: Yes.
15 MEMBER BLEY: -- and I had a lot of 16 questions. But I think, until we see those enclosures 17 -- and we just learned there's about eight enclosures, 18 and no idea which of those have the technical 19 information. So, we haven't seen any of it yet.
20 MR. HONCHARIK: Yes.
21 MEMBER BLEY: We need to look at that 22 first.
23 MR. HONCHARIK: Okay.
24 MEMBER BLEY: If you have anything to say 25 about what's in those enclosures and where you stand, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 1 that would be interesting, but I don't think you need 2 to do that yet, if you haven't --
3 MR. HONCHARIK: Yes, right now, we really 4 haven't had time to look at it, other than it's 5 responses to the RAIs and not a Technical Report.
6 MEMBER BLEY: This is more a process 7 question. Given this is the first time anybody has 8 taken this approach, I guess I'm a little surprised 9 they didn't, and you didn't ask for a Technical Report 10 on this, but I guess you have the analysis now to look 11 at.
12 MR. HONCHARIK: Well, I think part of 13 reviewing this, part of the RAI was provide the basis 14 for this, okay, determine the spectrum missiles and, 15 also, the analysis of the barrier, and document it and 16 provide the analysis. So, I think we're still looking 17 for that. Okay? So, it's not to say that, you know, 18 because, typically, for even a probabilistic way, you 19 have a Technical Report that supports that basis.
20 MEMBER BLEY: Yes.
21 MR. HONCHARIK: So, we're going to have to 22 look and to see whether or not they have sufficient 23 information. If not, we're going to have to ask for 24 more, whether that's a Technical Report or they have 25 to provide more information.
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109 1 MEMBER BLEY: We look forward to hearing 2 about it after you've had a chance to look and find 3 out where it's going.
4 MR. HONCHARIK: Okay. And with that, I 5 turn it over to BP to address the battery analysis.
6 MR. JAIN: Okay. So, this is BP Jain, 7 obviously, and I'll be talking about the barrier 8 design for the postulated missiles. There are three 9 missiles NuScale has postulated, as NuScale has said 10 before. One is a turbine blade, which weighs about 32 11 pounds. The second to the turbine blade was a rotor 12 fragment, 52 pounds, and half of the last-stage rotor 13 which weighs about 3,000 pounds.
14 MEMBER BLEY: Can you explain to us why 15 there's different velocities used for all three cases?
16 MR. JAIN: No, I cannot. Structurally, 17 our job is just to look, given a missile, what it will 18 do to the barrier. So, that's where I come in.
19 MEMBER BLEY: So, who are the guys who 20 look at whether the missile's postulated correctly?
21 MR. JAIN: My friend on the left.
22 (Laughter.)
23 MEMBER BLEY: Okay.
24 MR. HONCHARIK: Yes. I was going to do, 25 I was doing the turbine missile spectra, which NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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110 1 includes the weight, velocity, and speed. So, I 2 think, right now, I still haven't looked at it. But, 3 depending on -- the blade is further out. So, it's 4 actually spinning at a little higher RPM, I mean a 5 higher velocity.
6 MEMBER BLEY: Wait a minute.
7 (Laughter.)
8 MR. HONCHARIK: The same RPM, but a higher 9 velocity --
10 MEMBER BLEY: Okay.
11 MR. HONCHARIK: -- because of the radius, 12 the distance out.
13 MEMBER BLEY: Yes.
14 MR. HONCHARIK: So, that's why I think 15 you'll see different speeds.
16 MEMBER BLEY: So, that's a blade velocity 17 on the first two?
18 MR. HONCHARIK: Correct.
19 MEMBER BLEY: And they said they did it at 20 normal running speed at 150 percent and I think 220 21 percent, or something like that?
22 MR. HONCHARIK: Yes, they did like a 23 sensitivity analysis. I think they said their design 24 overspeed is 160 percent, which would be destructive 25 overspeed.
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111 1 MEMBER BLEY: Uh-hum.
2 MR. HONCHARIK: So, that's what pretty 3 much we're looking for as the bounding case, but they 4 also went up to like 220 percent.
5 MEMBER BLEY: Okay. So, you have all of 6 those to look at? Thank you.
7 MR. JAIN: Okay. So, given these 8 missiles, the staff used the guidance in 3.5.3 to the 9 design.
10 MEMBER BLEY: I'm sorry, this is just --
11 it's an interesting point, because if that's the 12 proposed destructive overspeed, but nobody's oversped 13 one of these and caused it to break, if, in fact, it 14 holds together better than you expected, it doesn't 15 come apart until a much higher RPM, it's sure not 16 conservative to just look at that destructive 17 overspeed.
18 MR. HONCHARIK: Correct, and that's why 19 part of the RAI was provide your basis for determining 20 their design overspeed. Because, like you said, 21 depending on the material that you're using, you could 22 have a higher destructive overspeed. If it's a crappy 23 material, you may have a lower destructive overspeed 24 versus a better.
25 MEMBER BLEY: Okay.
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112 1 CO-CHAIR MARCH-LEUBA: And how are you 2 going to handle that? Because this is like the better 3 you design your turbine, the worse your building is 4 going to be, right?
5 MR. HONCHARIK: Right.
6 MEMBER BLEY: So, you take no credit for 7 the overspeed trip, yes.
8 MR. HONCHARIK: Right, or the materials.
9 But they did do a sensitivity analysis up to 220, 10 which more than likely 220 is extremely high. I mean.
11 I don't think they've had one or two destructive 12 overspeeds where it was, I think, to 165, or something 13 like that.
14 MEMBER BLEY: But all of those were 15 keyed --
16 MR. HONCHARIK: Most of those were keyed, 17 correct.
18 MEMBER BLEY: I think not most. I think 19 all of them.
20 MR. HONCHARIK: Right. Well, for 21 destructive overspeed.
22 MEMBER BLEY: Yes.
23 MR. HONCHARIK: There have been others 24 that, like the other ones where I said that it was due 25 to material.
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113 1 MEMBER BLEY: Yes.
2 MR. HONCHARIK: And it actually flew apart 3 because of the inclusions and lower fracture 4 toughness.
5 MEMBER BLEY: Uh-hum. Okay.
6 MR. JAIN: Okay. So, the staff used the 7 guidance and started with 3.5.3 to review the better 8 design procedure for local and global loading effects.
9 And acceptance criteria, whether it meets the barrier 10 design is primarily to see if the wall is thick enough 11 to provide the backface scabbing.
12 Now in its SER, the staff is required to 13 make the following review findings: one, that the 14 procedure used for barrier design for the impact of 15 design basis turbine missiles are acceptable. And 16 acceptance criteria is, obviously, the staff guidance 17 is one of the documents staff uses for that.
18 And the other part is that information 19 presented in the DCD provides a reasonable assurance 20 that the reactor control building walls provide 21 adequate protection to essential SSCs.
22 So, those are the statements, the findings 23 we have to arrive at in our Safety Evaluation when we 24 are all done.
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114 1 find complete and technically-sufficient information 2 in the DCA. So, the staff asked questions, and we 3 have a submittal and the staff is going to be 4 reviewing that information.
5 MEMBER BLEY: So now, instead of just 6 using the previous guidance you had, you're going to 7 have to have somebody, if I understood them correctly, 8 look at their finite element analysis.
9 MR. JAIN: I'll come to that. Okay?
10 MEMBER BLEY: Okay. You're going to come 11 to that? Very good.
12 MR. JAIN: Yes. So, on the next slide, 13 I'm going to talk about some of the overview of the 14 key design issues, what makes this open item.
15 Well, one of the key reasons, obviously, 16 we didn't have the information to look at. So, we 17 can't come to a conclusion in terms of a safety 18 finding. But whatever we looked at so far, there were 19 certain areas which staff had concerns with. And so, 20 I'll go one by one.
21 One, the finite element procedure that 22 NuScale had used to calculate the penetration depth.
23 Staff has never looked at it before. There's no 24 precedence of using that approach for better design 25 for these high-speed missiles. So, staff will review NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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115 1 the validation and verification of the procedures, the 2 approach, and the computer program used to arrive at 3 those conclusions.
4 The other issue staff have, that the local 5 and global damage assessment was not provided for all 6 the three missiles. So, the staff did look at -- they 7 did provide the local damage for the first two 8 missiles, but there was no global assessment.
9 MEMBER CORRADINI: What do you mean by a 10 global assessment?
11 MR. JAIN: Global assessment means --
12 local is like punching and scabbing. It's a local 13 panel. But when you talk about global, it's when you 14 apply the force. There has to be an energy balance.
15 So, you apply kinetic energy, half mv-squared that 16 needs to be dissipated.
17 MEMBER CORRADINI: Somewhere?
18 MR. JAIN: And so, deformation of the 19 overall --
20 MEMBER CORRADINI: So, you want to know 21 what went in equals what --
22 MR. JAIN: Right.
23 MEMBER CORRADINI: -- got somewhere else?
24 MR. JAIN: Exactly.
25 MEMBER CORRADINI: Okay, fine. Thank you.
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116 1 DR. SCHULTZ: Bhagwat, are you 2 anticipating a need for you to perform confirmatory 3 analysis or do you feel that, in what you've seen so 4 far that is being supplied, you can do the review of 5 what the Applicant has done and --
6 MR. JAIN: Well, staff intends to review 7 what the Applicant has presented. At this point, we 8 don't plan to have any confirmatory analysis, unless 9 we find things which really require that.
10 DR. SCHULTZ: You've got a lot of careful 11 review that needs to be done --
12 MR. JAIN: That is correct.
13 DR. SCHULTZ: -- given that it's a new 14 application, and you've got the validation and 15 verification to look at, too.
16 MR. JAIN: New application, new approach.
17 There's no precedence.
18 MEMBER BLEY: And in all the parameters in 19 that finite element method, there's going to be a lot 20 of uncertainty. So, how they dealt with that is 21 something I hope you're looking at closely.
22 MR. JAIN: Right. And we have asked some 23 of those questions, and I understand that NuScale has 24 responded to those as well.
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117 1 design issue was the penetration depth. And that, 2 it's counterintuitive to the point that what they have 3 reported, that if you have a softer missile or 4 deformable missile, the penetration of that is reduced 5 by one-third compared to the missile that was rigid.
6 Now, granted, there will be a reduction in penetration 7 depth because the missile gets deformed and not all 8 the energy is input to the barrier. But, from what we 9 have seen in the literature, generally, the reduction 10 is of the order of 30-40 percent. Again, that's just 11 a number. But this is almost 300 percent. So, we 12 need to understand whether their models are okay or 13 the validation with the test results, and so on, that 14 they all makes sense.
15 MEMBER SKILLMAN: BP, do you have the 16 tools, does the staff have the tools that are required 17 to come to a determination on this topic?
18 MR. JAIN: Well, I mean, the NRC has 19 infinite resources. So, to answer your question, we 20 have the resources, but whether we will go that route, 21 it's premature at this point.
22 MEMBER CORRADINI: But I just want to make 23 sure I understand the bullet. The bullet is I have a 24 60 percent reduction in depth of penetration versus 25 what you see in the literature of 30 to 40 percent?
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118 1 That's what I hear you saying.
2 MR. JAIN: The concern we have --
3 MEMBER CORRADINI: If I'm one-third of the 4 depth, that's a 60 percent reduction, a 66 percent 5 reduction, versus what we see in the literature of 30 6 to 40 percent?
7 MR. JAIN: Well, it's not 66 percent.
8 It's a lot more. For a rigid missile, the penetration 9 is almost 60 inches. And for a deformable missile, 10 it's 20 inches. It's one-third.
11 MEMBER CORRADINI: Right. I understand.
12 Okay. What you use as the denominator is not what I 13 would use as the denominator.
14 MR. JAIN: Oh, okay. All right.
15 MEMBER CORRADINI: That's all I'm trying 16 to get at.
17 MEMBER BLEY: I'm just a little curious 18 because I haven't thought about this, either. From 19 what they told us, they assumed there is no 20 reinforcing steel. So, they're looking at just a 21 concrete wall being hit by a solid steel missile.
22 MR. JAIN: Uh-hum.
23 MEMBER BLEY: And I don't have a clue 24 about how deformable a solid steel missile is running 25 into concrete with no reinforcing that.
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119 1 MR. JAIN: Well, that's what we want to 2 understand because rigid and deformable is a relative 3 term compared to the barrier.
4 MEMBER BLEY: So, there's got to be some 5 good basis for it.
6 MR. JAIN: Right. When a missile becomes 7 rigid compared to the barrier, that's what we intend 8 to look at here.
9 MEMBER BLEY: I look forward to seeing 10 what you come up with.
11 MR. JAIN: Because that's where we are at.
12 That gives you a snapshot of some of the design issues 13 we are looking at. So, we will review this design 14 information and conduct audits, and whatever we need 15 to, to make safety findings.
16 MEMBER BLEY: Okay. So, we won't see the 17 results of your work until sometime later when you 18 come back with an SER with no open items, I guess?
19 MR. JAIN: That is correct.
20 MEMBER BLEY: Okay.
21 DR. SCHULTZ: The same as leak-before-22 break evaluation.
23 MEMBER BLEY: Yes.
24 MR. JAIN: And now, I will pass it on to 25 Sunwoo for Section 3.7.
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120 1 MR. PARK: Okay. Good morning.
2 My name is Sunwoo Park, Structural 3 Engineer at NRO. I'm going to cover Sections 3.7.2 4 and 3.7.3 of the staff's SER with the open items.
5 Section 3.7.2 addresses the Seismic System 6 Analysis, which covers two Seismic Category I 7 structures, the reactor building and the control 8 building and, also, the rad waste building, which is 9 a Category II, is also covered, but only from the 10 perspective of a Seismic II over I and the structure-11 soil structure interaction on the perspective.
12 Section 3.7.3 addresses the Seismic System 13 Analysis, covering, for example, the bioshield and the 14 other subsystems, such as the NuScale Power Module, 15 the reactor building crane are covered in other 16 sections of the staff's SER.
17 The Phase 2 SER had a four open items in 18 this area, and two of them have been resolved to date 19 with the excellent effort and the cooperation of the 20 Applicant. But, still, the two, the other open items, 21 one in 3.7.2 and the other in 3.7.3, still remain open 22 as of today.
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121 1 discuss more details of these two open items.
2 The only open item in Section 3.7.2 3 concerns seismic load events for NuScale Power Module 4 and reactor building interface supports which are 5 determined by analysis of the reactor building SASSI 6 model and the NPM ANSYS model. Here SASSI and ANSYS 7 refers to specific computer programs that are used in 8 the seismic analysis of a reactor building and the 9 NPM.
10 Then, the Applicant has expanded analysis 11 cases to include 130 percent of NPM nominal stiffness 12 as a way of accounting for the potential stiffness 13 variation of the NPM system in a seismic analysis that 14 involves the NPM.
15 The Applicant also has adopted a new 16 methodology for modeling hydrodynamic mass of pool 17 water, with a purpose of incorporating pool water mass 18 into the analysis model more realistically.
19 The Applicant has provided information 20 from a new set of analysis which is currently in staff 21 evaluation.
22 And then, the last bullet needs to be 23 updated because we have received the additional 24 information recently which is about seismic loads on 25 the pool walls from the analysis, which is, again, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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122 1 currently in staff evaluation.
2 Next slide.
3 The open item in Section 3.7.3 is about 4 the bioshields which are Seismic Category II, concrete 5 covers placed on top of each NPM as additional 6 radiological barrier. A bioshield is removed during 7 NPM refueling activities, and the removed bioshield is 8 stacked on top of an adjacent bioshield. During the 9 December 2018 regulatory audit, staff identified 10 issues concerning seismic design of with respect to 11 the bioshield. And Applicant, subsequently, has 12 revised the design approach to address identified 13 issues and, also, they provided information on new 14 seismic analysis and the design of bioshields, which 15 is currently in staff evaluation.
16 Next slide.
17 Okay. In summary, Applicant has been 18 undertaking actions for timely resolution of the open 19 items. And with the exception of two open item, staff 20 has found that the NuScale DCA demonstrates compliance 21 with the applicable regulatory requirements for a 22 seismic system and the seismic subsystem analysis.
23 That is about it for my part. Any 24 questions?
25 Thank you.
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123 1 MR. ROCHE-RIVERA: Thank you, Sunwoo.
2 Good morning. My name is Robert Roche-3 Rivera. I'm a Structural Engineer in the Office of 4 New Reactors, and I will be presenting the staff's 5 review of DCA 3.8.4.
6 Next slide.
7 Section 3.8.4 focuses on Seismic Category 8 I structures other than containment, which includes 9 the reactor building and the control building. The 10 staff's review scope includes Section 3.8.4 of the DCA 11 and the associated Appendix 3B.
12 The staff performed its review in 13 accordance with DSRS Section 3.8.4. We held biweekly 14 public meetings with the Applicant to discuss 15 technical issues and resolutions to RAIs. We 16 conducted an audit of the design report supporting the 17 information submitted in the DCA and RAI responses.
18 And Phase 2 SER identified five open 19 items. All since that time of submission of the Phase 20 2 SER have been resolved. And we will be presenting 21 the resolution of the representative open items in 22 later slides.
23 Next slide, please.
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124 1 to the applicable Code and standards, acceptance 2 criteria, and allowable. Based on our review, we 3 found the loads and load combinations considered and 4 the structural capacity determinations to be in 5 accordance with the applicable Code and/or standard.
6 We found the displacement and results with the 7 applicable Code and/or standard allowable, and we 8 found that the structural capacities are greater than 9 the design basis demands. Based on our review, we 10 concluded that the Applicant's methods for 11 demonstrating the design adequacy of the structures 12 are consistent with the NRC's regulatory requirements.
13 Next slide.
14 This is a representative open item. For 15 this section, it's open item No. 3.8.4-1, related with 16 RAI 8171, Question 3.8.4-13. Upon review of Rev 0 of 17 the application, we did not find an evaluation, a 18 design evaluation of temperature demands and/or 19 pressure demands. And we issued an RAI requesting 20 such evaluation. In response to our question, the 21 Applicant performed the design evaluations with 22 consideration of temperature, both operating and 23 accident temperature, and accident pressure demands 24 for the reactor building; and also provided results, 25 including strength results for concrete rebar and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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125 1 liners, steel liner. And all those results were shown 2 to be below the allowable limits from the applicable 3 Codes and standards.
4 This concludes my presentation.
5 MS. VERA: Any questions for the 6 structural?
7 MEMBER CORRADINI: Members, any questions?
8 No. Keep on going. Next group.
9 MS. VERA: Okay. So, for the mechanical 10 part, we're going to start with Nick Hansing for 11 Section 3.9.4. "Control Rod Drive Systems".
12 MR. HANSING: Good morning. My name is 13 Nick Hansing. I'm the leader here for Section 3.9.4.
14 I'd like to begin by going over some of the key design 15 considerations and features that made the staff's 16 review.
17 On the NuScale design, the pressure 18 housing and electromagnetic components are very 19 similar to the existing fleet, in that the pressure 20 housing is designed and constructed to ASME Boiler and 21 Pressure Vessel Code Class 1 requirements.
22 There is a long drive shaft and remote 23 disconnect mechanism which are unique for this design.
24 The design standards and testing programs were 25 emphasized in this review.
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126 1 To talk about the drive shaft first, 2 NuScale considers the drive shaft as an internal 3 structure and as designated Seismic Category I. DCA 4 stipulates additional design requirements for this SSC 5 as well. An example of one of these additional design 6 requirements would be some of the service-level 7 loading combinations and material testing.
8 For the remote disconnect mechanism, the 9 other unique feature, the Applicant confirmed that the 10 remote disconnect coil is always de-energized during 11 normal operations and remains in this state during a 12 reactor trip. This mechanism was tested during the 13 key feature mockup testing for its full design life of 14 150 cycles, which is five times the estimated cycles 15 expected, with satisfactory performance.
16 Next slide, please.
17 Drop testing, which was mentioned earlier 18 in the Applicant's presentation. The staff reviewed 19 this testing. It was Appendix-B-compliant with 20 prototypical components. The staff independently 21 verified the dimensions of the design documents, the 22 as-built dimensions of the test facility. Important 23 dimensions like the diametrical gap were consistent.
24 So, it was a very good representation of the actual 25 facility.
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127 1 Up to 1 inch of misalignment was possible 2 at each interface, and the testing considered 3 manufacturing tolerances and seismic displacements in 4 the calculation of the amount of misalignment.
5 CO-CHAIR MARCH-LEUBA: I assume you're 6 talking misalignment in the horizontal direction?
7 MR. HANSING: Correct.
8 CO-CHAIR MARCH-LEUBA: Did you test 9 anything with bending when it is sideways and 10 diagonal?
11 MR. HANSING: I believe that the testing 12 facility did have mid-span deflection of the fuel as 13 one of the attributes that was controlled in the --
14 CO-CHAIR MARCH-LEUBA: The concern is that 15 this system is the only system that can shut down the 16 reactor. I mean, it has to work extremely reliably.
17 For example, some fuels have spacers that they're 18 designed to collapse when you have horizontal loads.
19 And then, the control rods still have to come in. And 20 it's very long.
21 MR. HANSING: Yes.
22 CO-CHAIR MARCH-LEUBA: And it has to work.
23 MR. HANSING: Yes, it does.
24 CO-CHAIR MARCH-LEUBA: It does not have a 25 backup. So, you did consider all those, right? Can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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128 1 I sleep comfortably tonight that you looked at it?
2 MR. HANSING: Yes, I believe that the 3 misalignment testing was very thorough and did look at 4 these aspects.
5 So, with this testing, 14 different 6 configurations were used. There was a number of 7 variables reduced using fabrication constraints and 8 testing worst-case scenarios. If the interface points 9 were within the same weldment, they would probably 10 move very similar to each other. So, that could be a 11 reduced number of variables to look at.
12 And in this case, the tested displacement 13 exceeded the maximum expected displacement by more 14 than a factor of two. So, there was sufficient margin 15 in their testing.
16 The most limiting drop, the maximum 17 displacement and longest drop time was bounded by the 18 performance assumed in the Safety Analysis for control 19 rod drop time.
20 Additionally, the operability assurance 21 program, which was also mentioned earlier, included 22 performance testing, stability testing, endurance 23 testing, and production testing. And this will be 24 completed by a COL applicant.
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129 1 of the program in the DC application with a proposed 2 COL item to implement the program and provide a 3 summary of the testing program and results for staff 4 review.
5 DR. SCHULTZ: What are the components of 6 the production testing? What does that refer to?
7 MR. HANSING: The production testing 8 includes some of the ASME Code testing, the 9 hydrostatic tests, and I believe there's some material 10 testing associated with that as well.
11 DR. SCHULTZ: And going back to your 12 earlier discussion -- no, on this slide, on slide 13 27 -- you've got 14 configurations, maximum expected 14 displacement by more than a factor of two. What does 15 that refer to? The factor of two applied to what?
16 Not the 1-inch misalignment?
17 MR. HANSING: So, they had calculated the 18 maximum expected displacement between the 19 manufacturing tolerancing stackup and the maximum 20 expected seismic displacement from their calculations, 21 and the value that they used was more than two beyond 22 what the summation of those factors would be.
23 DR. SCHULTZ: All right. Thank you.
24 MR. HANSING: Thank you.
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130 1 Tom Scarbrough. I'm responsible for Section 3.9.6, 2 which is the functional design and qualification and 3 preservice testing and inservice testing of pumps and 4 valves.
5 Section 3.9.6, we reviewed their function 6 design qualification and preservice testing, PST, and 7 inservice testing, IST, for the NuScale safety-related 8 valves. The DCA provides a full description of the 9 PST-IST programs with a few SER confirmatory items 10 that are relatively minor. There are some table 11 adjustments that need to be made, and things of that 12 nature, and some clarifications of the section. But, 13 overall, it's pretty close to being a full 14 description. So, we're real close there.
15 There are first-of-a-kind emergency core 16 cooling system valves, and there are first-of-a-kind 17 containment isolation valves. We'll be looking at 18 those and we're going to be talking quite a bit more 19 about those during the closed session this afternoon.
20 They are currently conducting ECCS valve 21 design demonstration testing at Target Rock right now.
22 And that's ongoing, and we'll give you some more 23 details about that this afternoon as well.
24 The SER itself has two open items in this 25 section. There's the ECCS valve design, of course, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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131 1 that we're still currently reviewing, and then, also, 2 the ITAAC. And there's a specific ITAAC, which is the 3 valve installation verification ITAAC. As you know, 4 there's several of these valve systems that have long 5 hydraulic runs associated with them, and the proper 6 installation of those valves is crucial to ensuring 7 that these valves, all the various valves, the ECCS 8 valves, the CIVs, the VHRS actuation valves, work 9 properly, because of those long hydraulic runs.
10 So, we're working with NuScale on that.
11 They provided us a proposal. We had some suggestions 12 to maybe streamline it with some more of the details 13 in Tier 2. And they're working on that now. So, we 14 think that's very close to being resolved as well.
15 Okay. So, next slide, please. Is it 16 going to change? There we go.
17 Okay. So, this is really high-level --
18 MEMBER BLEY: I'm sorry.
19 MR. SCARBROUGH: Yes, sure, go ahead.
20 MEMBER BLEY: Back to the ECCS valve 21 design demonstration testing --
22 MR. SCARBROUGH: Uh-hum.
23 MEMBER BLEY: -- do you have the test 24 plan? How extensive is that test plan?
25 MR. SCARBROUGH: Oh, it's very extensive.
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132 1 We're going to talk about that quite a bit this 2 afternoon.
3 MEMBER BLEY: We'll do that in closed 4 session?
5 MR. SCARBROUGH: Yes. I talked to NuScale 6 yesterday. They showed me their slides. They lay out 7 that whole test plan in their slides --
8 MEMBER BLEY: Okay.
9 MR. SCARBROUGH: -- which is great.
10 MEMBER CORRADINI: More to come.
11 MR. SCARBROUGH: More to come, right.
12 Right.
13 MEMBER CORRADINI: As you can tell, we're 14 a bit interested.
15 MR. SCARBROUGH: We are, too.
16 (Laughter.)
17 MEMBER CORRADINI: Okay.
18 MR. SCARBROUGH: Okay. So, as you know, 19 there's the reactor vent valves. There's three of 20 those, and they're on top of the reactor vessel. And 21 then, there's two reactor recirculation valves, RRVs, 22 which are on the side.
23 MEMBER CORRADINI: Can you hold on now?
24 MR. SCARBROUGH: Yes.
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133 1 somebody's online in the closed -- in the open line, 2 but you need to mute your phone because we hear 3 clicking.
4 Thank you.
5 MR. SCARBROUGH: Thank you.
6 MEMBER CORRADINI: Tom?
7 MR. SCARBROUGH: Thank you.
8 Okay. And we'll get more into this, the 9 precise operation of these various design arrangement 10 valves. But there's a main valve. There's an 11 inadvertent actuation block -- it's called an IAB 12 valve. And then, there's a solenoid trip valve, and 13 there's a solenoid reset valve. And they're all 14 connected by this long hydraulic tubing, and we're 15 going to get into that this afternoon.
16 The valve design demonstration testing is 17 being conducted to demonstrate that 50.43(e) is 18 satisfied. That has to do with new design features, 19 safety features in passive plants. And so, NuScale is 20 conducting that, and we'll talk more about that this 21 afternoon. But we've been following it very closely.
22 We were just there last week and we may be there 23 again. So, that's where we are with that.
24 MEMBER BLEY: This has a close link to 25 Chapter 6 --
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134 1 MR. SCARBROUGH: Yes, it does.
2 MEMBER BLEY: -- which we'll do later.
3 MR. SCARBROUGH: Yes.
4 MEMBER BLEY: Will you be back for that 5 one?
6 MR. SCARBROUGH: Yes, I will.
7 MEMBER BLEY: Okay.
8 MR. SCARBROUGH: Yes. And some of the 9 Chapter 6 people are helping us with this as well.
10 And so, we're working together. It's a team effort.
11 Okay. So, function design qualification.
12 In the big picture, the DCA specifies ASME Standard 13 QME-1-2007 for the qualification of all the safety-14 related valves, as accepted in Reg Guide 1.100, 15 Revision 3. So, we're comfortable with that. There's 16 a lot more testing going on besides what's happening 17 this month. You know, there's full qualification 18 testing that has to be conducted under QME-1. So, 19 we're comfortable with that.
20 Also, the ITAAC, there's a functional 21 qualification ITAAC for safety-related valves which 22 points to the qualification report that's specified in 23 the QME-1 standard. So, there's a tie between the 24 functional qualification ITAAC and QME-1 itself. So, 25 we're comfortable with that as well.
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135 1 We've conducted audits of the valve design 2 specifications. We have a few open items, and most of 3 them had to do with there were some portions of the 4 design specifications that were not quite complete 5 yet. There were some open items in the 6 specifications. And we wanted to make sure the 7 specifications were complete, that we were seeing a 8 full-fledged version of those specifications. And 9 we've heard that NuScale has completed that. So, 10 we'll conduct an audit sometime to follow that up and 11 make sure that was all, that they were all completed.
12 There are no safety-related motor-operated 13 valves. So, none of those for me to be worried about 14 here. There's no safety-related pumps and no safety-15 related snubbers. So, just valves in this reactor.
16 Okay. The next slide.
17 Okay. Containment isolation valves.
18 These are interesting valves. Basically, they're 19 hydraulic-operated and have ball valves in them.
20 Okay? So, we're using the containment isolation for 21 ball valves.
22 MEMBER CORRADINI: Is that unusual?
23 MR. SCARBROUGH: Yes, usually they're 24 sealed.
25 MEMBER CORRADINI: They're gate valves?
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136 1 MR. SCARBROUGH: You know, the gate 2 valves. They seal very strongly and they're held 3 closed by the neck-gear mechanism, right? So, that's 4 something we've talked to them about in our audits, is 5 what type of ball valves. There is a new valve 6 design. For butterfly valves, you can have a triple 7 offset which you can actually torque into it, because 8 butterfly valves typically aren't very good for 9 containment isolation, either. But if you have a 10 triple offset where the disc plane is slightly off-11 center from the piping angle, then you can actually 12 torque these into the seat. And that's what we've 13 been talking about; maybe these ball valves may be the 14 same design as that, so they can torque them into the 15 seat.
16 MEMBER BLEY: The same valves they use for 17 the main steam isolation valves?
18 MR. SCARBROUGH: Yes, yes. These are, 19 yes, these are all the same type. Now, actually, how 20 they're set up is a little different, and I'll get to 21 that. But there's 16 of the primary system 22 containment isolation valves, PSCIVs, and they have 23 one valve body, but two valves, two ball valves in the 24 same valve body with two separate, completely separate 25 actuators that go out to separate skids. And so, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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137 1 they're completely separate that way.
2 But they're closed by hydraulic pressure 3 and they have a hydraulic fluid. But they're opened 4 by that, right, but to close, they have nitrogen 5 bottles that close them, right? And we can talk more 6 about the actual specifics this afternoon, maybe 7 during the closed session, but I don't get too far 8 into the proprietary portion.
9 MEMBER BLEY: If I'm moving to the closed 10 session, tell me to wait.
11 MR. SCARBROUGH: Okay.
12 MEMBER BLEY: But I have two questions 13 about this one. So, usually, a ball valve has equal 14 forces on both sides, so that there's no torque from 15 flow if they're partially open. But if they're 16 offset, as you were saying, the flow can actually 17 assist in closing them? Is that --
18 MR. SCARBROUGH: Well, these actually, 19 you're able to torque them into the seat. And one 20 thing we've talked about, it's a rack-and-pinion 21 design that's going to be operated.
22 MEMBER BLEY: Yes.
23 MR. SCARBROUGH: And the question is, are 24 you going to size the rack-and-pinion such that it's 25 a locking gear train, right, or are you going to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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138 1 torque them into the seat sufficiently, that that 2 force trying to open them cannot do it? And actually, 3 we've talked about that, and that's something they're 4 still thinking about, what the detailed design is 5 going to be.
6 MEMBER BLEY: You just added another 7 question for me.
8 (Laughter.)
9 MR. SCARBROUGH: Okay.
10 MEMBER BLEY: So, this afternoon will you 11 have pictures of the details of these? Are these 12 unique valves or have they been used in other 13 industries for a long time?
14 MR. SCARBROUGH: Ball valves are used 15 quite a bit.
16 MEMBER BLEY: I know, but --
17 MR. SCARBROUGH: But this use is unique.
18 MEMBER BLEY: Okay.
19 MR. SCARBROUGH: We can see if we can 20 track some down for you. We had a lot of ECCS valves 21 discussion, but we didn't have any pictures of the 22 ball valves.
23 MEMBER BLEY: Okay.
24 MR. SCARBROUGH: We'll see if we can track 25 some down. Okay?
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139 1 MEMBER BLEY: We're actually going out 2 there next month.
3 MR. SCARBROUGH: That's great. I think 4 that's really good. It's really good to see how it 5 all works.
6 So, that's the primary system containment 7 isolation valves. Now there are six secondary system 8 containment isolation valves, and these are the 9 feedwater and the main steam isolation valves. And 10 the feedwater, there's two of those, and 4 NPS, and 11 then, there's four main steam isolation valves and 12 those are the 12-inch NPS.
13 And the feedwater valves also have a 14 nozzle check valve. They only have one valve, one 15 ball valve, but, then, they have a nozzle check valve 16 also after that. And nozzle check valves are a little 17 unique. We're starting to see some of them like in 18 AP1000 and such, but nozzles --
19 MEMBER BLEY: I've never -- I'm not 20 familiar with those at all.
21 MR. SCARBROUGH: Oh, okay. They're, 22 rather than having swinging, they actually have a --
23 it's almost like a --
24 MEMBER CORRADINI: Plugs.
25 MR. SCARBROUGH: A plug, it's like a plug.
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140 1 But it's streamlined internal. And it's sort of 2 streamlined internal to the valve, and it moves back 3 and forth. And sometimes there's a small spring in 4 there to keep it in the right position. And under 5 AP1000, we had them do some testing. There were some 6 testing done at, I think, Utah State out there with 7 them, so to qualify those for AP1000. So, we have 8 some experience with them.
9 MEMBER BLEY: I guess I'm familiar with 10 those in very small-diameter, low-flow kind of 11 situations. Is this again a unique use?
12 MR. SCARBROUGH: Right, using this large 13 -- they -- typically the operating plants don't -- you 14 don't see these in this large size. I have heard 15 they've been using some non-safety-related systems and 16 things, but that's why we are so interested in what 17 the AP1000 was. It was sort of unique. They're sort 18 of in the whole passive core cooling system for the 19 AP1000.
20 So we wanted to make sure we had a good 21 understanding of how they operated and whether they're 22 qualified. So they also will be qualified per QME-1 23 as accepted in Reg Guide 1.100. We did conduct an 24 audit of the CIVs and the reactor safety valves, and 25 there are a few closeout items there. The same thing.
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141 1 There are some specs. There were some open items in 2 the specs, and they're working on those to get those 3 closed out. So we'll -- as soon as they let us know, 4 we'll finish that up. So that's where we are with 5 those.
6 And that's the end of my planned 7 presentation for the open session. We have more for 8 you during the closed session. Is there any high-9 level questions that we could answer for you on this 10 right now?
11 (No audible response.)
12 MR. SCARBROUGH: No? Okay. And Renee Li 13 is going to take over for Section 3.6.2.
14 MS. LI: I'm Renee Li and I review Section 15 3.6.2, and the review is to ensure that the NuScale 16 designs provide adequate protection against effects of 17 postulate pipe rupture and make the GDC for --
18 The review concentrates on those area that 19 were outside the staff guidance of BTP 3-4. That 20 include two isolation valve outside containment for 21 penetration piping and also the voltage connection of 22 the ECCS valve to the reactor vessel nozzle. I also 23 review a technical report that address applicant's 24 pipe rupture analysis and associate results. And we 25 have one SER open item related to the bolted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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142 1 connection of the ECCS valve to the reactor vessel.
2 So upon mentioning that the NuScale design 3 provide two isolation valve in a single valve body 4 outside the containment and the single valve body is 5 welded to the containment vessel nozzle safe end. And 6 those NuScale's isolation valve configuration deviate 7 from the BTP 3-4 containment penetration area. The 8 specific -- the accessibility of this tube of outside 9 containment, rather than one in/one out would be 10 provide in Chapter 6 presentation. And in my 11 presentation I focus on the breaks exclusion for those 12 containment configuration.
13 So as you may know, in Branch Technical 14 Position 3-4 it defines the containment penetration 15 from the inboard containment isolation valve through 16 the penetration to the outboard containment isolation 17 valve. And that also the BTP 3-4 provides staff's 18 guideline that including certain design stress and the 19 fatigue limit and augment inspection to ensure the 20 extremely low probability of pipe failure in this area 21 and usually refer as spray exclusion area.
22 Since the NuScale configuration is beyond 23 the BTP 3-4 guideline, the staff ask a question for 24 NuScale to justify the application of break exclusion 25 in the area identify in the FSAR. In its RAI response NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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143 1 the applicant state that the design stress and the 2 fatigue limit criteria for the piping segment within 3 the NuScale specific break exclusion area are 4 consistent with the Branch Technical Position 3-4 5 guideline, and they also in the technical report I 6 mention earlier. It shows the results of their 7 calculation. And the results are within the relevant 8 BTP 3-4 stress and the fatigue limit for break 9 exclusion.
10 In addition, augment 100 percent biometric 11 in-service examination requirement for all the welds 12 within the break exclusion area is consistent with 13 Branch Technical Position staff guideline, and detail 14 of the ISI described in FSAR Section 6.2 and 6.6.
15 So with this the NRC staff found the applicant 16 justification for its of breaks exclusion area 17 acceptable.
18 Next slide, please? So the next area is 19 related to the break exclusion of RVV and RRV, the 20 bolted connection to the reactor vessel. As Tom 21 mention earlier, that the RVV and the RRV allowed the 22 natural circulation of the emergency core cooling, and 23 each of those RVV and RRV is bolted directly to the 24 reactor vessel nozzle. The NRC staff's key concern is 25 that this bolted flange connection must not fail to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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144 1 cause a LOCA. So the staff issue RAI requesting 2 NuScale to provide a justification for why the 3 probability of gross rupture is extremely low if 4 NuScale desire to treat the bolted connection as a 5 break exclusion area.
6 The applicant response the RAI and 7 codified that for both bolted connection, the NuScale 8 design classified them as break exclusion area and 9 they also provide the following justification: First, 10 the bolting stress design criteria per ASME Section 11 III NB 3230 meet the intent of BTP 3-4 stress 12 acceptance guidance for typical piping system.
13 Second, they also demonstrate why the cumulative usage 14 factor of 1.0 is acceptable.
15 For that first is the ASME Section III NB 16 3230.3(c) for the fatigue evaluation require a fatigue 17 strength reduction factor of not less than four being 18 apply. So that fatigue strength reduction would 19 provide a conservative safety margin for the 40 20 design. And also they identify that a list of the 21 phenomena which would -- which might adversely affect 22 fatigue evaluation and then went through each 23 phenomenon to demonstrate why those phenomena are not 24 susceptible, why the NuScale bolt connection are not 25 susceptible for those phenomena.
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145 1 Here I'd like to point out the BTP 3-4, 2 the fatigue usage factor criteria of 0.1, you know, 3 some technical -- I'd like to provide some technical 4 rationale. So in BTP 3-4 the postulate break is 5 really a mechanistic approach and based on the ASME 6 design requirement that including the stress and the 7 fatigue usage factor. So for the stress is 80 percent 8 of the ASME design stress allowable. And for the 9 fatigue usage factor the design is 1.0, but in BTP 3-4 10 is -- use 0.1 as threshold and if stress fatigue is 11 considered, then it would be a COF equal to 0.4.
12 So you see a -- I would say a larger 13 margin that -- for the COF. The basis for the 0.1 is 14 to ensure a conservative margin on the cycle is made 15 available to take into account for example potential 16 40 design, improperly control fabrication, or 17 installation error, an expect mode of operation, and 18 certainty in the vibration load and other degradation 19 mechanism such as corrosive environment, water hammer.
20 So those are the phenomenon that I mention early that 21 NuScale provide a justification of why those phenomena 22 are not applicable to this particular RVV, RRV bolt 23 connection.
24 Next slide. So --
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146 1 have a few thoughts on that. The fatigue strength, 2 some of them were positive; some of them were 3 negative. The fatigue strength reduction factor for 4 bolting is in there because you have the threads and 5 the threads give you a high stress concentration 6 factor. So it's -- as opposed to a welded joint. But 7 that's kind of a -- so you can't take the full credit 8 for the fatigue strength reduction factor of four as 9 providing margin because you got the stress raiser.
10 On the other hand, isn't -- in my mind 11 there's just some natural redundancy in a bolt that --
12 in a bolted flange connection compared to a welded 13 connection. In other words, to get a rupture you have 14 to fail essentially all the bolts. I mean, it's --
15 you're more likely getting a leak, but I think you're 16 less likely to get an actual rupture because you'll --
17 if you can -- if it starts leaking, you detect a leak 18 and then you do something about it --
19 MS. LI: Yes.
20 CO-CHAIR RICCARDELLA: -- if you fail one 21 or two bolts, right?
22 MS. LI: I agree. Actually I cover that 23 because to justify the COF of 1.0 --
24 CO-CHAIR RICCARDELLA: Yes.
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147 1 already cover --
2 CO-CHAIR RICCARDELLA: Yes.
3 MS. LI: -- the next slide is also NuScale 4 design employ augment fabrication and the in-service 5 exemption requirement. The in-service inspection, the 6 UT exam for the bolt would be at least once every 10 7 years.
8 CO-CHAIR RICCARDELLA: Okay.
9 MS. LI: And also the bolting design adapt 10 NUREG-1389 guideline with highly SS -- SCC-resistant 11 material. That's the alloy-718. And lastly, the 12 highly-sensitive leakage detection system that 13 sensitive to leak rate as low as 0.001 gallon per 14 minute.
15 So with all the five justification the NRC 16 staff found that applicant justification provide a 17 reasonable assurance of the -- for break exclusion.
18 MEMBER CORRADINI: But just so I'm on the 19 same page, I see what you're saying. I'm fine with 20 that. But on the other hand, the RRV is assumed to 21 open up because from the standpoint of the transient 22 accident analysis this is one of their breaks that 23 they consider anyway.
24 MS. LI: Yes, I think Chapter 6 may cover 25 that.
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148 1 MEMBER CORRADINI: This maximizes pressure 2 in containment.
3 MS. LI: Right, but I think in here the 4 focal point is not to break the --
5 MEMBER CORRADINI: I'm with you. Okay.
6 MS. LI: Okay. So with that, NuScale has 7 not yet complete the stress and the fatigue 8 calculation for this particular bolt connection, and 9 we going to audit that when the information become 10 available and to close this open item. So that --
11 DR. SCHULTZ: Did they have a complete --
12 in terms of the augmented fabrication examination 13 requirements has that been completed? Have they 14 described that completely to your satisfaction, the 15 fabrication examination requirements?
16 MS. LI: Right. Yes.
17 DR. SCHULTZ: Okay. Thank you.
18 MS. LI: So that conclude my presentation.
19 MEMBER CORRADINI: Thank you.
20 MR. TSIRIGOTIS: Good morning. My name is 21 Alexander Tsirigotis and I'm a mechanical engineer in 22 the New Reactors Office. This will be a brief 23 presentation of the review the staff performed for the 24 NuScale design certification Sections 3.8.2, 3.9.3, 25 and 3.9.5.
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149 1 The staff performed a design specification 2 audit as part of its review. There the staff found 3 that the stress analysis and fatigue evaluations of 4 the reactor pressure vessel, containment vessel and 5 reactor internals were not available. The position in 6 SECY 90-377 is that this level of detail should be 7 completed and included at the design certification 8 stage. Because these evaluations were not available, 9 this remains an open item.
10 NuScale is updating the stress analysis 11 and performing fatigue evaluations at critical 12 locations in accordance with ASME Boiler and Pressure 13 Vessel Code Section III. The calculations for the --
14 to complete the stress analysis and the fatigue 15 evaluation are scheduled to be available for audit by 16 the end of July, therefore this remains an open item.
17 The staff also reviewed the design of the 18 trip/reset valve of the ECCS valve. You've heard 19 previously my colleagues discussing the ECCS valves 20 and you will hear more about this later.
21 This trip/reset valve serves as both RCS 22 and containment pressure boundary. The next slide 23 will show a schematic presentation of the boundaries 24 for this valve. According to the ASME Boiler and 25 Pressure Vessel Code Section III examination of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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150 1 valve body is required during fabrication. The NRC 2 staff questioned the adequate level of quality for the 3 design of this valve. The NRO office though finally 4 accepted the ASME Boiler and Pressure Vessel Code 5 Section III as the acceptable level of quality.
6 Information on the function, operation, 7 performance and testing of the ECCS valves will follow 8 in the closed session as Tom mentioned earlier.
9 And this sketch here shows the schematic 10 of the trip/reset solenoid valve and it shows the 11 boundaries, the pressure boundaries. Any questions?
12 CO-CHAIR MARCH-LEUBA: Yes, on the fatigue 13 analysis these valves are set on their operating 14 position and kept there for two years. They only move 15 once every two years during refueling. So what 16 fatigue analysis are you doing?
17 MR. TSIRIGOTIS: The fatigue analysis that 18 I mentioned earlier --
19 CO-CHAIR MARCH-LEUBA: Yes.
20 MR. TSIRIGOTIS: -- is for the containment 21 vessel, the reactor vessel --
22 CO-CHAIR MARCH-LEUBA: Oh, not for the 23 valves?
24 MR. TSIRIGOTIS: -- and the internals.
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151 1 somebody's going to explain to us how this valve 2 works, because I'm looking at the design that you --
3 MR. TSIRIGOTIS: In the closed session --
4 CO-CHAIR MARCH-LEUBA: -- and I don't see 5 nothing.
6 MR. TSIRIGOTIS: -- you will see more 7 about this.
8 CO-CHAIR RICCARDELLA: Would you put that 9 sketch back up, please?
10 MR. TSIRIGOTIS: Yes.
11 CO-CHAIR RICCARDELLA: So this is the 12 containment valve attached in the containment. So 13 then there has to be a break exclusion zone here, 14 right, where you can't assume a break?
15 MR. TSIRIGOTIS: Well, did you -- I don't 16 know if Renee spoke about the containment, about the 17 break exclusion area. I think she did. Yes.
18 CO-CHAIR RICCARDELLA: I just wanted to 19 confirm. In this sketch right here we're talking 20 about the area between the valve and the containment 21 as a break exclusion zone, correct? Renee?
22 MEMBER CORRADINI: We're confused.
23 MR. TSIRIGOTIS: I can see that. These 24 valves are very small to begin with, right? I mean, 25 as you see there the size is --
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152 1 CO-CHAIR RICCARDELLA: Yes. Okay.
2 MR. TSIRIGOTIS: Now I am not quite sure, 3 but I think the break exclusion area is for piping and 4 these valves have tubing, small tubing.
5 CO-CHAIR RICCARDELLA: Okay.
6 MEMBER CORRADINI: So --
7 MR. TSIRIGOTIS: To postulate breaks below 8 one inch, I believe that BTP 3-4 does not postulate 9 breaks below one inch.
10 MEMBER CORRADINI: Yes, help us out, Tom.
11 MR. SCARBROUGH: These -- those tubes are 12 all inside that valve, that body. Safe-in -- there's 13 a safe-in and those tubes run through there and then 14 they connect to the solenoid valves. The rest valve 15 is on the top and the trip valve is on the bottom, how 16 they're set up. And we'll show those in the closed 17 session --
18 MEMBER CORRADINI: Okay.
19 MR. SCARBROUGH: -- exactly how those are.
20 MEMBER CORRADINI: So you're going to 21 educate us in closed session?
22 MR. SCARBROUGH; With NuScale's help, yes.
23 Right.
24 CO-CHAIR RICCARDELLA: CORRADINI: Good.
25 Okay. Thank you.
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153 1 MR. SCARBROUGH: Yes.
2 MR. TSIRIGOTIS: Thank you, Tom.
3 MS. VERA: This concludes Chapter 3.
4 PARTICIPANT: It appears you're done.
5 MS. VERA: Yes.
6 PARTICIPANT: Thank you, gentlemen, 7 ladies.
8 MEMBER CORRADINI: Questions the members?
9 (No audible response.)
10 MEMBER CORRADINI: Okay. We are going to 11 postpone any closed session until the end of the day.
12 And we will take up Chapter 6 next, but we're close 13 enough to lunch I'll declare victory and we'll go to 14 lunch. Can we -- I assume we can start earlier.
15 PARTICIPANT: Yes.
16 MEMBER CORRADINI: So why don't we come 17 back at quarter to 1:00? We have an hour. Have a 18 good lunch.
19 (Whereupon, the above-entitled matter went 20 off the record at 11:48 a.m. and resumed at 12:49 21 p.m.)
22 CO-CHAIR MARCH-LEUBA: Rebecca, you're up.
23 MS. NORRIS: Good afternoon, everyone.
24 Thank you for having us. This presentation is the 25 NuScale presentation on FSAR Chapter 6, Engineered NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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154 1 Safety Functions. This is in support of the staff 2 discussion including open items.
3 The presentation team for today is Dan 4 Lassiter, a mechanical design engineer; Zack Houghton, 5 mechanical design engineering manager; and myself, 6 Rebecca Norris, the licensing supervisor for this 7 chapter.
8 Today we will be covering Chapter 6 as 9 shown with the following notes: As you can see, 6.7 10 does not apply to the NuScale design, so we will not 11 be covering that. Also 6.3, we have detailed valve 12 drawings and cutaways in the closed session later 13 today, so we're going to be giving a brief overview in 14 this open session then later providing more details.
15 We also will be giving an overview of 16 Section 6.2.1 and 6.2.2., but we will not be going 17 into the analysis portion. This is because we share 18 the RELAP analysis with Chapter 15. The Chapter 15 19 presentation is tomorrow, so it will be covered then.
20 Also as a final note, in the back of your 21 packet there are quite a few backup slides. We tried 22 to organize them in the same order as the main 23 presentation. We do not intend to cover them for time 24 constraints, but if you have any questions on them, 25 feel free to ask.
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155 1 And with that, I will turn it over to Zack 2 for the presentation.
3 MR. HOUGHTON: All right. Thank you, 4 everyone. Good afternoon.
5 So first we'll start with Section 6.1, 6 Engineered Safety Feature Materials. So components 7 for the engineered safety features have been selected 8 to be compatible with the environmental conditions of 9 normal operation, maintenance, testing and accidents.
10 Components for the NuScale design of the engineered 11 safety features have a 60-year design life. One 12 material choice of note that we put on this slide is 13 the use of a forged XM-19 material in the core 14 vicinity of the lower containment vessel.
15 MEMBER BALLINGER: Is this with -- well, 16 XM-19, Nitronic 50, whatever you want to call it, is 17 it -- is this section size bigger than usual for use 18 of this material? In other words, is this an unusual 19 size for XM-19?
20 MR. HOUGHTON: I believe it is larger than 21 previous experience, but I'll consult with our 22 materials expert H.Q. Xu, who should be on the line.
23 H.Q., do you have any more insight on that 24 question?
25 MR. XU: Yes, this is H.Q. Yes.
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156 1 MR. HOUGHTON: Yes is the answer.
2 MEMBER BALLINGER: I guess that's an 3 answer.
4 (Laughter.)
5 MR. HOUGHTON: And it was chosen 6 specifically because it has shown good resistance to 7 neutron embrittlement. So the lower -- the 8 containment valve being a metal vessel, I think the 9 members are familiar with it, and we'll see it more in 10 future slides. But that's why we made the specific 11 material choice there for the lower containment vessel 12 in the core region.
13 MEMBER BALLINGER: It's also got twice the 14 yield strength of regular stainless steel.
15 MR. HOUGHTON: And we do not allow any 16 protective coatings within the containment vessel or 17 for any components within the containment vessel.
18 MEMBER CORRADINI: So there's no organics 19 at all?
20 MR. HOUGHTON: Correct.
21 MEMBER CORRADINI: Okay.
22 CO-CHAIR MARCH-LEUBA: When you say on, 23 you mean inside, or outside, too?
24 MR. HOUGHTON: Inside of containment.
25 CO-CHAIR MARCH-LEUBA: It is more like in?
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157 1 MR. HOUGHTON: Yes.
2 All right. For Section 6.2.1 the primary 3 functions of the containment itself are to provide 4 enclosure and support of the reactor pressure vessel, 5 the reactor coolant system and the associated 6 components. It is a credited fission product release 7 barrier. It provides containment of postulated mass 8 and energy releases inside containment, so that would 9 be a steam or feed line break or a chemical volume 10 control system line break. It supports operation of 11 the emergency core cooling system by providing 12 retention of reactor coolants during ultimate heat 13 sink operation and also by providing a heat removal 14 pathway to the ultimate heat sink.
15 It's designed to support heat removal 16 capability to maintain pressure and temperature less 17 than the design allowables; that is in accordance with 18 GDC-50, and also to rapidly reduce the peak pressure 19 and temperature to less than 50 percent within 24 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. And as Rebecca mentioned earlier, the actual 21 details of the pressure temperature analysis will be 22 with Chapter 15 tomorrow.
23 All right. And again, hopefully the 24 members are familiar with the containment design at 25 this point. We can see the figure on the left of this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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158 1 slide here is the -- showing the outline of the 2 containment vessel and the reactor vessel associated 3 components located with that. The containment vessel 4 is designed and stamped as an ASME Class 1 vessel for 5 NuScale. It is 1,050-PSI design pressure, 550-degree 6 design temperature. It's located directly within the 7 ultimate heat sink, which is the reactor building 8 pool. And as mentioned, we use an XM-19 for the lower 9 containment material and an SA-508 Grade 3, Class 2 10 for the upper containments.
11 CO-CHAIR RICCARDELLA: So you're going 12 from a low-alloy steel to a stainless steel? You have 13 a bi-metallic weld there?
14 MEMBER BALLINGER: Just a flange.
15 MR. HOUGHTON: The difference is made at 16 the flange connection.
17 MEMBER BALLINGER: Oh, it's at the flange?
18 I see. Okay.
19 MEMBER CORRADINI: So where is the water 20 level?
21 MR. HOUGHTON: So the water level is 22 located --
23 MEMBER CORRADINI: The ultimate -- or 24 the --
25 (Simultaneous speaking.)
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159 1 MR. HOUGHTON: The water level is located 2 right below the head region of the containment vessel.
3 PARTICIPANT: That's the pool water level.
4 MR. HOUGHTON: The pool water level, 5 that's correct.
6 MEMBER CORRADINI: Okay.
7 MR. HOUGHTON: During normal operation the 8 containment is maintained at a vacuum, so there would 9 be no water inside of containment.
10 MEMBER CORRADINI: Have you guys decided 11 what you're going to define as vacuum? Has that been 12 decided, or is that private -- proprietary what you 13 choose finally as the value?
14 MR. HOUGHTON: Well, our tech specs 15 require that we operate sufficiently low to maintain 16 our containment evacuation system and containment 17 leakage detection operable. The exact value that we 18 would be operating at in normal operation will vary 19 cycle to cycle, plant to plant based on if you have 20 any in-leakage or out-leakage, of course, and the 21 actual performance of the containment evacuation pumps 22 themselves. So we don't have a set vacuum limit for 23 normal operation.
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160 1 suction?
2 MR. HOUGHTON: So it is taking its suction 3 at the head of the containment, in the upper region.
4 And so all of the piping and electrical penetrations 5 are located on the head of the containments. One of 6 the pipelines at the head of containment is connected 7 to a containment evacuation pump, which would be 8 located out in the reactor building gallery area.
9 MEMBER CORRADINI: So one of the members 10 at lunch after a refreshing meal at the cafeteria --
11 (Laughter.)
12 MEMBER CORRADINI: -- had a question about 13 dynamic effects about measuring leakage. So what --
14 you tell us when there's an appropriate time to ask 15 because I'm trying to understand where an accumulation 16 of condensate would be, how you then -- either by 17 measuring pressure or by measuring moisture condensate 18 from essentially pulling a vacuum how you determine 19 that and how the dynamics of it would be affected.
20 Because we couldn't answer it at lunch.
21 MR. HOUGHTON: Okay. I'll give a high-22 level overview and then please ask any clarifying 23 questions.
24 MEMBER CORRADINI: At the appropriate time 25 in your discussion, whenever it is.
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161 1 MR. HOUGHTON: This is probably as good a 2 time as any. I really don't have any slides that talk 3 specifically about the evacuation system. That would 4 have been presented with Chapter 9.
5 MEMBER CORRADINI: Yes, I know, but --
6 MR. HOUGHTON: So we have --
7 MEMBER CORRADINI: -- we need to be 8 reminded.
9 MR. HOUGHTON: Sure. So we have two 10 methods for detecting any leakage into containment.
11 There's the pressure of the containment itself and 12 then there is the condensate collection vessel in the 13 containment evacuation system, which is a very 14 sensitive method for seeing if we have any condensate 15 collected.
16 MEMBER CORRADINI: But the second method, 17 which is essentially measuring what you condense by 18 taking sampling, or by taking a continuous pool like 19 the containment, just have some time window at which 20 you do the averaging over. That time window can't be 21 as small as a minute. It strikes me that that's too 22 small.
23 So what is the time window of averaging?
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162 1 happened.
2 MR. HOUGHTON: I don't have a specific 3 answer for that on the time window, for the amount of 4 time. That's something I'd have to go discuss with 5 the team and come back with a --
6 MEMBER CORRADINI: Okay.
7 MR. HOUGHTON: -- better answer on. And 8 I don't think we have anybody on the line --
9 (Simultaneous speaking.)
10 MEMBER CORRADINI: Or we can put it down 11 on a list of things to chat with you about in July.
12 MR. HOUGHTON: Yes, that's --
13 MEMBER BALLINGER: There has to be a tech 14 spec on unidentified leakage.
15 MR. HOUGHTON: Correct.
16 MEMBER BALLINGER: So you have to meet 17 that somehow. And if each plant -- each module may be 18 slightly different, then that implies that the vacuum 19 -- the system is maybe slightly different. You have 20 to meet the tech spec on unidentified leakage.
21 MR. HOUGHTON: Correct.
22 MEMBER BALLINGER: Yes.
23 MR. HOUGHTON: Yes, and there are specific 24 timelines for the amount of time that is allowed to 25 identify a specific leakage amount. I don't recall NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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163 1 the exact numbers off the top of my head, but we're 2 well within the regulatory requirements for how 3 quickly we can detect leakage, our design being a 4 metal containment vessel it maintain the vacuums. We 5 have very sensitive measuring --
6 (Simultaneous speaking.)
7 MEMBER CORRADINI: Right, but the reason 8 the question got generated over lunch was because the 9 containment vessel has no insulation and it's always 10 condensing there's got to be -- you'd get a pulse --
11 if you had a leak, you'd see a pulse in pressure and 12 an equivalent increase in measure condensate, because 13 I assume the evacuation pumps are not changing flow 14 rate. They're pulling at some fixed rate. So I get 15 to a different quasi-equilibrium pressure which tells 16 me that somehow I have an additional leak that I can't 17 detect, and that is either above or below the tech 18 spec limit.
19 MR. HOUGHTON: Correct, and there's also 20 tech spec limits on being able to maintain the 21 containment evacuation system operable. And so to 22 support that there are graphs in Chapter 5 which 23 dictate the pressure versus temperature that you would 24 see inside containment. Basically we're trying to 25 avoid a scenario where we have condensate collecting NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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164 1 inside the containment because that then wouldn't make 2 it out to our collection vessel. So we preclude 3 operation within a window that would have condensate 4 on the containment vessel.
5 MEMBER CORRADINI: And that's in Chapter 6 6?
7 MR. HOUGHTON: The graph is in Chapter 5.
8 That's correct.
9 MEMBER CORRADINI: Okay. Thank you.
10 CO-CHAIR MARCH-LEUBA: Yes, see, now that 11 we've gone through that, I'm thinking that this leak 12 detection only works if the leak rate is larger than 13 the capacity of the vacuum to remove it. If you have 14 -- I always assume that the vacuum pumps pull vacuum 15 and then they stop.
16 MR. HOUGHTON: Yes.
17 CO-CHAIR MARCH-LEUBA: But if you have 18 been continuously running --
19 MR. HOUGHTON: The vacuum pumps are 20 continuously running, so if the leakage rate were 21 beyond the capability of the vacuum pumps, then you 22 would have a build up in pressure and then we would 23 eventually exceed our tech spec --
24 (Simultaneous speaking.)
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165 1 than the vacuum pumps, then you will not see it?
2 MR. HOUGHTON: Then we will see it. So 3 the containment evacuation system is always running.
4 Any vapor in the containment space would be pulled 5 out, condensate would be collected off of the --
6 CO-CHAIR MARCH-LEUBA: Oh, so condensate 7 is collected by the vacuum pumps, not at the bottom of 8 the containment?
9 MR. HOUGHTON: By the vacuum pumps.
10 That's correct. There's a -- excuse me, by the vacuum 11 system. So the vacuum pumps are constantly taking 12 suction. There's a condenser on that system.
13 CO-CHAIR MARCH-LEUBA: So the condenser is 14 downstream of the -- is on the high-pressure side of 15 the vacuum pump?
16 MR. HOUGHTON: Yes, and so we would 17 collect any condensate out of that. That's how we 18 measure the specific --
19 (Simultaneous speaking.)
20 CO-CHAIR MARCH-LEUBA: I misunderstood.
21 I thought you were condensing it and collecting it in 22 the bottom of the vessel.
23 MR. HOUGHTON: No. No, we preclude 24 condensing inside of the vessel with our operating 25 limits.
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166 1 CO-CHAIR MARCH-LEUBA: Oh, you hope you 2 don't.
3 MR. HOUGHTON: Yes.
4 CO-CHAIR MARCH-LEUBA: So as long as the 5 leak is smaller than the capacity of the vacuum pumps, 6 you will be okay if you continue to operate? You will 7 detect it and you will be able to continue to operate?
8 MR. HOUGHTON: Correct. And as long as 9 the leak is lower than our allowable leakages per our 10 tech specs.
11 CO-CHAIR MARCH-LEUBA: Yes. All right.
12 CO-CHAIR RICCARDELLA: I'm sorry. I had 13 something I wanted to ask about.
14 The selection of the materials for the 15 containment -- would you go back to the previous 16 slide? So you said you picked the FXM-19 because of 17 better radiation resistance, but yet the reactor 18 vessel, which is at a higher radiation field, is still 19 a low-alloy steel, right?
20 MEMBER BALLINGER: Yes, the reason I asked 21 the question, it's otherwise known as Nitronic 50.
22 That material is used in light water reactors in in-23 core or -- you know, inside the vessel for a lot of 24 things. So that's that.
25 It also has twice the yield strength.
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167 1 CO-CHAIR RICCARDELLA: I would assume 2 though that the radiation, the neutron embrittlement 3 concern is greater at the vessel? Higher neutron flux 4 than at the containment, right, down in that region?
5 MR. HOUGHTON:
6 MR. HOUGHTON: That's true. And so we do 7 all -- I mean, we take appropriate precautions for 8 dealing with neutron embrittlement at the reactor 9 vessel, but that uses the SA-508 material. We have 10 coupons in place in the Surveillance Program for 11 monitoring embrittlement.
12 CO-CHAIR RICCARDELLA: Well, I guess I was 13 just curious as why you felt it was necessary to go to 14 this other material here.
15 MR. HOUGHTON: It was a design decision 16 that we made for a number of reasons, one of which 17 though was questions on whether or not we would be 18 able to get appropriate coupon data from test coupons 19 located in the containment vessel to support that 20 containment vessel material.
21 CO-CHAIR RICCARDELLA: Okay.
22 MR. HOUGHTON: So it was trying to avoid 23 the concern of embrittlement of that vessel because of 24 its unique conditions.
25 CO-CHAIR RICCARDELLA: And are the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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168 1 radiation levels close? I mean, they are fairly close 2 together, so maybe they're not that big a difference.
3 PARTICIPANT: It's one over R.
4 MR. HOUGHTON: Yes, I don't have the exact 5 numbers for the difference between the two.
6 CO-CHAIR RICCARDELLA: And then on the 7 low-alloy steel on the containment is that clad on the 8 outside?
9 MR. HOUGHTON: Correct.
10 CO-CHAIR RICCARDELLA: Inside and outside 11 or just --
12 MR. HOUGHTON: Yes.
13 CO-CHAIR MARCH-LEUBA: Okay.
14 MR. HOUGHTON: It is clad on both the 15 inside and the outside.
16 CO-CHAIR RICCARDELLA: Okay. And the 17 reactor vessel is clad inside and outside, too?
18 MR. HOUGHTON: Correct.
19 CO-CHAIR RICCARDELLA: Okay. Thank you.
20 MEMBER SKILLMAN: Did the design of the 21 reflector have anything to do with choice of material 22 for the containment vessel? In other words, are you 23 depending on that reflector to reduce the neutron 24 fluence sufficiently to not have to consider NVT on 25 the containment vessel?
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169 1 MR. HOUGHTON: I don't think I'd be 2 prepared today to give a complete answer on design 3 decisions around the reflector and how that related 4 back to the containment vessel. Certainly we do 5 consider the reflector when we're accounting for 6 neutron embrittlement. Design decisions of the 7 reflector blocks, not something I could address 8 completely today. If you have more interest, it's 9 something we could discuss when the membership is out 10 in July.
11 MEMBER SKILLMAN: One more question: A 12 couple weeks ago, maybe a month or two ago, Dr.
13 Ballinger and Dr. March-Leuba and I were really kind 14 of on the warpath for safety level of boric acid 15 addition and CVCS, but as we've explored more 16 thoroughly the importance of leak detection, it's 17 almost as if the containment evacuation system may 18 have much more of a primary role in terms of alerting 19 the operators to an emerging issue in terms of 20 leakage.
21 So I understand you use the PRA to set the 22 safety requirements for the -- or to establish what 23 are the safety requirements for the system for what is 24 nuclear safety and what is not, but does this warrant 25 a more thorough examination of those decisions given NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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170 1 the importance of your CES, your evacuation system?
2 I mean, that's going to be the leading indicator of a 3 leak.
4 MR. HOUGHTON: I would certainly agree 5 with the importance of that system and, you know, you 6 look at our -- the design of our containment being a 7 smaller volume system maintained at a vacuum. I 8 believe the sensitivity of the system to leakage is 9 going to be a very important operating point for the 10 plant and something the operators will care greatly 11 about, so I fully agree with that statement. And it's 12 -- we have built in the appropriate redundancy into 13 the system to make sure that we can maintain that 14 equipment operational. It's -- the system is 15 classified the way it is because it is not credited 16 for dealing with any accident scenarios.
17 CO-CHAIR MARCH-LEUBA: Yes, but one poor 18 man's definition of safety grade is if it is used to 19 maintain the tech specs, it is safety grade, right or 20 not?
21 MR. HOUGHTON: I don't believe that's --
22 CO-CHAIR MARCH-LEUBA: Or it's only --
23 (Simultaneous speaking.)
24 MR. HOUGHTON: -- definition of safety-25 related.
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171 1 MEMBER SKILLMAN: It's the right feeling, 2 but it's not the right legal test.
3 MR. HOUGHTON: Okay. So we'll move onto 4 slide 8 if there are no further questions.
5 So we handed out a hard copy of this slide 6 as well because it's very difficult to read the 7 figure, but it's an important figure.
8 So this figure shows the class breaks per 9 Reg Guide 1.26 of the components that make up the 10 containment isolation system. So there's the 11 containment itself, but then these are the containment 12 isolation valves located on the head.
13 The areas show in blue crosshatching are 14 the ASME Class 1 isolations. And again, that's per 15 Reg Guide 1.26. And then in the green crosshatching 16 that shows the piping and the valves that are 17 designated as ASME Class 2. And I'll say -- I'll 18 point out as well downstream -- going back to the ASME 19 Class 1 isolations, downstream of the second outboard 20 isolation to the next isolation is designated as ASME 21 Class 3, and that's again in accordance with Reg Guide 22 1.26.
23 One point that will come up later in the 24 slides but I'll point out now, because the design of 25 the isolations which isolate lines that connect NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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172 1 directly the containment are similar to the design of 2 those valves which isolate the reactor coolant system, 3 those valves are all designed, constructed and stamped 4 in accordance with ASME Class 1. That doesn't show up 5 on this figure. I'll describe that in more detail 6 later, but it's essentially the valves that you see on 7 the right side of the screen. The main steam and the 8 feedwater valves, those are designed in accordance 9 with ASME Class 2.
10 CO-CHAIR RICCARDELLA: Help me understand.
11 Most existing reactors, PWRs and BWRs, when they have 12 double containment isolation valves, there's one 13 inside and there's one outside, right? And your 14 system is designed where they're both on the outside.
15 Help me understand the technical differences between 16 those two in terms of isolating the breaks and break 17 exclusion zones.
18 MR. HOUGHTON: We'll have a slide on 19 that --
20 CO-CHAIR RICCARDELLA: Okay.
21 MR. HOUGHTON: -- in a few more slides.
22 CO-CHAIR RICCARDELLA: All right. Thank 23 you. I'll wait.
24 MR. HOUGHTON: Okay. The COL item for 25 Section 6.2.1 is really related to the transient NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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173 1 analysis and nothing that we've discussed so far.
2 All right. Section 6.2.2, Containment 3 Heat Removal. Again, the containment itself is the 4 containment heat removal component for the NuScale 5 design, so we don't have a separate active system of 6 containment spray or ice condensers or any -- these 7 other designs for removing heat from the containment.
8 The designs have active equipment.
9 As mentioned, the pool water is just below 10 the upper head during normal operation, and that 11 provides greater than 30 days cooling volume.
12 The containment steel walls allow direct 13 passive heat transfer to the pool. For our limiting 14 peak pressure cases the containment pressure is 15 reduced to less than 50 percent of peak pressure 16 within two hours. We have requested exemption from 17 General Design Criteria 40 which requires testing of 18 containment heat removal systems. Because there are 19 no active components there's no real operability or 20 performance testing to be done for our containment 21 heat removal components. And the containment removes 22 heat from module as the accident progresses with no 23 operator action.
24 MEMBER CORRADINI: So there's no testing 25 of DHRS? Maybe I'm not following.
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174 1 MR. HOUGHTON: This is just the 2 containment heat removal, so this is the containment 3 itself.
4 MEMBER CORRADINI: Not the DHRS?
5 MR. HOUGHTON: This is not -- does not 6 include the DHRS.
7 MEMBER CORRADINI: Okay. Sorry.
8 MR. HOUGHTON: All right. So Section 9 6.2.4 on Containment Isolation. So the containment 10 isolation valves meet the General Design Criteria 55.
11 MEMBER CORRADINI: Can we go back?
12 MR. HOUGHTON: Yes.
13 MEMBER CORRADINI: So there's no testing 14 of it. What is the inspection process that's 15 necessary for -- because you rely on it for long-term 16 cooling, which means you've got to make sure the inner 17 surface is relatively continuously clean so that you 18 don't have to take account of some sort of following 19 factor of the inner surface, or the outer surface. So 20 what is done there in terms of inspection upon 21 refueling?
22 MR. HOUGHTON: So there are in-service 23 inspection requirements within the in-service 24 inspection, the ISI Program, for doing visual 25 examinations of containment surfaces during refueling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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175 1 outages.
2 MEMBER CORRADINI: Is there some limit as 3 to if you see fouling that you have to take -- go 4 inside and do cleaning?
5 MR. HOUGHTON: That would have to be in 6 the applicant's Inspection Program where they would 7 just make that determination on exactly when cleaning 8 will happen versus not. I'll say in our analyses we 9 do account for fouling. It's one of the factors that 10 we look at in our analyses.
11 MEMBER CORRADINI: But you don't account 12 for any sort of cleaning operation on an every-two-13 year basis? That's not accounted for? What I guess 14 I'm trying to get at is there will be an accumulation 15 of a fouling factor and you don't take account of the 16 fact that it needs to be cleaned off in terms of 17 performance? You take what you'd consider to be a 18 conservative value for fouling? You see my question?
19 MR. HOUGHTON: Yes, I see your question.
20 I'd want to confirm with the Chapter 15 folks to make 21 sure that we've got a --
22 MEMBER CORRADINI: Okay.
23 MR. HOUGHTON: -- complete answer there.
24 But there is an in-service inspection requirement to 25 review the visual that the actual In-Service NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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176 1 Inspection Program would be developed as a COL item.
2 And that would require cleaning as needed of the 3 vessel.
4 MEMBER CORRADINI: Okay. Thank you.
5 MR. HOUGHTON: All right. Next slide? So 6 I was discussing, the containment isolation valves, 7 they meet GDC-55, 56 and 57 with two exemptions being 8 requested. The first is for locating both of the 9 containment isolation valves outside of the 10 containment for GDC 55 and 56 penetrations. And the 11 second is with regard to the decay heat removal system 12 which is unique in that it constitutes a closed system 13 both inside and outside of the containment. So that's 14 not clearly addressed by the General Design Criteria, 15 so we have an exemption for locating the isolations 16 separately.
17 To support this exemption the piping 18 between the containment vessel and the containment 19 isolation valve is designed to meet the break 20 exclusion zone criteria. I mentioned previously the 21 designed and stamped primary system containment 22 isolation valve is Class 1; the secondary system 23 containment isolation valve is designed and stamped as 24 ASME Class 2. And as the staff mentioned previously 25 in their Chapter 3 review, the containment isolation NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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177 1 valves are hydraulically operated and all fail closed.
2 To the member's question on the basis for 3 this, the NuScale containment environment is unique.
4 It's a high-pressure, steam environment during 5 accident conditions. The containment design is unique 6 in that it is a metal vessel. So we determined that 7 the most appropriate design choice would be to put 8 both isolation valves outside of containment but as 9 close to the boundary as possible and have taken 10 appropriate cautions to eliminate the potential for 11 break in that area.
12 MEMBER CORRADINI: Remind me, the cartoon 13 shows as if they're bolted to the vessel. That's not 14 correct?
15 MR. HOUGHTON: They are welded to the 16 vessel. So this next slide shows that. So this is 17 looking at a representative primary system isolation 18 on the containment head. So you can see the nozzle 19 integral to the containment head itself, safe end 20 welded to the nozzle, and then the containment 21 isolation valves welded to the safe end.
22 CO-CHAIR RICCARDELLA: Would you go 23 through the materials for these various components, or 24 is that in closed session?
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178 1 the materials for these components. That information 2 is in the design certification document though.
3 MEMBER BLEY: You refer to these, and 4 there's a bunch of other valves in Chapter 6 that get 5 referred to as hydraulically operated. They're 6 hydraulically moved to their position for normal 7 operation of the plant, but they're -- are these 8 spring-closed or are they --
9 MR. HOUGHTON: So they use a --
10 MEMBER BLEY: -- some other mechanism?
11 MR. HOUGHTON: -- gas bottle that 12 essentially is the --
13 MEMBER BLEY: So they're gas-closed?
14 MR. HOUGHTON: -- spring that closes a 15 valve, yes.
16 MEMBER BLEY: Okay.
17 MR. HOUGHTON: So they're pneumatic-18 hydraulic --
19 (Simultaneous speaking.)
20 MEMBER BLEY: So, yes. And when we look, 21 like in the PRA, and we're going to talk about this 22 when we come visit you, the part of the valve that 23 moves them to the safe position is the part we're 24 really interested in. So using hydraulically-operated 25 failure rates for valves like this isn't appropriate.
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179 1 You want to use the, either air-operated or spring-2 operated, which is true for a bunch of other valves 3 we'll look at later, failure rates. Just wanted to 4 mention that we'll be looking at that more late.
5 MR. HOUGHTON: Thank you.
6 MEMBER SKILLMAN: Zack, what is the 7 hydraulic medium? What is the composition of it?
8 MR. HOUGHTON: So it is -- we do not have 9 the composition of the medium specifically identified 10 in the design certification or specifically selected 11 today. It's required to be compatible with the pool 12 environment and non-combustible. Tentatively we've 13 been using Houghto-Safe, which is a -- no relation and 14 a currently available hydraulic fluid that's used in 15 industry today.
16 MEMBER SUNSERI: It would also have to be 17 resistant to radiation changes, too, right?
18 MR. HOUGHTON: It'll have to be resistant 19 for the radiation in the environment that's in, but, 20 no, we're at the top of the containment vessel so a 21 ways a way.
22 MEMBER SUNSERI: Yes, but still I mean, 23 they've got a bioshield and all that stuff in there.
24 MR. HOUGHTON: Sure.
25 MEMBER SKILLMAN: Is it hydrocarbon? Does NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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180 1 it have a carbon derivative, carbon base?
2 MR. HOUGHTON: I don't have that answer 3 off the top of my head. Do we have somebody in 4 Corvallis that would have the answer to that question?
5 MEMBER BLEY: We'll be interested in 6 this --
7 (Simultaneous speaking.)
8 PARTICIPANT: -- give us a moment.
9 MEMBER BLEY: We'll be interested in the 10 same question on other hydraulically-operated valves.
11 MR. HOUGHTON: Okay. It may be something 12 for us to discuss when the staff's out, members are 13 out in July.
14 MEMBER CORRADINI: Is somebody on the line 15 that wants to help out here?
16 MR. McGEE: We'll have to get back to you 17 on this one.
18 MEMBER CORRADINI: Okay. Thank you.
19 CO-CHAIR RICCARDELLA: What is that in-20 service insert all about?
21 MR. HOUGHTON:
22 MR. HOUGHTON: Oh, thank you. So because 23 these valves are the containment isolation boundary 24 they have to be local leak rate tested in accordance 25 with Appendix J, which I'll discuss in a future slide.
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181 1 You can see on the figure that there's an in-service 2 insert shown installed on the inboard side of the 3 containment isolation valve set.
4 MEMBER BLEY: Can somebody point to that?
5 MR. HOUGHTON: Yes, so it's -- so here's 6 the in-service insert in the figure. So this would be 7 the normal operating configuration for leakage rate 8 testing. During and outage this insert would be 9 removed. A testing insert would be installed that 10 creates a leakage path barrier and that way we can do 11 testing of the individual valves.
12 PARTICIPANT: So that's a bolted flange?
13 MR. HOUGHTON: Correct.
14 MEMBER BLEY: How often do you do that 15 kind of in-service testing, or do you expect it to be 16 done?
17 MR. HOUGHTON: Two years and as needed, 18 right? As needed after maintenance.
19 MEMBER BLEY: Okay. But at least every 20 two years?
21 MR. HOUGHTON: Yes, I'll look for Gary to 22 confirm that.
23 Do we do local leak rate testing every two 24 years regardless of maintenance?
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182 1 Option A of Appendix J and then eventually we would --
2 I would imagine go to Option B, performance-based, but 3 initially the plant is going to be at Option A, which 4 would be every two years.
5 MEMBER BLEY: Okay. And what kind of a 6 functional test do you do afterwards to make sure you 7 can pass water when you need to? That you put the 8 right --
9 (Simultaneous speaking.)
10 MR. McGEE: -- what you asked.
11 MEMBER BLEY: -- insert back in after the 12 test.
13 MR. McGEE: Well, the inserts are clearly 14 marked. I mean, you'd have administrative control.
15 MEMBER BLEY: They always are, but if you 16 don't test, you're going to have one in the wrong --
17 wrong one in there one day.
18 MR. McGEE: Right, well, you're not going 19 to flow your process fluid when you start up. You're 20 not going to flow CVCS. You're not going to flow --
21 you're going to -- operations is going to know right 22 away that they're not --
23 MR. HOUGHTON: Yes, we'd have to go look 24 at each of the penetrations, but I believe Gary's --
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183 1 right, all of these during the startup would have flow 2 through them so you'd have an actual verification that 3 the right insert is in the --
4 MR. HOUGHTON: I think that would be 5 generally true. We'd want to go through line by line.
6 To my knowledge I don't recall this question coming up 7 in the past, so I --
8 MEMBER CORRADINI: But there's got to be 9 some sort -- I think where Dennis is going is there's 10 got to be some sort of testing upon restart of the 11 module to make sure that you have what you thought was 12 the proper --
13 MEMBER BLEY: Well, I am, yes. I mean, 14 and years ago we didn't do that kind of testing and 15 sure enough we had things left in place that shouldn't 16 have been left there.
17 Yes, sir?
18 MR. RAD: Hey, this is Zach Rad, Director 19 of Reg Affairs. So to your question, generally 20 speaking there is testing of your process systems 21 after you've put them back together following an 22 outage, but it's important to note that with that 23 insert in, this is the safest condition it could be in 24 relative to nuclear safety is that the containment is 25 isolated.
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184 1 MEMBER BLEY: I have to think that -- and 2 you probably ought to think that all the way through.
3 I mean, that's where you want it to be, under say 4 accident conditions, but if most of these are one-way 5 and some are another way, you might not be right.
6 MR. RAD: I didn't mean to imply that it 7 was desirable.
8 MEMBER BLEY: I know.
9 MR. RAD: I just mean to imply that --
10 MEMBER BLEY: But I'm not sure it's always 11 the safe thing.
12 MR. RAD: -- it's not going to prevent the 13 operation of a containment isolation valve from a 14 safety function perspective.
15 MEMBER BLEY: That's for sure.
16 MR. HOUGHTON: We understand the question 17 and it's something we'll take away and consider 18 internally.
19 MEMBER BLEY: Okay.
20 CO-CHAIR RICCARDELLA: So this is typical 21 of all of the Class 1 penetrations?
22 MR. HOUGHTON: Correct. Yes. For all of 23 the Class 1 penetrations, yes.
24 CO-CHAIR RICCARDELLA: And so looking at 25 this, the break exclusion zone has to be from the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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185 1 valve all the way to the containment vessel, right?
2 MR. HOUGHTON: That's correct.
3 CO-CHAIR RICCARDELLA: And you can not 4 take a break in there?
5 MR. HOUGHTON: That's correct.
6 CO-CHAIR RICCARDELLA: And you're doing 7 some special materials, inspections and things for 8 that, right?
9 MR. HOUGHTON: The materials, the 10 inspections are all listed in the design 11 certification, but we have applied the break exclusion 12 zone criteria, that full section.
13 All right. Next slide?
14 MS. NORRIS: Now seems like a good time to 15 bring up an answer. We had a question on the fouling 16 of the containment heat transfer surfaces. So I've 17 been referred to Section 6.2.2.4 of the FSAR, and this 18 addresses some of this. So hopefully that answers 19 your question.
20 MEMBER CORRADINI: I'll look. Thank you.
21 MEMBER BLEY: Say those again?
22 MS. NORRIS: 6.2.2.4 of the FSAR.
23 MR. HOUGHTON: All right. So we covered 24 this a little bit in the staff's presentation on 25 Chapter 3, but just to cover it again for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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186 1 containment isolation valves. So every containment 2 isolation valve has solenoid valves that actually 3 control the hydraulics. Those are located remotely on 4 a hydraulic control skid.
5 Each module has two separated dedicated 6 hydraulic control skids which are located in separate 7 areas of the reactor building and each hydraulic 8 control skid operates one division of valves. So in 9 that last figure that we saw each of the valves shown 10 in that one body would have their own dedicated 11 solenoid valves which are located on dedicated control 12 skids that support each module and are located in 13 separate areas. So we've remained divisional 14 separation as much as we can given the unique 15 consideration of our design of our isolation valves 16 being in the same area. And then this figure just 17 shows what I described in a bit more detail.
18 MEMBER REMPE: So I'd like to follow up on 19 your follow-up about Section 6.2.2.4, because it 20 refers us to Section 6.2.1.6 without any definition of 21 what periodic inspections is. And if I look at 22 6.2.1.6, all it says is we're going to have periodic; 23 again, undefined, in-service inspection to ensure 24 compliance with GDC-39, just -- that's for surface 25 fouling.
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187 1 Is that enough of an answer for you, Mike?
2 MEMBER CORRADINI: I'm still trying to 3 find it. I'm sorry.
4 MEMBER REMPE: I mean, didn't you want to 5 know how often they're going to inspect and --
6 MEMBER CORRADINI: I want to know -- I'm 7 just trying to understand; and maybe this is for the 8 accident analysis portion that's closed, what they 9 used to bound it? And if it's bound, then what are 10 they looking for to decide I'm out of bounds; I've got 11 to go clean it?
12 MR. HOUGHTON: Understood.
13 MEMBER CORRADINI: But I found it on 6.2-14 25 and it doesn't say very much. It's a paragraph.
15 MEMBER REMPE: And it refers -- 6.2.2.4, 16 which is where she referred us, says go to 6.2.1.6, 17 which again doesn't say what you're asking about, what 18 is the -- when do you start cleaning up?
19 MEMBER CORRADINI: So it doesn't have to 20 be answered today, but to the extent that you take 21 credit for some sort of fouling, I'd like to know what 22 it is and how do you know you've exceeded it?
23 MR. HOUGHTON: Understood.
24 MEMBER SUNSERI: Let me ask a follow-up 25 question. I'm trying to remember. So when you refuel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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188 1 though, you flood the containment, right? So it's 2 flooded and drained each refueling outage, right, at 3 a minimum?
4 MR. HOUGHTON: Yes.
5 MEMBER SUNSERI: Okay.
6 MR. HOUGHTON: All right. So moving on to 7 Section 6.2.5, Combustible Gas Control. So for the 8 NuScale design, again it would be the unique 9 containment arrangement with the containment held at 10 vacuum conditions during normal operation. We have 11 designed for the complete effects of hydrogen burn or 12 detonations within the containments. We have 13 requested an exemption from 10 CFR 50.44(c)(2), which 14 is a requirement for an inerted atmosphere or limited 15 hydrogen concentration.
16 The design itself provides a mixed volume.
17 Just inherent to its design of having no sub-18 compartments it relies on natural convection for 19 emergency core cooling system operation which creates 20 a mixing environment within the containment.
21 So we do however provide equipment for 22 monitoring of in-containment hydrogen concentrations 23 and we are capable of continuously monitoring hydrogen 24 and oxygen gas concentrations after both design-basis 25 and beyond-design-basis accidents. There are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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189 1 analyzers provided within the process sampling system.
2 I will note however that the exact performance 3 requirements of that monitoring equipment is still 4 something that is under discussion with the staff.
5 MEMBER CORRADINI: So is there a standing 6 hydrogen and oxygen concentration just due to the 7 fact --
8 SIRI ON CELL PHONE: Okay. I found this 9 on the web for hey, Sara, standing hydrogen --
10 (Laughter.)
11 MEMBER CORRADINI: I didn't do anything.
12 PARTICIPANT: Does she have an answer?
13 MEMBER CORRADINI: She didn't.
14 PARTICIPANT: Thank you, Alexis.
15 MEMBER CORRADINI: Thank you, Alexis.
16 Siri.
17 So there's a standing hydrogen and oxygen 18 concentration in the atmosphere, is that correct?
19 MR. HOUGHTON: That is not correct. So 20 the atmosphere is a vacuum under normal operations, so 21 it would be air --
22 MEMBER CORRADINI: But you have a relative 23 humidity in there all the time, so you have a standing 24 concentration due to essentially decomposition, 25 radiolitic decomposition, don't you?
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190 1 MR. HOUGHTON: Well, no. So we're 2 maintaining in a vacuum continuously during 3 operations. So are you speaking specifically during 4 operation or during an accident scenario?
5 MEMBER CORRADINI: Well, let's start 6 during operation.
7 MR. HOUGHTON: So during operation we're 8 maintaining in a vacuum continuously. So anything 9 that you have in there would be negligible, I'll say, 10 volumes or mass of anything left behind.
11 MEMBER CORRADINI: Okay. And now if I 12 take it to an accident, I'm kind of going with your 13 previous slide where you said you considered 14 detonation. So I'm trying to understand how you 15 evolve hydrogen concentration and what's the 16 associated oxygen concentration.
17 MR. HOUGHTON: So there would be some 18 amount of oxygen that would come out from dissolution, 19 as you mentioned. For our analysis -- okay, so in an 20 accident condition the regulations require you to 21 consider 100 percent fuel-clad water interaction.
22 MEMBER CORRADINI: Right.
23 MR. HOUGHTON: So hydrogen is very 24 dominating in what would have to be a severe accident 25 case for our plant to generate hydrogen under those NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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191 1 conditions because we maintain the core covered for 2 all design-basis accidents.
3 For the purposes of our integrity 4 evaluation we assume that there was sufficient oxygen 5 to detonate the hydrogen under all conditions, 6 however, in actual design you'll have very limited 7 amounts of oxygen available. It's a very hydrogen-8 limited scenario for our plant.
9 MEMBER CORRADINI: So you're making the 10 assumption there's sufficient oxygen? You're not --
11 there's not a calculation that computes what it would 12 be?
13 MR. HOUGHTON: Correct.
14 MEMBER CORRADINI: Okay. All right.
15 Thank you.
16 MR. HOUGHTON: And the only oxygen would 17 be from air and leakage or dissolution, so relatively 18 small amounts.
19 And then last bullets, we do include 20 connections for clean up of the environment after an 21 accident, should it be needed. So there's connections 22 to bring in skidded equipment as needed.
23 Next slide.
24 MR. SEXTON: Hi, this is Colin Sexton. If 25 I could just step in real quick and discuss options.
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192 1 When we do our calculation for how large of a 2 combustion could occur, the main element of that 3 calculation is how much oxygen could be present, 4 because oxygen is a limiting ingredient out of oxygen 5 and hydrogen in our combusting calculations. And the 6 elements that contribute to that oxygen in our 7 calculations are initial oxygen that could be present 8 in the containment and radiolitic production of 9 oxygen.
10 MEMBER CORRADINI: So, Colin, nice to hear 11 from you. So what is that value at -- so it's a 12 calculation in terms of that concentration and then an 13 assumption of combustion?
14 MR. SEXTON: Yes, what we do in our 15 calculation is we estimate the maximum amount of 16 oxygen that could be present by being quite 17 conservative on how much could be there initially and 18 how much could be developed radiolitically. And then 19 we calculate how big of an explosion you could 20 generate with that oxygen by optimizing the hydrogen 21 conditions to match. So oxygen limiting, and we use 22 the oxygen as effectively as we can to generate a 23 limiting combustion event in our analyses.
24 MEMBER CORRADINI: Okay. I think I got 25 it. Thank you. Thank you very much.
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193 1 MR. HOUGHTON: Okay. So containment 2 leakage testing. All the containment isolation valves 3 and the passive containment isolation barriers -- so 4 these would be manways and bolted joints.
5 MEMBER CORRADINI: So can we go back? I'm 6 sorry. I'm reading my notes.
7 So there's an exemption -- if I'm 8 remembering correctly, there's an exemption on 9 something to do with sampling. Does this somehow 10 intersect with combustible gas control, that you're 11 sampling to determine what your concentration is in 12 containment for hydrogen monitoring?
13 MR. HOUGHTON: That is the portion that's 14 still under discussion with the staff.
15 MEMBER CORRADINI: So tell me more about 16 that since I can't remember. I wrote a note to myself 17 that I should ask.
18 MR. HOUGHTON: It's related to our 19 alternate source term discussions that are ongoing.
20 I wouldn't want to give any preliminary information in 21 that area.
22 MEMBER CORRADINI: So this is more a 23 matter of the method in which you're going to do the 24 sampling so you can determine the concentrations?
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194 1 around the exact conditions under which an operator 2 would have to go out, establish the equipment, what 3 sort of environment are they going to see, how are we 4 accounting for those aspects of taking the sample.
5 MEMBER CORRADINI: Well, I know it's an 6 open item, but it's still being discussed with staff.
7 There's not been a response back in terms of the open 8 item, or is it still being reviewed by staff? What's 9 the status of this?
10 MR. HOUGHTON: I'll look for the -- it's 11 being reviewed?
12 MEMBER CORRADINI: Well, if -- I'll ask 13 the staff the same question. So if we're going to go 14 -- if you're going to go look for the staff, we'll 15 just get them next.
16 MR. HOUGHTON: Okay.
17 MEMBER CORRADINI: Okay. Thank you.
18 MR. HOUGHTON: Thank you.
19 All right. So Section 6.2.6. So we 20 designed all of the containment barriers to support 21 local leak rate testing, and local leak rate testing 22 would be done per 10 CFR 50, Appendix J for all type 23 B and type C tests. So those are -- passive barriers 24 such as manways, bolted connections would constitute 25 type B testing. Type C testing is your valves, right, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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195 1 the containment isolation valves that we discussed.
2 We have requested exemption from GDC-52 3 for doing the type A integrated leak rate testing, 4 again because of the unique aspects of our containment 5 design. Some of the aspects of our design which 6 support that exemption are that it's fabricated and 7 tested similar to reactor pressure vessel requirements 8 because it is a -- stamped as a Class 1 vessel. All 9 known leakage pathways are tested through local leak 10 rate testing. We have a comprehensive In-Service 11 Inspection Program meeting Class 1 requirements for 12 all those welds, so those welds that we saw that make 13 up the connection to the valves. And then the 14 containment is also hydrostatically tested.
15 MEMBER CORRADINI: So the argument is that 16 you basically are always monitoring? So you know what 17 your leak rate is just by the continuous monitoring of 18 the evacuation system? I'm trying to understand 19 the --
20 MR. HOUGHTON: That's a correct statement, 21 but it really doesn't play heavily into our exemption 22 request because we still have to do testing under 23 different conditions. The monitoring that we're doing 24 under normal operation is a different set of 25 conditions than we would see during design-basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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196 1 accident --
2 MEMBER CORRADINI: Sure.
3 MR. HOUGHTON: -- conditions, which is 4 what we would have to try to replicate in a type A 5 test.
6 MEMBER CORRADINI: But the argument is by 7 your individual isolation valve by isolation valve 8 testing that meets the standard of knowing what your 9 leak rate would be?
10 MR. HOUGHTON: Correct.
11 MEMBER CORRADINI: Okay. Thank you.
12 MR. HOUGHTON: So going to the next slide.
13 So there are some additional commitments that we made 14 to support this exemption. As I mentioned, the 15 containment is hydrostatically tested, however, the 16 controls over hydrostatic testing are a little bit 17 different. We test at a -- you test at a higher 18 pressure under hydrostatic conditions.
19 The code for hydrostatic testing doesn't 20 have requirements on bolt pre-load, bolt gasketing, so 21 we created a separate -- what we call a pre-service 22 design pressure leakage test that is controlled by 23 ITAAC that sets additional controls over bolt pre-24 load, gaskets used configuration so that prior to 25 operation we're doing a test to show that no leakage NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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197 1 pathway develops under limiting conditions.
2 We have done an analysis to support our 3 design certification and we have a COL item to follow 4 up on this analysis for the as-built containment to 5 show that we maintain flange contact pressure at 6 accident temperature concurrent with peak accident 7 pressure conditions.
8 So we use a finite element analysis model 9 to look at the prying that would happen at every one 10 of our bolted joints and ensure that we maintain 11 appropriate contact where the gasket itself is. And 12 that's important because it supports our position that 13 the leakage that we see under type B or type C testing 14 conditions is identical to what we would see under 15 type A conditions.
16 MEMBER SKILLMAN: Are all your bolted 17 connections -- you have a fiber flange, a flexitallic 18 gasket. What kind of gasket do you have such that 19 you're determining whether or not that gasket is 20 threatened by that pressure?
21 MR. HOUGHTON: The exact gasket selection 22 is in the technical report where we talk about the 23 details about our assumptions for this analysis.
24 That's also part of the purpose of the COL item is to 25 ensure that the actual -- if there's changes in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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198 1 final selected gasket material type, that that all 2 gets considered in this analysis.
3 MEMBER SKILLMAN: Thank you.
4 MR. HOUGHTON: And then finally we made a 5 commitment to verify the preload that was -- that's 6 been applied. So after maintenance, that sort of 7 thing, we'll ensure that we've got the appropriate 8 preload because it is important for that analysis that 9 we've done.
10 Next slide. And here are the COL items.
11 CO-CHAIR RICCARDELLA: But you actually 12 hydrostatically test that containment pre-service at 13 1,000 PSI, some really high pressure like --
14 MR. HOUGHTON: Correct, at pre-service, 15 yes.
16 CO-CHAIR RICCARDELLA: No in-service?
17 MR. HOUGHTON: No, we do not make a 18 commitment for in-service hydrostatic testing.
19 CO-CHAIR RICCARDELLA: Or leak tests? No 20 in-service leak tests?
21 MR. HOUGHTON: Local leak rate testing 22 only.
23 PARTICIPANT: That's the type A they're 24 asking for the exemption.
25 CO-CHAIR RICCARDELLA: Okay. Got it.
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199 1 MR. HOUGHTON: All right.
2 MS. NORRIS: So coming back real fast to 3 the hydraulic fluid question, so we have a vendor-4 specified fluid, which is Fyrquel, F-Y-R-Q-U-E-L. But 5 this is not specified in the DCA. This just is what 6 is suggested. That is a hydrocarbon.
7 So the words in the DCA are in Section 8 5.6.1 and 5.6.2. Basically the hydraulic actuator 9 shall use water or fluid demonstrated compatible with 10 reactor building pool water cleanliness requirements 11 as provided in the reference so that fluid leakage in 12 the pool is not detrimental. And 5.6.2. says any 13 hydraulic fluid including corrosion inhibitors if 14 added shall be non-flammable and non-combustible. So 15 that's what we have in the DCA.
16 MEMBER SKILLMAN: Fyrquel. The reason I 17 asked; and I mentioned this I think at a first 18 meeting, you've got a -- what, a 71/2-8 million gallon 19 pool, 12 reactors?
20 MR. HOUGHTON: Correct.
21 MEMBER SKILLMAN: You're going to keep the 22 temperature in that pool to probably 100, under a 100 23 so you're not steaming inside your building. You're 24 going to work to keep your relative humidity low.
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200 1 fluid. We thought that would be dandy. What we 2 didn't realize is that through the course of events we 3 actually had coliform in the water because we used a 4 clean water supply, but it was not pure from coliform.
5 And the de-fueling at TMI-2 stopped for 6 several months because we grew leafy green vegetables 7 in the reactor that completely blocked all of our 8 activity. We were on standby for probably three 9 months. And what happened was the hydrocarbon; and 10 Fyrquel is one of them, became a food source for the 11 material that was in the water and we had to use 12 hydrogen peroxide to purify the water.
13 And so if there is any hydrocarbon 14 exposure, I would suggest that that is something you 15 want to act on now. You do not want to continue with 16 any hydraulic fluid that can add hydrocarbon to your 17 pool. You want either having clear water with boric 18 acid or some other substance, but not anything that 19 would feed a hydrocarbon base. That is a multi, 20 multi-million dollar decision.
21 MS. NORRIS: Yes, sorry about that. As we 22 were talking corrections came in. It is not 23 hydrocarbon. It is in fact phosphate-based, probably 24 for those reasons. Also that's not from the FSAR.
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201 1 earlier.
2 MEMBER SKILLMAN: Thank you.
3 MEMBER BLEY: Zack, can I take you back to 4 the containment evacuation system?
5 MR. HOUGHTON: Yes.
6 MEMBER BLEY: I'm looking at the drawing.
7 And you come up out of the containment and you go into 8 those vacuum pumps and into a condenser and drain, but 9 it also shows line coming with valves open into the 10 vacuum pumps from the service air system. Is that 11 some kind of cooling for the vacuum pumps or 12 something? You're not mixing these streams. That 13 wouldn't make sense.
14 MR. HOUGHTON: Let me --
15 MEMBER BLEY: From SAS, which I'm guessing 16 is service air system.
17 MR. HOUGHTON: Yes, I believe that's only 18 for startup, although there may be some other unique 19 conditions where we use that. So I'd want to go 20 confirm with the systems engineers.
21 MEMBER BLEY: Okay. And have them take a 22 look at the drawing that's in the FSAR, because that 23 shows the valves open from those systems, which is a 24 little funny.
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202 1 fairly complex way to collect the sample over time.
2 As I read Chapter 6, or 9, it seemed as if you just 3 let it collect until you get some amount and you're 4 keeping a time run on that so you can get a leak rate.
5 And then I guess if it starts to fill up, you would 6 dump it. That's what it looks like, except you didn't 7 -- it also says you can do grab samples in there, 8 which would mean you'd have to drain it and then do 9 a --
10 (Simultaneous speaking.)
11 MR. HOUGHTON: Right, and the grab samples 12 would be for chemistry purposes, but we do let it 13 collect. When it --
14 MEMBER BLEY: Over time?
15 MR. HOUGHTON: -- as it fills up it would 16 have to be dumped.
17 MEMBER BLEY: Yes, you don't have some 18 kind of meter --
19 (Simultaneous speaking.)
20 MR. HOUGHTON: But we can monitor it as 21 it's collecting to see --
22 MEMBER BLEY: Yes.
23 MR. HOUGHTON: -- what the rate is.
24 MEMBER BLEY: That's what I thought.
25 Okay. Thank you.
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203 1 MR. LASSITER: And I would add that the 2 containment pressure is a fast response leakage 3 mechanism, but the --
4 MEMBER CORRADINI: Well, but I --
5 (Simultaneous speaking.)
6 MR. LASSITER: -- would build over time as 7 a longer average.
8 MEMBER CORRADINI: We're talking -- but I 9 guess I'm back to the question that kind of led from 10 Member Ballinger, which is if the leakage is below 11 your evacuation pump capacity, I wouldn't see it. I'd 12 see it in the collection. If it's above the 13 evacuation pump capacity, I'd see it in the pressure.
14 At least that's how I'm thinking this things 15 functions.
16 MR. LASSITER: And if the pressure 17 continues to build, we have --
18 MEMBER CORRADINI: Right.
19 MR. LASSITER: -- automatic trip on 20 containment pressure.
21 MEMBER CORRADINI: Okay.
22 MR. HOUGHTON: And also mention it's not 23 exactly a 0.1 --
24 MEMBER CORRADINI: Oh, I understand.
25 MR. HOUGHTON: -- sort of problem, right?
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204 1 As leakage increases your pump performance changes 2 depending on now much volume it's removing.
3 MEMBER CORRADINI: Sure.
4 MR. HOUGHTON: Yes.
5 MEMBER CORRADINI: Okay. Thank you.
6 MR. HOUGHTON: All right. Section 6 --
7 MR. ARTHUR: This is J.J. Arthur again.
8 I can answer the question about the service air 9 connection. That's actually used to remove liquid 10 from the containment after refueling, so we actually 11 use pressurized air to drive out water.
12 MEMBER BALLINGER: Oh, back to the vacuum 13 system again.
14 MEMBER CORRADINI: Are you happy?
15 (Simultaneous speaking.)
16 MEMBER BLEY: -- this drawing and that 17 didn't -- I understand that, but that didn't seem to 18 match up to this, but maybe you can help me.
19 MEMBER BALLINGER: The vacuum system 20 connects to the top of the containment, right?
21 MR. HOUGHTON: Correct.
22 MEMBER BALLINGER: You've got this giant 23 condensing wall that's the containment. What if water 24 condenses down into the bottom?
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205 1 refers back to the graph that I mentioned that's in 2 Chapter 5 that describes -- we -- if we maintain the 3 pressure low enough inside containment, then you won't 4 have condensation --
5 MEMBER BALLINGER: Right.
6 MR. HOUGHTON: -- on the walls, right? So 7 we have a limit on vacuum pressure in order to support 8 the containment evacuation system monitoring feature.
9 MEMBER BALLINGER: So you are not going to 10 allow condensation on the walls?
11 MR. HOUGHTON: Correct, we'll maintain the 12 pressure below the saturation pressure that --
13 MEMBER BALLINGER: Okay.
14 MR. HOUGHTON: -- corresponds to the 15 temperature of the pool.
16 MR. LASSITER: That's why have a safety-17 related pressure trip below atmospheric.
18 MR. HOUGHTON: All right. Section 6.2.7 19 covers --
20 MEMBER SKILLMAN: Let me ask one more 21 question: So let's take a module. You remove it from 22 its operating bay. You take it to the refueling 23 stand. You disassemble it. And all of that equipment 24 is now bathed in borated water, 800 PPM or -- it's a 25 number like that. Over the course of the life of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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206 1 plant will not that boric acid become a coating on the 2 inside and the outside, particularly on the inside of 3 that vessel? And is there a wash-down before you 4 reassemble and bring that vessel back into its 5 operating bay?
6 MR. HOUGHTON: Well, that goes I think to 7 the other member's question. So the exact frequency 8 that -- at which we would have to do a cleaning isn't 9 defined today, but there's an inspection requirement 10 to look for exactly the collection that you're 11 describing and there would be a requirement to clean 12 it off if you were --
13 MEMBER SKILLMAN: Okay.
14 MR. HOUGHTON: -- building up a scale.
15 MEMBER BLEY: The kind of thing I've been 16 stewing over; and we talked to the PRA people about 17 this, is -- and I hadn't thought about this scenario, 18 but if you begin to get some leakage -- I mean, you 19 walk around a current PWR and anywhere boric acid 20 solution leaks out and sees the atmosphere you start 21 getting crystallization and this gooey stuff.
22 I don't know why you wouldn't get that 23 inside on the containment surface. And if then this 24 expanded to be an accident, then would you get the 25 same heat transfer? And that's certainly not NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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207 1 reflected anywhere in the PRA if that's a potential 2 problem. We were worried about other kinds of 3 contamination on the surface as well, but that's an 4 obvious one that --
5 (Simultaneous speaking.)
6 CO-CHAIR MARCH-LEUBA: Boric acid is very 7 volatile and certainly at atmospheric pressure it will 8 be a crystal, maybe not at that --
9 MEMBER BLEY: What it leaves behind --
10 CO-CHAIR MARCH-LEUBA: Nothing. It 11 becomes a gas and it goes to the pump. Maybe. I 12 mean, I know it boils off at -- it evaporates at 300 13 degrees CM in --
14 (Simultaneous speaking.)
15 MEMBER BLEY: Oh, I believe that, yes.
16 CO-CHAIR MARCH-LEUBA: Yes, so the 17 pressure.
18 MEMBER BLEY: I don't know what the 19 containment temperature is.
20 CO-CHAIR MARCH-LEUBA: It's --
21 (Simultaneous speaking.)
22 MEMBER CORRADINI: Containment -- well, it 23 can't be any -- it can't be much higher than the pool 24 temperature.
25 MEMBER BLEY: I wouldn't think so.
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208 1 CO-CHAIR MARCH-LEUBA: Yes.
2 MEMBER BLEY: So that --
3 CO-CHAIR MARCH-LEUBA: I don't know 4 what --
5 MEMBER BLEY: -- whether it comes out 6 at --
7 (Simultaneous speaking.)
8 CO-CHAIR MARCH-LEUBA: I --
9 (Simultaneous speaking.)
10 MEMBER BLEY: -- it leaves behind a hell 11 of a residue.
12 CO-CHAIR MARCH-LEUBA: I don't know what 13 evaporation temperature is at, but whatever pressure 14 they operate at, but it is a possibility, I guess.
15 MEMBER CORRADINI: But I thnk what you're 16 getting at is kind of what I was getting at, is at the 17 end of cycle before I would go through a refueling, 18 which they would then fill it with water, everything 19 would dissolve, everything would be nice and rosy 20 again. There would be some sort of fouling that you'd 21 have to consider.
22 MEMBER BLEY: And it seems like over time 23 there might be --
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209 1 Inspection Program are you doing on the CNV?
2 MR. HOUGHTON: So there's a requirement to 3 do a visual inspection of the CNV to look for exactly 4 this. There's also a requirement to have a Boric Acid 5 Control Program, right? So applicants will have to be 6 inspecting, looking for the boric acid crystals that 7 might get left behind from a leak, for example. So 8 there's requirements in the In-Service Inspection 9 Program and in the Boric Acid Control Program to look 10 for these sorts of indications.
11 CO-CHAIR RICCARDELLA: So visual at every 12 outage, but then is there also a volumetric inspection 13 as part of this slide that's up here now?
14 MR. HOUGHTON: There's volumetric 15 inspections of welds as described by the In-Service 16 Inspection Program.
17 So Section 6.2.7 covers fracture 18 prevention of the containment vessel. As mentioned, 19 the containment vessel meets the relevant parts of 20 GDC-116 and 15 and is there for ASME Section III, 21 Subsection NB. The ferritic pressure boundary 22 materials meet Section III, Subsection NB fracture 23 toughness requirements. And again, as discussed the 24 austenitic stainless steel XM-19 is not subject to 25 neutron embrittlement at the fluence levels we've --
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210 1 (Simultaneous speaking.)
2 CO-CHAIR RICCARDELLA: Again, following on 3 what I said, it's -- then is it also Section 11, IWB 4 in your In-Service Inspection Program, which is like 5 what -- the way you would inspect the reactor vessel?
6 MR. HOUGHTON: That's -- the inspections 7 for the containment vessel are all captured in Table 8 6.2-3, so I'd refer you there to see the exact 9 requirements.
10 CO-CHAIR RICCARDELLA: Okay.
11 MR. HOUGHTON: Okay. And with that I'll 12 turn it over to Dan to discuss emergency core cooling.
13 MR. LASSITER: All right. Thanks, Zack.
14 So we're going to talk about the NuScale 15 emergency core cooling system design. I know there's 16 been some questions about detailed figures on the 17 valves, and those figures will be in the closed 18 session this afternoon, but there will be some 19 simplified figures in this presentation.
20 The NuScale --
21 CO-CHAIR MARCH-LEUBA: Can you talk closer 22 to the microphone?
23 MR. LASSITER: Sure.
24 CO-CHAIR MARCH-LEUBA: Otherwise, I won't 25 hear you.
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211 1 MEMBER BLEY: And speak a little louder, 2 if you can.
3 MR. LASSITER: Okay. The NuScale ECCS 4 system cools the core when it cannot be cooled by 5 other means such as loss of coolant accidents. As Tom 6 described earlier, the ECCS consists of five valves:
7 three reactor vent valves on the top of the reactor 8 vessel and two reactor recirculation valves on the 9 sides. The ECCS works in conjunction with the 10 containment isolation system to retain the required 11 inventory for operation. ECCS valves are normally 12 held closed by electrical power and the ECCS is 13 actuated on a high containment level and loss of AC 14 power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
15 CO-CHAIR MARCH-LEUBA: Wait, the RRVs are 16 held open by electric power, AC power and pressure 17 difference. The IAB keeps them closed, right?
18 MR. LASSITER: The IAB is not normally 19 acting to -- acting in the ECCS valve system. They're 20 normally open and unengaged. The only thing keeping 21 the main valves closed is the trip solenoid valves 22 staying closed.
23 CO-CHAIR MARCH-LEUBA: So you can trip the 24 AC power and then even though the pressure difference 25 is larger than the set point, it would open?
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212 1 MR. LASSITER: Okay. Well, it is 2 conditional on the prevailing reactor pressure as to 3 whether the IAB is --
4 (Simultaneous speaking.)
5 CO-CHAIR MARCH-LEUBA: So the IAB is not 6 holding it closed, but it will the moment you remove 7 the power?
8 MR. LASSITER: Dependent on the reactor 9 pressure, correct.
10 CO-CHAIR MARCH-LEUBA: All right. So it's 11 both AC power and pressure that keeps them closed?
12 MEMBER CORRADINI: So let me ask the 13 question a little differently because eventually 14 you're going to explain this to us and after you give 15 us the quiz we'll pass.
16 If the IAB fails in the direction of 17 allowing the ECCS upon a loss of power to open, the 18 effect is a higher containment pressure with the 19 blowdown for a brief time. That's the only difference 20 that I can see. Am I missing something? Maybe we're 21 into accident analysis and you can tell me to wait 22 until tomorrow, which I'm happy to do or wait until 23 closed session.
24 But I'm trying to understand a failure of 25 the IAB upon a demand to open essentially allows for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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213 1 just a starting point of a higher stored energy and 2 pressure in the primary system.
3 MR. HOUGHTON: It would be dependent on 4 the exact event that you're looking at. So one of the 5 limiting events is inadvertent opening of a relief 6 valve.
7 MEMBER CORRADINI: Right.
8 MR. HOUGHTON: So multiple valves 9 opening --
10 MEMBER CORRADINI: Right.
11 MR. HOUGHTON: -- would be a -- critical 12 heat flux limits or something that we also look at in 13 addition to the peak pressure.
14 MEMBER CORRADINI: Right. Okay. But 15 memory -- well, I don't want to get ahead of 16 ourselves. We're going to be back there tomorrow.
17 So, okay. Fine.
18 MR. HOUGHTON: Correct.
19 MEMBER BLEY: You said they're 20 electrically. The solenoid-operated FSAR says they're 21 hydraulically operated.
22 MR. LASSITER: They're hydraulic or pilot-23 operated main valves, but the actuator which controls 24 the hydraulic fluid --
25 MEMBER BLEY: Oh, that's right.
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214 1 MR. LASSITER: -- is a solenoid, correct.
2 MR. HOUGHTON: Well, see this in more 3 detail in the closed session in future slides.
4 MR. LASSITER: Yes, maybe we'll --
5 MEMBER CORRADINI: Yes, well, you can just 6 tell us to wait.
7 MR. LASSITER: Okay.
8 MEMBER CORRADINI: That's acceptable. And 9 I have a feeling --
10 (Simultaneous speaking.)
11 MEMBER BLEY: -- that's probably a lot 12 easier, yes.
13 MR. LASSITER: Yes, I would defer these --
14 this review of the detailed valves to the closed 15 session.
16 Next slide, please?
17 CO-CHAIR MARCH-LEUBA: Oh, no, no, no.
18 Don't move. This is something I should have brought 19 up during Chapter 19, but let me just put it on the 20 record.
21 It actuates on high CNV level the moment 22 the level goes a little bit higher than the valves, 23 but that level is measured by this famous 24 laser-guided --
25 MEMBER REMPE: Radar.
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215 1 CO-CHAIR MARCH-LEUBA: -- radar --
2 MEMBER REMPE: Yes.
3 CO-CHAIR MARCH-LEUBA: -- radar bouncing 4 back and forth from the top of the vessel. Now my 5 suspicion; and I haven't seen any detail, is that 6 those radar-level sensors are digital and they will --
7 they would have a microprocessor on them that would 8 process the radar signal to see the phase delay on the 9 bouncing back.
10 MR. LASSITER: Right.
11 CO-CHAIR MARCH-LEUBA: And therefore once 12 you consider common-cause failure between all those 13 four microprocessors that those four level sensors may 14 fail at the same time. And why is this a problem?
15 Because for two years that radar is out of sensor. I 16 mean, there is no level and you're sending the radar, 17 nothing bounces back. So you are not testing -- those 18 level sensors are working for two years and you're 19 hoping that because you tested them during refueling 20 they still work now.
21 And I wish I had asked this question 22 before during the Chapter 19 evaluation, but the 23 common-cause failure of the instrumentation is not 24 beyond the realm of possibility, and especially a 25 complex instrumentation that has its first nuclear NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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216 1 application. Now your question is a Chapter 19 guy, 2 but --
3 MR. HOUGHTON: We understand the question 4 and we'll take that back and discuss with both chapter 5 19 and our I&C counterparts as well to fully consider 6 that.
7 CO-CHAIR MARCH-LEUBA: And I know that Dr.
8 Corradini would like us to help your design by myself.
9 I'm looking for some analog switches, that red line, 10 and then put the radar as a non-safety grade system to 11 measure the level.
12 MR. HOUGHTON: We appreciate the comment.
13 CO-CHAIR MARCH-LEUBA: But it is something 14 that we may have to come back to chapter 19 when we 15 come back to chapter 19.
16 MEMBER DIMITRIJEVIC: Chapter 19 only goes 17 through there to assign the failure rates. How the 18 valve operates, that belongs here. And we agreed to 19 discuss operation of this with either a visit or here, 20 so the chapter 19 just models reality. We have to 21 know what reality actually is so we know what will be 22 appropriate model for it.
23 MEMBER REMPE: Along with those other 24 concerns, degradation with radiation? Is that so? Is 25 that a concern?
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217 1 CO-CHAIR MARCH-LEUBA: The main concern I 2 have is that for two years that level sensor doesn't 3 measure nothing.
4 MEMBER REMPE: Right.
5 CO-CHAIR MARCH-LEUBA: If it fails, you 6 will know it failed.
7 MEMBER REMPE: And what causes it? Well, 8 many things -- is a radiation environment, there's 9 it's a lot of things. I'm very interested in --
10 CO-CHAIR MARCH-LEUBA: My main concern is 11 for two years you have a radar that is blinking and 12 not receiving anything back. And then you hold that 13 in a severe accident and it probably will work, but --
14 MEMBER BLEY: It's not seeing water, but 15 it should be seeing the metal at the bottom of the 16 vessel.
17 CO-CHAIR MARCH-LEUBA: Somebody that know 18 how it works it should --
19 MEMBER BLEY: Just a heads up, when we get 20 back to chapter 7, we didn't talk about that during 21 chapter 7.
22 MEMBER REMPE: Yes, there was no evidence 23 about this. We learned about it after chapter 7.
24 MR. LASSITER: We'll take that question 25 back to our instrumentation design group.
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218 1 MEMBER CORRADINI: It will be one of the 2 questions in July.
3 MR. LASSITER: Okay. Moving on, there are 4 a couple of slides animating the ECCS system and the 5 NuScale module. It's getting ahead of me a bit here, 6 but on the ECCS demand, the ECCS valves are opened.
7 Steam is relieved out of the top reactor vent valves.
8 In a situation such as a loss of coolant accident, a 9 CVC pipe break which is an un-iceable pipe break 10 inside containment, high containment level is reached.
11 By that time reactor pressure has decreased to some 12 level and ECCS is actuated.
13 Go to the next slide, please.
14 CO-CHAIR MARCH-LEUBA: What signal opens 15 the vent valves?
16 MR. LASSITER: MPS, our module protection 17 system, sends a signal to actuate all the ECCS valves 18 into the open position. They're actuated by removing 19 power to the solenoids.
20 CO-CHAIR MARCH-LEUBA: Yes.
21 MR. LASSITER: And the signals are high 22 containment level. That's really the only signal 23 other than a loss of AC power --
24 CO-CHAIR MARCH-LEUBA: How are you going 25 to get to that level unless you have the vent valves NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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219 1 going?
2 MR. LASSITER: If there's a pipe break.
3 CO-CHAIR MARCH-LEUBA: In the LOCA.
4 MR. LASSITER: Correct.
5 CO-CHAIR MARCH-LEUBA: That would be the 6 only time the ECCS would be actuated.
7 MR. LASSITER: Correct.
8 MEMBER CORRADINI: So let me just repeat 9 what you said just so we're on the same page. I would 10 expect with any sort of LOCA inside containment, I'd 11 see a pressure rise that can't be maintained by the 12 evacuation system. Therefore, it would be a high 13 pressure signal or a rate of pressure rise that would 14 essentially trigger the RVVs. Am I wrong?
15 MR. HOUGHTON: Well, before you get to 16 that point, you'd have a containment isolation and a 17 reactor trip on other signals.
18 MEMBER CORRADINI: Thank you.
19 MR. HOUGHTON: So if you continue to lose 20 volume into the containments and raise level, then 21 once level is high enough is when you get the ECCS 22 actuation.
23 MEMBER CORRADINI: Okay, and it's the RVVs 24 and the RRVs will then open in some sequence.
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220 1 simultaneously.
2 MEMBER CORRADINI: Okay.
3 MEMBER BLEY: But they don't open until 4 there's a enough water that you have water to move 5 back into the vessel through the ECCS valve.
6 CO-CHAIR MARCH-LEUBA: This movie is kind 7 of flaky because you seem to open the valves before 8 the level loops and to open the valves you need to 9 have the water outside. Your level opens valves and 10 the water -- the level wasn't really dropped inside.
11 MR. HOUGHTON: Correct.
12 MEMBER CORRADINI: Let me just go back to 13 my question so that I'm not -- because you said it 14 much better than I did which is I had some sort of 15 detection of a pressure rise or a rate of pressure 16 rise above some sort of specification. And then I go 17 to containment isolation, reactor trip, turbine 18 isolate, turbine trip, reactor trip. And then I would 19 assume I sequentially first open the RVVs and then 20 after some delay time open the RRVs. Otherwise, I'd 21 start blowing down in both locations and losing 22 inventory. But I guess I don't care because I get to 23 some sort of level which would cause a recirculation 24 anyway. Am I there?
25 MR. LASSITER: Yes.
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221 1 MEMBER CORRADINI: So all five of them 2 upon a signal will open?
3 MR. LASSITER: Correct.
4 MEMBER CORRADINI: Okay.
5 CO-CHAIR MARCH-LEUBA: But that signal 6 being the level in containment is higher.
7 MEMBER CORRADINI: No, it would be a 8 pressure rise.
9 CO-CHAIR MARCH-LEUBA: What is it?
10 MR. LASSITER: I think what Zack was 11 describing earlier is that earlier in the sequence 12 there would be a reactor trip and a containment 13 isolation based on other signals such as high 14 containment pressure or low pressurizer level. And if 15 the event progresses and the core is not being -- able 16 to be cooled by other means, DHR, for example, and 17 there's water spilling into containment, once there's 18 sufficient level for ECCS operation, that's where ECCS 19 is actuated by the ECCS.
20 CO-CHAIR MARCH-LEUBA: As it says in the 21 previous slide, the only thing that opens the five 22 valves is high level in containment, nothing else 23 opens it.
24 MEMBER BLEY: Loss of all electrical 25 power.
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222 1 CO-CHAIR MARCH-LEUBA: Loss of DC power.
2 MR. LASSITER: Correct.
3 CO-CHAIR MARCH-LEUBA: If it is good for 4 you, you should encourage it.
5 MEMBER BLEY: Or an operator can open --
6 MEMBER CORRADINI: An operator can 7 override and do a manual actuation.
8 MR. LASSITER: Operator can actuate ECCS 9 from the control room.
10 MEMBER BLEY: And that's like one switch 11 that opens both valves.
12 MR. LASSITER: Correct.
13 MR. HOUGHTON: But the IAB would go off if 14 it was above the IAB pressure. This is all going to 15 be -- we'll see that the pictures that will show this, 16 I think in more detail in the closed session where you 17 can actually see the components and the pressure 18 chambers and how -- the operation of the valve --
19 MEMBER CORRADINI: Thank you for 20 explaining. Appreciate it.
21 MR. LASSITER: Move onto the next slide of 22 the animation here. So I think we're aware now but 23 the steam vents into the containment region, condenses 24 in the containment due to the containment being 25 submerged in the reactor pool. Liquid level rises NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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223 1 above the core and also above the reactor 2 recirculation valves.
3 MEMBER CORRADINI: Can you point to recirc 4 valves? I think I see them. Oh, here they are.
5 Thank you.
6 MR. LASSITER: Okay, reactor recirculation 7 valves and the water level was rising up in that 8 animation there.
9 You can go to the next slide.
10 MEMBER CORRADINI: While you're still 11 there -- oh, well.
12 MR. LASSITER: So the cold water comes 13 back into the reactor vessel by the reactor 14 recirculation valves down the downcomer, reenters the 15 core, the inlet to the core. You see that again the 16 liquid travels through the riser. It's steamed off 17 back through the reactor vessel vent valves, the ECCS 18 vent valves on the top and this establishes the long-19 term cooling of ECCS in two phase natural circulation 20 cooling system.
21 Next slide.
22 MEMBER CORRADINI: So let me ask again if 23 we're straying into chapter 15 just tell us to wait 24 until tomorrow.
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224 1 of the lower part of containment is a much smaller gap 2 than all the things we've seen in the animation. So 3 how much water does it take to essentially fill up to 4 the RRV?
5 In this picture, it strikes me as about 6 one tenth of the total free volume to fill to the RRV.
7 And in all the animations, it looks more.
8 CO-CHAIR MARCH-LEUBA: The more important 9 question is how high above the fuel is the inside 10 level when the outside level reaches the set point?
11 MR. LASSITER: I'll answer the question 12 this way. The reactor recirculation valves are about 13 six feet above the top of the core. And the level of 14 -- you know, the settling level of retained coolant 15 inside the reactor module is dependent on the 16 accident. So I think I'd like to defer the question 17 about the level of water to tomorrow to chapter 15 18 which is tomorrow, I believe.
19 MEMBER CORRADINI: That's fine. That's 20 fine.
21 MR. LASSITER: Because it varies depending 22 on if it's a -- what type of break is occurring.
23 MEMBER CORRADINI: Right, but the actual 24 volume of water doesn't change, so I'm just looking 25 for the volume up to the RRVs which is the trip point.
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225 1 MR. HOUGHTON: We don't have that exact 2 number. That's something we can look at, but your 3 point is correct that in some of the artistic 4 renderings that we see that clearly shows more space 5 there, but in the figure that you held up that was 6 shown earlier in the slides here which is a very small 7 volume in there that's for two reasons, right, it's 8 the exact number we can come back to it.
9 MEMBER CORRADINI: That's fine.
10 MR. HOUGHTON: But it's small for two 11 reasons, right, so that we can quickly correct 12 condensate and get volume into the lower region, but 13 it's also there to minimize the heat retransfer 14 resistance once we're in ECCS operation, so we keep a 15 very short distance there between the reactor vessel 16 and the containment vessel in that lower region.
17 MEMBER CORRADINI: Okay.
18 MEMBER BLEY: Mike, which figure did you 19 hold up?
20 MEMBER CORRADINI: Slide seven of his 21 packet we just had which is an engineering drawing.
22 MR. LASSITER: Correct, taken from an 23 engineering drawing.
24 MEMBER REMPE: I'm sure you'll talk about 25 it tomorrow, but do you have an idea of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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226 1 uncertainty in all the condensation heat transfer? I 2 mean do you think your numbers are accurate to plus or 3 minus 20 percent, 50 percent?
4 MR. LASSITER: That is a question for the 5 chapter 15 folks tomorrow.
6 MEMBER REMPE: Okay.
7 MEMBER CORRADINI: They're taking a lot of 8 penalties that a best estimate wouldn't take. That, 9 at least from what we've seen in the PRA calculation 10 when we went to NuScale at the offices, I see a lot of 11 penalties.
12 MEMBER BLEY: Can you go back to seven?
13 You can just it a seven and it will jump there. Yes, 14 at least you learned one thing today.
15 (Laughter.)
16 MEMBER BLEY: Since this is a better 17 rendering and I did like it because the picture is a 18 little different, just above the reactor core out 19 inside in the containment it's kind of something 20 sticking out, annular ring around the reactor vessel 21 down lower. Right there. What is that?
22 MR. LASSITER: This is a lower reactor 23 flange, so that's where the refueling --
24 MEMBER BLEY: Okay, those are vertical 25 bolts?
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227 1 MR. LASSITER: Correct.
2 MEMBER BLEY: That I'm seeing in there.
3 Okay. So that's where you open it up to get down to 4 front. Okay. And some of the other pictures look 5 like a bit flow restricter, but on this picture it's 6 essentially maintaining the same annulus that you have 7 down below it it looks like. So I probably doesn't 8 figure in very much.
9 MEMBER CORRADINI: Since we're on that is 10 the RRV above or below that flange? It's above the 11 flange. I don't see it in the cartoon.
12 MR. LASSITER: It's above the flange.
13 MEMBER CORRADINI: Okay. Thank you.
14 CO-CHAIR MARCH-LEUBA: I'm looking at 15 figure 15.6-62 about how much water is in the vessel.
16 And when you open the RRV and you let the cooling 17 rate, the inside and the outside, they reach ten feet 18 above the core, core top, to guard the fill.
19 MEMBER BLEY: Inside.
20 CO-CHAIR MARCH-LEUBA: Yes, approximately, 21 because if it did have flow there will be a one foot 22 difference between left and right.
23 Roughly, the cooling units are about ten 24 feet above the part.
25 MR. LASSITER: Okay, next slide here.
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228 1 So only two vent valves and one 2 recirculation valve is required for operation, at 3 least in chapter 16 analyses. Core stays covered for 4 all design basis events and ECCS, as we have 5 discussed, is automatically actuated when required by 6 the module protection system, based on the high 7 containment level where loss of power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
8 So operator action is not required. However, 9 operators can manually actuate ECCS from the control 10 room. The ECCS valves form part of the reactor 11 coolant pressure boundaries, so they're designed and 12 fabricated in accordance with class 1 as ASME class 1 13 components.
14 CO-CHAIR MARCH-LEUBA: If an ECCS valve 15 fails to open, is there anything the operator can do 16 to push it, to drive it to position?
17 In the cartoons that you show, no. The 18 answer is no.
19 MR. LASSITER: No, they are instructed to 20 not wait until there's an active ECCS demand before 21 they attempt an override, but physically, you know, 22 it's not turning a knob, turn something harder in 23 valve --
24 CO-CHAIR MARCH-LEUBA: Other than going 25 there with some scissors and cut the power to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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229 1 solenoids, he cannot drive it in with a wrench.
2 MR. LASSITER: No.
3 CO-CHAIR MARCH-LEUBA: Because the reason 4 it would fail would be you have corrosion inside 5 stopping the motion, failing of two solenoids would 6 de-energize it is unbelievable.
7 MR. LASSITER: Okay, and we haven't talked 8 about this yet, but an auxiliary function of the ECCS 9 is to provide low temperature over pressure protection 10 for the reactor vessel in the start up portion or I 11 guess I should say in the low pressure start up or 12 shut down of the module. And there's a --
13 MEMBER BLEY: You don't have a brittle 14 failure problem, no. That's when people used to use 15 LTOP all the time.
16 MR. LASSITER: It precludes the module 17 from getting into that region.
18 MEMBER BLEY: Yes.
19 MR. LASSITER: So there's a pressure 20 temperature curve built into the module protection 21 system logic accordingly to prevent over 22 pressurization at low temperature.
23 MEMBER BLEY: So it is for brittle failure 24 protection.
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230 1 generous curve, but it still is at pressure 2 temperature.
3 MEMBER BLEY: I think that would be very 4 generous. I mean it seems to me that it would 5 probably generous to the point you didn't even need it 6 except for operational reasons where you might not 7 want to over prep.
8 MR. LASSITER: The curve for that is in 9 chapter 5. I don't have the table off the top of my 10 head.
11 MEMBER BLEY: Table 5.
12 MR. LASSITER: Next slide. This is a 13 simplified diagram of the ECCS valves. We're talking 14 about valve operation and configuration. This figure 15 in particular looks more like the recirculation valve 16 on the side of the reactor vessel.
17 We have -- I'll point with the mouse here.
18 We have the main valve attached to the reactor vessel.
19 We have the inadvertent actuation block device which 20 has been discussed previously attached to the main 21 valve or as part of the main valve design. We'll see 22 the detailed drawings.
23 MEMBER BLEY: It's actually a separate 24 valve, at least it looks like from the cartoon in 25 chapter -- we'll see that later. I said separate NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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231 1 within a housing it's a separate valve.
2 MR. LASSITER: Correct. There's a trip 3 line tubing going through the containment region and 4 the trip reset actuator assembly which is attached to 5 the containment vessel wall which was discussed in 6 chapter 3 at some extent.
7 MEMBER CORRADINI: So is the reason you 8 need an outside containment for maintenance and 9 servicing?
10 MR. LASSITER: Primarily it's for solenoid 11 cooling.
12 MEMBER CORRADINI: Oh, it is.
13 MR. LASSITER: So we would either need to 14 bring cooling water into the vacuum region of 15 containment or put them on the outside and submerge in 16 the reactor pool to provide cooling to the solenoids 17 which are --- especially for the trip valve which is 18 normally energized to stay closed.
19 MEMBER CORRADINI: So all five have to be 20 sitting inside the swimming pool.
21 MEMBER BLEY: Wait a minute, all five of 22 those pilot valve assemblies.
23 MR. LASSITER: Correct.
24 MEMBER CORRADINI: All five of the things 25 that are S and S. IABs are inside. S and S are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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232 1 outside. P is with a real valve.
2 MEMBER BLEY: IAB that's not solenoid 3 operated?
4 MEMBER CORRADINI: No, it's simply a 5 pressure differential measure. You pressurize it 6 closed and then you wait for a pressure to allow to 7 open, a pressure difference --
8 MEMBER BLEY: I'm showing what it 9 accomplishes.
10 MEMBER CORRADINI: I'm showing what it 11 accomplishes. You're asking me how a valve actually 12 works. It's something different.
13 MR. LASSITER: Yes, we'll look at the cut 14 away in the closed session. I think that will be much 15 more clear.
16 Next slide.
17 MEMBER BLEY: So is it just air cooled 18 outside of containment?
19 MR. LASSITER: Water, because they're 20 submerged in the reactor pool.
21 MEMBER BLEY: Oh, they're in the pool.
22 That's right. Thank you.
23 MEMBER CORRADINI: What is the failure 24 that you're concerned about that you can't have it?
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233 1 materials? I guess I don't understand.
2 MR. LASSITER: It's longevity of the 3 solenoids for operational purposes.
4 MEMBER BLEY: They just keep heating up in 5 a vacuum.
6 CO-CHAIR MARCH-LEUBA: If it's in a 7 vacuum, you can only cool them by radiation.
8 MR. LASSITER: It would be too hot.
9 MEMBER BLEY: They get really hot.
10 Typically, a solenoid when it gets too hot, whatever 11 kind of insulation, whatever it is, it'll disappear.
12 MEMBER BALLINGER: You'll get gamma 13 heating as well.
14 MEMBER BLEY: Slightly more than I know.
15 MEMBER BALLINGER: And in a vacuum, you 16 get a heat record. There ain't nowhere for it to go.
17 MR. LASSITER: Okay, on to the next slide.
18 So as we talked about removing power to those solenoid 19 actuators, the trip valves are located out on the 20 containment vessel wall. That actuates the main valve 21 and the main valves operate using reactor coolant 22 pressure and they're also assisted in movement by a 23 spring, an internal spring.
24 MEMBER BLEY: So in our closed session, 25 I'd be interested in -- it states it the other way in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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234 1 chapter 6 is they're spring operated with some 2 hydraulic assist. The proportion of that would be of 3 some interest.
4 MR. LASSITER: The proportion is largely 5 hydraulic. The spring is a low-pressure spring.
6 MEMBER BLEY: So you really don't need the 7 spring.
8 MR. LASSITER: Around 30 psi, but we need 9 the spring to hold the valve open indefinitely. So 10 after the vessels have depressurized and there's no 11 driving hydraulic force to open the valve.
12 CO-CHAIR MARCH-LEUBA: I guess we'll go 13 through this maybe closed session or maybe tomorrow, 14 but if your level on the containment is high enough to 15 treat ECCS, your delta P is not that much.
16 MR. LASSITER: So it's event dependent as 17 to how much delta P is --
18 CO-CHAIR MARCH-LEUBA: Delta P will be 19 negligible. If you rely on that to be existing, I 20 think you went into the view of all the analysis more 21 accurately.
22 MR. LASSITER: I think we should defer 23 that question to --
24 (Simultaneous speaking.)
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235 1 disagree, but my worry is that the simplified using of 2 hydraulic evaporated valve data has given us a biased 3 estimate in the PRA.
4 CO-CHAIR MARCH-LEUBA: I cannot wait until 5 --
6 MEMBER BLEY: And over confidence, yes.
7 CO-CHAIR MARCH-LEUBA: I cannot wait to 8 see all of these hands.
9 MR. LASSITER: Okay, continue here. Each 10 ECCS main valve includes an inadvertent actuation 11 block device. You know the primary purpose of ECCS is 12 cooling in postulated accidents. But now due to the 13 design of ECCS, an inadvertent actuation block device 14 is engineered to prevent ECCS main valve opening due 15 to an unexpected failure, such as solenoid or power 16 failure while at operating conditions.
17 The inadvertent actuation block device is 18 not normally engaged and it's strictly a differential 19 pressure device and we'll see detailed figures in the 20 closed session and some diagrams of its operation, but 21 it does not come into play for any valid ECCS demands 22 in chapter 15 analyses. It stays open. It's only for 23 an unexpected failure during operational conditions.
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236 1 their normal position, right? And spring operated to 2 their failed, but desired safety position.
3 MR. LASSITER: That's correct. Well, even 4 the small valves are solenoid operated valves.
5 They're just on a much smaller scale, but there is a 6 spring. The spring assist is also there. They're 7 based on some --
8 MEMBER BLEY: So the solenoid is not 9 driving them in both directions.
10 MR. LASSITER: No, the solenoid is 11 energized to close, and fail to open which when the 12 fail to open is by spring and hydraulic actually.
13 MEMBER BLEY: On both?
14 MR. LASSITER: Even on the small ones.
15 MEMBER BLEY: Okay. You'll show us more 16 later.
17 MR. LASSITER: Yes.
18 MEMBER BLEY: I sure didn't get that.
19 Okay.
20 MR. LASSITER: Okay, let's go to the next 21 slide.
22 So another -- talk about the regeneration 23 in section 6.3. And the regeneration in the NuScale 24 containment is limited by restricting of insulation, 25 paint, and coatings in containment. There is a debris NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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237 1 effects evaluation performed, see section 6.3.3.1.
2 And the debris assumed in that evaluation is based on 3 latent debris from refueling or maintenance 4 activities. Therefore, we have a COL item for the 5 applicant to implement a cleanliness program to limit 6 debris within the containment and support the 7 assumption of that evaluation.
8 And the ECCS is capable of post-accident 9 extended long term cooling beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
10 MEMBER CORRADINI: So let's talk about 11 that for a minute. So I want to ask the staff as 12 well. I want to make sure. Have you responded to 13 their RAIs on potentially boron non-equilibrium 14 concentrations within the core and outside the -- in 15 the containment? 8930 is the RAIs that I have written 16 down to remind myself to ask about.
17 MR. LASSITER: I think that's being 18 addressed by our chapter 15 analysis team.
19 MEMBER CORRADINI: So you escape again.
20 But not for so long.
21 But I want to make sure because I thought 22 this is part of the requirement for the long term 23 cooling for ECCS actuation. Because if I get into a 24 situation that all goes well in the short term, I 25 eventually have the situation where I can essentially NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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238 1 create a -- not a disequilibrium, but a difference in 2 concentration inside the core and outside in 3 containment in terms of boron concentration. Am I not 4 correct?
5 CO-CHAIR MARCH-LEUBA: I think we can 6 discuss this at length tomorrow. So they can get the 7 24-hour relief.
8 MEMBER CORRADINI: Okay, fine.
9 CO-CHAIR MARCH-LEUBA: But we will discuss 10 at length tomorrow.
11 MR. LASSITER: But the long term cooling 12 topical report that's addressed --
13 MEMBER CORRADINI: The RAI that I was 14 tracking was 8930. I'm looking at staff, are they 15 nodding that I have the right RAI?
16 MS. KARAS: This is Becky Karas from 17 Reactor Systems. Yes, that's correct and that is 18 going to be described as part of the chapter 15 19 presentation.
20 MEMBER CORRADINI: Okay, so this is the 21 one that you guys will not talk about today that 22 you'll talk about tomorrow. So everybody gets a day.
23 Great. Thank you. Move on.
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239 1 we'll move on to 6.4.
2 Switch gears to control room habitability 3 system. The control room habitability system provides 4 stored breathing air and pressurization to the control 5 room in the event that the normal control room 6 ventilation is unavailable. It's designed to support 7 20 personnel for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The air inventory is 8 monitored by pressure and temperature instrumentation.
9 The control room habitability system is automatically 10 actuated by the plant protection system or manually 11 inside the control room envelope.
12 CO-CHAIR MARCH-LEUBA: Just out of 13 curiosity how big are those times? I mean 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 14 worth of air for 20 guys.
15 MEMBER CORRADINI: A lot.
16 CO-CHAIR MARCH-LEUBA: Because I know when 17 you go diving you carry something like this and it 18 lasts you for an hour.
19 MR. LASSITER: There are -- I'm trying to 20 remember the quantity of tanks, but they are about 21 five feet by a foot in diameter. And there are excess 22 inventory to take some out of service.
23 CO-CHAIR MARCH-LEUBA: You have confidence 24 that's sufficient?
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240 1 a connection to offsite air supply source if needed 2 past 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. However, power is expected to be 3 available to the normal control room ventilation 4 system after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
5 MEMBER CORRADINI: The requirement is 72 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
7 MR. LASSITER: Correct.
8 MEMBER CORRADINI: Okay.
9 MR. LASSITER: Next slide. This is a 10 simplified diagram of the control room habitability 11 system. Air bottles providing the stored breathing 12 air to the control room envelope. There's 13 automatically actuated supply valves. There's a 14 manual valve as well and there are the -- I guess I 15 could point with the mouse -- and the balancing of 16 pressure relief valves. These are solenoid valves 17 which similarly relate to the ECCS and actually are 18 held in the closed position by energizing to close and 19 to fail to open.
20 DR. SCHULTZ: Is the control room normally 21 pressurized?
22 MR. LASSITER: The control room is 23 normally pressurized to 1/8th inch water gauge by the 24 normal control room ventilation system.
25 DR. SCHULTZ: Thank you.
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241 1 MR. LASSITER: And the control room 2 ventilation system operates in conjunction with the 3 normal control room ventilation system to close the 4 normal supply and return dampers which are shown at 5 the bottom of this figure here.
6 Go to the next slide.
7 Due to the fact that NuScale does not rely 8 on operator action for chapter 15 analyses or 9 postulated events, the control room habitability 10 system is classified as non-safety related. However, 11 augmented quality requirements are applied such as all 12 the components which store the air and supply the air 13 are seismic category one. And the COL item which 14 requires the applicant to periodically test and 15 inspect the integrity of the control room envelope as 16 well as the control room habitability components.
17 The control room habitability system is 18 automatically actuated on loss of AC power to both 19 normal control room ventilation, air handler units for 20 ten minutes; high radiation downstream of the normal 21 control room ventilation intake filtration train; or 22 loss of AC power to the four EDSS battery chargers.
23 Next slide.
24 Here are the COL items for section 6.4.
25 That concludes my presentation for section 6.4.
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242 1 MR. HOUGHTON: All right, so section 6.5 2 covers fission product removal and control. So we do 3 not have any credited fission product removal systems 4 other than the containment and other components that 5 we've discussed already.
6 We don't credit the reactor building or 7 any of the HVAC systems in our dose analyses and we do 8 not employ a containment spray system in the NuScale 9 design.
10 Section 6.6, in-service inspection and 11 testing covers ISI, in-service inspection program, for 12 class 2 and 3 components based on 10 CFR 50.55(a) 13 (g)(3). Pre-service inspections and in-service 14 inspections are performed on class 2 and 3 components 15 in accordance with boiler pressure vessel section 11 16 requirements and I'll add a clarification here. The 17 containment vessel inspections are performed for class 18 1 requirements and those are described in table 6.2-3.
19 The in-service inspection program is 20 discussed during the chapter 3 discussion this 21 morning. It includes augmented volumetric and surface 22 inspections to protect against postulated piping 23 failure areas. So that's specifically looking at the 24 high energy fluid system piping and break exclusions 25 on piping. You'll see those augmented inspections.
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243 1 There are COL items for development of the 2 actual programs by the applicant. And that concludes 3 the presentation.
4 MEMBER CORRADINI: Questions by the 5 members?
6 Okay, no questions. Before we have the 7 staff come up with chapter 6, thanks to NuScale.
8 Let's take a break. We'll be back at 20 til.
9 (Whereupon, the above-entitled matter went 10 off the record at 2:26 p.m. and resumed at 2:40 p.m.)
11 MEMBER CORRADINI: Okay, let's get back in 12 session.
13 Omid, you're up.
14 MR. TABATABAI: Yes. Good afternoon. My 15 name is Omid Tabatabai. I'm a senior project manager 16 in the Office of New Reactors and I just want to first 17 thank the members for giving us an opportunity to 18 present to you our draft phase 2 SER for chapter 6, 19 engineered safety features.
20 The reason I mention draft phase 2 is 21 because we still have some open issues that we're 22 still working with NuScale to find resolutions to. So 23 they're not quite open items as we are used to 24 defining them that we haven't mutually agreed --
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244 1 items.
2 MR. TABATABAI: Yes.
3 MEMBER CORRADINI: Uncommon.
4 MR. TABATABAI: Uncommon open items.
5 MEMBER CORRADINI: So I counted three 6 uncommon open items in chapter 6 that I want to make 7 sure I understand. So when we get to the right point 8 I want to make sure we understand them.
9 MR. TABATABAI: Sure.
10 MEMBER CORRADINI: Thank you.
11 MR. TABATABAI: Thank you. Staff's agenda 12 today, I'm just going to recognize the team who worked 13 on chapter 6 for the staff provide an overview of the 14 NRC staff's presentation and because we want to focus 15 this presentation on chapters -- I'm sorry, 16 subsections that have open items or exemption 17 requests. So we'll focus only on section 64.5.6, 6.3 18 and 6.4. So we don't have really because of the time 19 and important issues that will be discussed in chapter 20 3 and chapter 6 issues, we're not going touch on 6.1, 21 material issues, or other things that we don't have 22 any open items for.
23 This is a list of NRC staff reviewers who 24 were heavily involved in the review of chapter 6.
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245 1 are presenting to you today.
2 As I mentioned earlier, we won't focus on 3 section 6.2 which is containment systems; 6.3, 4 emergency core cooling system; and control room 5 habitability. These are the sections that either have 6 open items or exemption requests in them.
7 As we have already informed the members, 8 we won't discuss 6.2.1 containment structures, 6.2.2, 9 containment heat removal systems as part of chapter 15 10 presentation later this week. And during NuScale's 11 presentation, it required a few questions with respect 12 to these sections that were deferred to to tomorrow 13 and Thursday's presentation.
14 MEMBER CORRADINI: So let me stop you here 15 and make sure I understand. So today, we're not going 16 to hear about 6.2.1 nor 6.2.2?
17 MR. TABATABAI: That's correct.
18 MEMBER CORRADINI: But the uncommon open 19 items are part of 6.2 --
20 MR. TABATABAI: 6.2.1.
21 MEMBER CORRADINI: 6.2.1. I'm sorry.
22 MR. TABATABAI: That's correct.
23 MEMBER CORRADINI: 6.2.1. I get all my 24 subsections mixed up. Are we even going to discuss 25 those today? Are we going to be discussing those NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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246 1 tomorrow?
2 MR. TABATABAI: They will be discussed 3 tomorrow.
4 MEMBER CORRADINI: Okay. The reason I'm 5 asking is I read in two places in there that the staff 6 has done audit calculations and I'm curious about are 7 we going to see the audit calculations and what tools 8 were used with the audit calculations. So that is 9 planned for tomorrow.
10 MR. TABATABAI: That's correct.
11 MEMBER CORRADINI: Okay, good.
12 MR. TABATABAI: And also, I just want to 13 make sure you know that there are some open items, 14 uncommon open items for today's presentation as part 15 of chapter 6 which Shanlai will be talk about later 16 on.
17 MEMBER CORRADINI: Okay. Thank you.
18 MR. TABATABAI: Just for your information, 19 we issued about 34 requests for additional information 20 which included 114 questions. And we have received 21 about 111 of those questions responded to as of today.
22 So with that, I'm going to turn the 23 microphone to Clint Ashley to discuss 6.2.4.
24 Clint?
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247 1 afternoon. As Omid indicated, my name is Clint 2 Ashley. And I'll be presenting the staff's review of 3 section 6.2.4, containment isolation. There are 4 actually no open items in this review area, but there 5 are four containment isolation related requirements 6 where NuScale seeks an exemption.
7 MEMBER CORRADINI: You have to speak up, 8 Clint. Even though you've got your green thing on.
9 We want to hear your words.
10 MR. ASHLEY: So this slide just contains 11 a containment isolation regulatory basis and related 12 requirements. I don't intend to read these aloud.
13 They're just listed here for information purposes.
14 And in an upcoming slide, I'll talk 15 specifically about those requirements where NuScale 16 requests an exemption.
17 This slide shows the review guidance 18 applicable to containment isolation. Again, I just 19 show this for information purposes. And I don't 20 intend to read these aloud.
21 Next slide, please.
22 So NuScale's DCA contains several 23 exemption requests associated with containment 24 isolation requirements. They were touched on earlier 25 by Zack Houghton from NuScale.
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248 1 Just to remind each of us about what these 2 requirements are, for GDC 55, the provisions require 3 in part that each line that is part of the reactor 4 coolant pressure boundary and penetrates primary 5 reactor containment shall be provided with isolation 6 valves.
7 GDC 55 specifies the location of the 8 valves with one inside and one outside containment.
9 And you can think of the systems that traditionally 10 fall within GDC 55 are like the CVCS system.
11 The provisions in GDC 56 are nearly 12 identical to GDC 55 except that GDC 56 applies each 13 line that is connected directly to the containment 14 atmosphere. And for both of these GDCs, redundant 15 barriers are required to account for a single act of 16 failure in the isolation provisions. And this is 17 typically achieved by providing two barriers, two 18 isolation valves in series.
19 NuScale's GDC 55 and 56, piping 20 penetration lines, provide a containment isolation 21 design consisting of two automatic isolation valves 22 located outside containment in series and therefore 23 the applicant requests an exemption request from the 24 requirements of GDC 55 and 56 just regarding valve 25 location.
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249 1 MEMBER CORRADINI: Historically, have 2 there been any containments with outside isolation 3 valves such as this?
4 MR. ASHLEY: There have been exemption 5 requests for having isolation valves located both 6 outside.
7 MEMBER CORRADINI: Okay, so that's 8 happened in the past?
9 MR. ASHLEY: It has.
10 MEMBER CORRADINI: Okay. And you approved 11 them?
12 MR. ASHLEY: I'm sorry?
13 MEMBER CORRADINI: And the staff approved 14 them?
15 MR. ASHLEY: The Agency at the time 16 approved them, that's correct.
17 MEMBER CORRADINI: I assumed that. I 18 should have asked that. Thank you.
19 CO-CHAIR RICCARDELLA: But there are 20 exemptions and they're limited, right?
21 MR. ASHLEY: They are. They are limited.
22 And we'll talk more about that in the next few slides.
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250 1 boundary or connects directly to the containment 2 atmosphere. So we're looking at things like confluent 3 cooling water, cooling systems, those things that 4 aren't directly tied in the reactor coolant pressure 5 boundary or the containment atmosphere. Those 6 typically just have at least one isolation valve. And 7 as specified in GDC 57, the valve is located typically 8 outside containment.
9 In NuScale there are two independent decay 10 heat removal system trains, each with a decay heat 11 removal system steam supply line and a decay heat 12 removal system condensate return line. And the 13 applicant poses the use of closed loop decay heat 14 removal system outside containment as an alternative 15 to the isolation valve requirement. So therefore, 16 because of the difference, they've sought an exemption 17 request for that particular requirement.
18 And the last item on the slide is a TMI-19 related requirement. It's 10 CFR 50.34(f)(2)(xiv)(E).
20 And it requires containment isolation systems that 21 include automatic closing on a high radiation signal 22 for all systems that provide a task to the 23 environment.
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251 1 the environment, but they don't use the high radiation 2 signals. And so as a result, the applicant requests 3 an exemption from the TMI-related requirement.
4 Next slide, please.
5 The purpose of this slide is to identify 6 that the Commission will not consider granting an 7 exemption unless special circumstances are present.
8 Special circumstances in part are present whenever 9 application of the regulation in the particular 10 circumstances would not serve the underlying purpose 11 of the rule or is not necessary to achieve the 12 underlying purpose of the rule. For the exemptions 13 that we just discussed on the prior slide, NuScale's 14 position is that special circumstances are present in 15 that the underlying purpose of the requirement or rule 16 is achieved through alternate containment isolation 17 provisions. And we'll talk more about how NuScale 18 meets the underlying purpose of the requirement due to 19 the alternate needs on the next few slides.
20 So just to tee up in order to assess 21 whether special circumstances exist, it's important to 22 understand the underlying purpose of the requirements.
23 For GDC 55, 56, and 57, the underlying purpose is to 24 provide containment isolation capability that supports 25 the safety function of containment, to provide a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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252 1 barrier to the uncontrolled release of radioactivity 2 to the environment. And generally, this requires a 3 redundant isolation barrier such as achieved by either 4 a valve or a closed system.
5 The underlying purpose of the TMI-related 6 requirement is to limit radiological releases by 7 ensuring containment isolation for systems that 8 provide a task to the environment. And this is most 9 notably associated with NuScale's containment 10 evacuation system.
11 Okay. Two more slides. This slide 12 summarizes the staff finding for GDC 55, 56, and 57.
13 And I'd like to emphasize that the design specific 14 review standards section 6.2.4 does discuss adding two 15 isolation valves outside containment. And I think 16 Zack Houghton from NuScale teed up this important 17 issue about the uniqueness of the NuScale's 18 containment design, the harsh environment inside 19 containment.
20 And if you were going to have these two 21 isolation valves outside the design specific review 22 standards speaks to basically assessing that region 23 between the containment and the first outboard 24 isolation valve. And you want to look at that region 25 with the lens of standard review plan 3.6.2 and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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253 1 associated Branch Technical Position 3-4 and adopt 2 this break exclusion zone. So that's why GDC 55, 56, 3 and 57 have satisfied the alternate containment 4 isolation position.
5 Next slide, please.
6 For 50.34(f)(2)(xiv)(E), the NuScale 7 applicant requests an exemption from this requirement 8 as applied to the containment evacuation system. In 9 the DCA, the applicant describes the NuScale design 10 meets the underlying purpose of the rule by isolating 11 the containment evacuation system using two automatic 12 containment isolation signals. The first one, not 13 necessarily in this order, but one would be a 14 containment vessel pressure signal and the second one 15 would be low pressurize the level. And that's not all 16 the signals that are available, but these are the two 17 that would be called upon in the event of trying to 18 contain that radiation or radiological release inside 19 containment.
20 So the applicant goes on to explain that 21 the NuScale design differs from the traditional large 22 light water reactors designs because reactor core 23 uncovery and the resulting core damage cannot occur 24 without reaching a low pressurized level contained in 25 isolation set points. Therefore, an event similar to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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254 1 TMI Unit 2 accident is precluded by the NuScale plant 2 design.
3 The applicant also states that in the 4 NuScale plant design the pressurizer is located well 5 above the level of the reactor core, not connected to 6 the reactor vessel by piping. And any decrease in 7 reactor vessel and inventory to the level of the core 8 would result in complete emptying of the pressurizer 9 and initiation of the pressurizer level contained in 10 isolation signal. And as such, the applicant 11 describes the automatic isolation of the containment 12 evacuation system on a high radiation signal is not 13 required to meet the underlying purpose of the rule, 14 in part, mainly in part, because there's alternate 15 means that would preclude that release of radiation to 16 the environment.
17 But based on that information and the 18 unique NuScale design, and our review of the TMI Unit 19 2 related events, we felt very comfortable that 20 alternate means are provided to prevent radiological 21 releases to the environment and in this case, 22 automatic isolation on a high radiation signal is not 23 required to meet the underlying purpose of the rule.
24 For both of these exemption requests, the 25 staff finds that NuScale's exemption meets the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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255 1 requirements as defined in 50.6. That concludes my 2 presentation.
3 MEMBER SUNSERI: So just -- I mean maybe 4 I'm reading more into it, but you say it meets the 5 requirements for an exemption, but are you accepting 6 it?
7 MR. ASHLEY: Well, I think it's ultimately 8 the Commission. I certainly recommend it. They meet 9 the requirements.
10 MEMBER SUNSERI: Okay. I get it. All 11 right. Thanks. But technically, there's nothing.
12 MR. ASHLEY: Yes.
13 MEMBER SUNSERI: Okay.
14 MS. GRADY: Good afternoon. I'm Anne-15 Marie Grady and I'm the reviewer for combustible gas 16 control and also for the next presentation containment 17 grate testing.
18 The regulatory basis for -- I should start 19 off by saying this section combustible gas control has 20 one exemption request which is what I'm going to talk 21 about now and it also has one open item.
22 The regulatory basis for combustible gas 23 control is to have -- is to meet GDC 41, 42, and 43 of 24 being able to have atmospheric clean up and in section 25 4, of combustible gas control system and also to meet NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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256 1 10 CFR 50.44(c). 50.54(c) has five aspects that it 2 controls. It requires a mixed containment atmosphere.
3 It requires that hydrogen be controlled below ten 4 percent for severe accident. It requires equipment 5 survivability. It requires containment, hydrogen, and 6 oxygen monitoring post-accident. And it also requires 7 containment integrity post-accident.
8 And I should also emphasize because this 9 has been -- this has changed in the early 2000s, we're 10 only talking about a severe accident. This regulation 11 used to apply to the design basis accident. Now it's 12 an accident basically that can have 100 percent 13 zirconium oxidation and release all of the hydrogen 14 that it does produce.
15 Some key design features to evaluate in 16 looking at the exemption request is that the NuScale 17 containment is kept at a very low pressure, sub-18 atmospheric, during normal operation which severely 19 limits the amount of actual hydrogen -- I'm sorry, 20 actual oxygen in the containment which is a key 21 factor.
22 NuScale's containment is designed to 23 accommodate a bounding combustion event resulting from 24 hydrogen generation at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without a loss of 25 integrity or loss of supporting system structures and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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257 1 components.
2 NuScale has asked, requested an exemption 3 to not have a system to control the hydrogen in the 4 containment and that is primarily because the way that 5 this plant is operated and it really is going to be an 6 oxygen-starved environment, so even though there's 7 significant hydrogen produced from the oxidation of 8 the zirconium, there's barely ever enough oxygen to 9 support combustion. So it almost behaves as if it 10 were a BWR, although certainly not described as such.
11 So it's oxygen starved. There's sufficient hydrogen, 12 but so what?
13 Now NuScale would say that they don't need 14 a system to control that because they can, in fact, 15 meet the conditions of the combustion and/or 16 detonation in the containment within the first 72 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> considering that there is barely enough oxygen 18 to support the combustion and/or detonation.
19 Next slide, please.
20 We focused on the fact that combustible 21 gas control does meet design of NuScale, does meet the 22 other aspects required by 50.44(c). It does have a 23 mixed containment environment. It does provide 24 hydrogen and oxygen monitoring. I'll get to that in 25 a minute. And it does have equipment survivability NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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258 1 program and the containment integrity is maintained 2 even with a bounding accident within the first 72 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
4 Therefore, staff recommends that the 5 exemption request be granted because a system to 6 control the hydrogen in the containment is not 7 necessary within the combustion and/or detonation 8 would not be supported.
9 CO-CHAIR MARCH-LEUBA: If you have the 10 right LOCA all that 100 percent hydrogen mass can move 11 to outside of the containment into the main building.
12 What would be the concentration of hydrogen there?
13 MS. GRADY: The analysis that was done, 14 the calculations that were done for this design were 15 for LOCA with an intact containment.
16 CO-CHAIR MARCH-LEUBA: So the LOCA 17 bypasses containment?
18 MS. GRADY: No. Intact.
19 CO-CHAIR MARCH-LEUBA: Or intact 20 containment.
21 MS. GRADY: Intact. So there is nothing 22 leaving the containment except what's --
23 CO-CHAIR MARCH-LEUBA: Are there some 24 circumstances where highly can move outside 25 containment, what would happen then?
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259 1 MS. GRADY: There has been an accident 2 analyzed in chapter 19 which was a bypass accident, 3 but I'm not talking about that today. I'm talking 4 about only the LOCAs that could take place in inside 5 containment, containment -- unless the combustion or 6 detonation failed the containment. We're talking 7 about an intact containment.
8 CO-CHAIR MARCH-LEUBA: We are asking for 9 an exception to the rule. I'm asking is not the 10 equivalent to this rule for the particular conditions 11 of NuScale hydrogen on the pool building instead of 12 hydrogen in containment?
13 MS. GRADY: I'm sorry, Dr. March-Leuba, 14 but I don't understand your question.
15 MR. LU: Actually, I can help a little 16 bit. Based on LOCA analysis, at this point there is 17 no issue with the core recovered and then based on 18 what was presented to us. It is unlikely to have that 19 the design base, the LOCA generated --
20 CO-CHAIR MARCH-LEUBA: What you're saying 21 --
22 MR. LU: Chapter 19, you're correct.
23 CO-CHAIR MARCH-LEUBA: You're saying there 24 is no way to get that much hydrogen anyway.
25 MS. GRADY: Well, that's --
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260 1 MR. LU: For design basis. Yes.
2 MS. GRADY: From the perspective of 3 Chapter 15, he's absolutely right. But that has no 4 bearing on combustible gas control regulation.
5 Because it says, you shall model an accident with 100 6 percent zirconium oxidation, and produce the hydrogen 7 that would be produced.
8 That's the requirement. It's not a 9 deterministic accident.
10 CO-CHAIR MARCH-LEUBA: The requirement 11 says that you should produce 100 percent hydrogen and 12 put it in containment. That's what the requirement 13 says.
14 MS. GRADY: It says model that --
15 (Simultaneous speaking.)
16 CO-CHAIR MARCH-LEUBA: I'm asking what are 17 they thinking -- if they had been thinking about 18 NuScale, would it have had a requirement that says 19 also generate 100 percent hydrogen and put it on the 20 building?
21 MEMBER CORRADINI: No. I'll say no.
22 That's beyond the design base. That's a Chapter 19 23 question.
24 CO-CHAIR MARCH-LEUBA: I'm thinking of a 25 picture that is what everybody's doing for Fukushima, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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261 1 of a place that was not supposed to have hydrogen 2 blowing up.
3 MEMBER CORRADINI: But with current 4 operating plants, that's not a requirement. So it 5 would seem -- I'm going to speak only for myself.
6 That would seem inappropriate for asking this of 7 NuScale if I'm not asking this for current plants.
8 MS. GRADY: I know what you're asking now.
9 Combustible gas control is combustible gas control in 10 containment, that's the regulation. Now if you're 11 asking me what might occur --
12 MEMBER CORRADINI: But I think where 13 Jose's going to is what you talked to us about a month 14 ago --
15 MS. GRADY: Yes.
16 MEMBER CORRADINI: -- which is 17 nevertheless, you are still worried about hydrogen 18 control within the bioshield relative to --
19 MS. GRADY: Absolutely.
20 MEMBER CORRADINI: -- a hydrogen control 21 issue.
22 MS. GRADY: Yes.
23 MEMBER CORRADINI: Even though it's not a 24 Chapter 15 issue, it's still an issue.
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262 1 evaluated in Chapter 19.
2 CO-CHAIR MARCH-LEUBA: And the bioshield 3 was modified to move the hydrogen into the building.
4 What I'm asking is, what will be the concentration --
5 that's a very big building. Is the concentration 6 there that undetonable?
7 (No audible response.)
8 CO-CHAIR MARCH-LEUBA: You don't know.
9 MS. GRADY: It is a big building, and 10 that's probably a saving grace because there's not 11 that much hydrogen. And there's a significantly large 12 atmosphere. So the concentration would not be a 13 problem in the building.
14 CO-CHAIR MARCH-LEUBA: I don't think 15 reaching concentration is a problem. But the better 16 answer is to say, even if you put 100 percent hydrogen 17 in the building, nothing happens. So I'm saying that 18 doesn't apply.
19 MS. GRADY: It would be a low 20 concentration of hydrogen and probably would not lead 21 to combustion.
22 CO-CHAIR MARCH-LEUBA: I'll be very 23 surprised that it does.
24 MEMBER CORRADINI: I want to take us back 25 to -- I want to make sure we're all on the same page.
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263 1 So within containment, I think Anne-Marie has answered 2 the question. But I think where Dr. March-Leuba is 3 going is, we still want to somehow get back to the 4 staff's eventual review --
5 MS. GRADY: Yes.
6 MEMBER CORRADINI: -- findings on the 7 appropriateness of the bioshield and any sort of 8 hydrogen concentrations within the bioshield above the 9 container.
10 MS. GRADY: That's correct.
11 MEMBER CORRADINI: And that's yet to be --
12 that's in the process of review.
13 MS. GRADY: Right. We discussed it 14 preliminarily when we were here with Chapter 19. We 15 will be answering the question in Phase 4, yes. But 16 that's a specific event identified by NuScale.
17 DR. SCHULTZ: As you do that are you going 18 to go further than the region of the bioshield, 19 because the design of the bioshield reportedly is 20 going to take care of that concentration possibility.
21 Are there other areas that we ought to be concerned 22 about?
23 MS. GRADY: Well, looking at that from the 24 perspective of multi-module effects, that whether or 25 not there could be combustion under the bioshield NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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264 1 and/or detonation under the bioshield, could lead to 2 impacting another module. That's the scope of the 3 review.
4 DR. SCHULTZ: Thank you.
5 MS. GRADY: Okay, now the open item. One 6 of the -- as I mentioned earlier, one of the 7 requirements for 50.44(c) is to have hydrogen and 8 oxygen monitoring. And NuScale originally intended to 9 provide the hydrogen and oxygen monitoring in 10 containment post-accident by using an aspect of the 11 sampling system, the post-accident sampling system.
12 Now, the sampling system is an online 13 system during normal operation. It runs continuously.
14 It records in the main control room. It tells the 15 operators what's going on inside the containment 16 atmosphere. That system can be, in fact, reinstated 17 after an accident.
18 And in this case, we have to postulate an 19 accident, 50.44(c), that can be reinstated by -- after 20 they have a LOCA, you have the containment isolating, 21 the isolation valves close.
22 In some time, and it doesn't have to be 23 quickly. In some time, within the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the 24 containment isolation valves could be opened up again 25 in the containment evacuation system and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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265 1 containment flooding and drain system and establish 2 that flow path again to sample the containment 3 atmosphere. And that was how NuScale proposed to meet 4 the requirements of 50.44(c)(4). So they would be 5 monitoring the containment atmosphere post-accident.
6 Now NuScale has come in, as you know, with 7 an accident source term topical report, which has 8 associated with it a post-accident sampling exemption 9 request. And they have, as of today, requested that 10 they be able to not meet the dosage requirements for 11 establishing this flow path, this hydrogen and oxygen 12 monitoring.
13 So we are discussing how this function, 14 monitoring hydrogen and oxygen in containment 15 post-accident, can be reinstated safely. That is 16 under evaluation now, there's been an RAI issued 17 asking them to provide the information to us that this 18 can be done safely, and we're waiting for the 19 response. And their RAI number is 9682, and that's --
20 so it's an open item.
21 MEMBER CORRADINI: That's one of the three 22 yet to be responded to in terms of the tallying that 23 Omid showed at the beginning. That still --
24 MS. GRADY: That has not been responded to 25 yet, but yes.
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266 1 MEMBER CORRADINI: Okay. Thank you.
2 MEMBER SKILLMAN: Anne-Marie, I find the 3 wording in that write-up at your fourth bullet 4 interesting. RAI has been issued to determined that 5 this configuration could be safely established.
6 Why isn't the word, that, if?
7 MS. GRADY: Oh, I don't know. I guess it 8 was arbitrary.
9 MEMBER SKILLMAN: I'll give you four 10 inches of lead on each one of those pipes, so maybe 11 six based on real experience. I don't think you want 12 to be drawing a sample after DBA unless you have 13 really shielded all the piping, all the tubing, and 14 everywhere that tubing and piping might go. You do 15 not want to tap into that containment if you've had an 16 accident. That's all I'm going to say.
17 MS. GRADY: Okay, I will take that under 18 consideration in the review of the response.
19 MEMBER SKILLMAN: There's real operating 20 experience on that.
21 MS. GRADY: Hence, the question has asked 22 for a dose analysis to, in fact, establish this 23 configuration. We are mindful of that but thank you.
24 MR. TABATABAI: I tried to actually 25 quickly change, that, with, if, but I forgot this is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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267 1 a PDF file so I couldn't do it, sorry.
2 MEMBER CORRADINI: Nicely done.
3 MS. GRADY: I won't forget.
4 Okay, so this is containment leakage 5 testing, and this is another topic that NuScale has an 6 exemption request. And the exemption request is that, 7 as you know, containment leak rate testing is 8 required. It's required by Appendix J to do the 9 integrated leak rate test on the containment, which is 10 Type A testing.
11 It's required to do the local mechanical 12 joints testing on the containment which are the Type 13 B testing, the piping -- the flange openings and also 14 the electrical penetration assemblies and the hatches.
15 And it also requires Type C testing which is on the 16 containment penetrations, which is essentially testing 17 the containment isolation valves.
18 NuScale proposes to test the containment 19 isolation valves. The Type C testing, as required by 20 Appendix J, proposes to do Type B testing on the 21 mechanical joints, which is the flanges and the 22 electrical penetration assemblies.
23 But NuScale has requested not to do a Type 24 A test, which is typically a test on the entire 25 vessel, which would be a pneumatic test, and NuScale NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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268 1 has several reasons for requesting not to do it. But 2 primarily, their basic reasoning is that they would --
3 could meet the underlying reason or goal of 4 demonstrating the total containment leakage by other 5 means.
6 And the other means are that they would, 7 first of all, have met -- that this is an ASME, 8 Section 3, Class 1 vessel, and all that that implies.
9 That they would be doing inservice inspection on this 10 vessel, which is significantly different from the 11 existing plants because they -- the vessel is 12 inspectable both inside and outside at 100 percent of 13 its surface, which is different from the plants that 14 tend to rely on the Type A testing for results.
15 That they -- let's see, they are -- they 16 have also -- the ASME code requires that they do a 17 hydrostatic test and they will do that as a matter of 18 design, but they have also promised -- proposed they 19 do a pre-service design pressure test. And that 20 really is a pressure test that has more stringent 21 requirements than the ASME hydrostatic test because 22 the ASME hydrostatic test on the vessel would in fact 23 allowed leakage at the mechanical joints on the 24 containment, and that would still pass the test.
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269 1 test, as a material flaw test, as anything like that 2 on the entire vessel, and not so much focusing on 3 that. So the pre-service design pressure test that 4 NuScale proposes to perform, once for each containment 5 vessel, is to have a hydrostatic test as the 6 hydrostatic test would be per the code, hold the 7 pressure for 30 minutes, and then if there is any 8 leakage -- not just some, any leakage at all, the 9 vessel would not pass the pre-service design pressure 10 test. It would be an indication that the joints 11 weren't tight enough, and they would have to go back 12 and evaluate what that really means.
13 Okay, that test is a NuScale proposal. We 14 have accepted that idea, and it's a test that would be 15 ITAACed as well separate from any other ITAAC that is 16 involved in this vessel.
17 Now, significant time on NuScale's part, 18 I'm sure, certainly on mine was to what you evaluate 19 the design of the mechanical joints. As many of you 20 people know here, the typical leakage -- the 21 preponderance of the leakage from a containment is 22 from the penetrations and the mechanical joints.
23 NuScale has done an analysis of their mechanical 24 joints per the ASME code.
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270 1 so that the flanges would be tightened down 2 sufficiently. They have established what the -- using 3 ANSYS and a finite element analysis. They have 4 established what the minimum bolt preloads would be 5 for all these mechanical joints.
6 Then they have -- recognizing the fact 7 that -- obviously, a phenomenon called prying that 8 could occur. I mean, these flanges have many bolts.
9 And if one were to fail, the pressure -- the internal 10 pressure could, in fact, cause the increase in 11 pressure to the adjacent bolts, and before you know it 12 you've unzipped the flange. So they haven't stopped 13 with the minimum preload per the ASME code, they've 14 also determined what the maximum preload would be for 15 the bolts they have designed and the materials they 16 have chosen.
17 And then establishing a bolt preload for 18 each of the different joints, and they're all 19 different. The preloads are all different.
20 And I should say one thing about the 21 mechanical joints. There are circular closures. They 22 have two concentric O-rings with mechanic -- with 23 metal seals that are set in grooves. They're all 24 specified individually because they are different 25 sizes. They bolt preloads are all different.
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271 1 And then they did a finite element 2 analysis on all of these mechanical joints to 3 determine what the gap would be after imposing 4 pressure -- internal pressure, which would represent 5 two different LOCAs in the containment.
6 And one of them was -- one of the 7 scenarios they analyzed just from -- well, on 8 sensitivity -- one of them was the RRV inadvertent 9 actuation, and the other one was the CVCS line 10 breaking side containment. And they got pressures and 11 temperatures resulting from those accidents. They 12 analyzed the joints.
13 They calculated what the gaps would be on 14 the flanges, and they showed flange by flange that the 15 gaps would be very small. They would be very small 16 typically on the inside of the flange, but the gap 17 would not proceed past the first seal in the flange 18 joint and progress to the outer one.
19 So, in other words, the design tells them 20 that under these pressure and temperature conditions 21 that the flanges would not open, would not leak.
22 That's what they say. And that is why they have 23 proposed to have the pre-service design pressure test 24 to confirm what they've already analyzed and shown in 25 their analysis.
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272 1 Since these features -- the features of 2 the fact that this vessel is an ASME design since they 3 have done an analysis that shows that leakage is 4 entirely unlikely. Since the other types of testing 5 -- types being types and testing will still be done 6 pneumatically per Appendix J.
7 And since -- I think that's the only other 8 feature I want to mention. Then I would recommend 9 that the Commission grant an exemption request from 10 Type A testing.
11 And if you have any questions?
12 (No audible response.)
13 MS. GRADY: Thank you.
14 MEMBER CORRADINI: I'm happy. I'm happy.
15 I'll let the members speak for themselves. Thank you.
16 MR. LU: Okay. Shanlai Lu from staff, 17 Reactor System Branch. I'm going to cover ECCS system 18 6.3.
19 I think the staff is expecting the 20 committee to have questions, as Michael mentioned 21 already at the beginning that you really understand 22 the open items, especially those uncommon open items.
23 So we do have limitations here. This is 24 an open session, so staff and NuScale are preparing to 25 have a closed session this afternoon after this one.
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273 1 We'll talk about details of the valve designs so I can 2 answer any questions about how the ECCS system 3 functions and why it works.
4 And then after that one and tomorrow, you 5 can see a lot of my open items in Section 6.3, they're 6 actually related to the ECCS performance evaluation, 7 LOCA and non-LOCA, and then also single-filler stuff 8 there.
9 So it's all -- and, you know, will be 10 presented tomorrow. And tomorrow there is also two 11 sessions. There is an open session. There is also a 12 closed session. So while it's our staff's intention 13 to brief and inform the committee as much as possible 14 to really understand all those issues we have on our 15 plate.
16 So the strategy I have for this 17 presentation is I'm going to go through a top-down 18 approach instead of going through issue-by-issue 19 explained one by one. What is the problem -- it goes 20 through top-down so that you are saved within the 21 limited amount of time you can see what's the big 22 picture of the puzzle looks like.
23 And then, in this open session, if you 24 have questions, you can ask, but if there is detailed 25 information, we need to defer to the closed session to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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274 1 do that.
2 And then there a lot of questions will be 3 related to Chapter 15 analysis, which we're going to 4 cover tomorrow. Including myself, I am going to 5 support other reviewers to do that.
6 So with that, I'm going through the first 7 slide. Okay, NuScale has already covered this one.
8 And you can see that from the ECCS system designs, 9 that they are extremely simple, three RVVs, two RRVs, 10 and each ECCS has its own IAB trip valve and trip 11 reset valve. That's it.
12 But on top of that, there are additional 13 -- its containment functions as the part of ECCS.
14 Once the valve opens, containment provides the 15 cooling, and then directly to the pool. So not only 16 does it works with the LOCA event, it works for all 17 other long-term cooling events too, once the ECCS is 18 actuated.
19 So with that one -- and then, I'm going to 20 point out that that's a unique feature of that -- you 21 can stick them back -- a unique feature of this ECCS 22 system. No other design or possibly line has such a 23 simple ECCS system. It's very unique.
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275 1 was required by GDC 27. That's the reason for a while 2 that the staff has already engaged with the committee 3 to talk about that one.
4 And that's also unique because no other 5 PWRs do not have that feature. They all can inject 6 additional boron poison into the reactor during a LOCA 7 event or ECCS actuation including AP1000 capacity line 8 too.
9 So that will lead to the next --
10 MEMBER BLEY: Can I ask you a question 11 about --
12 MR. LU: Sure.
13 MEMBER BLEY: -- how the design works? I 14 don't think we talked about this.
15 MR. LU: Okay.
16 MEMBER BLEY: The vendor -- so under our 17 LOCA we had a long talk about how much you flood up --
18 MR. LU: Right.
19 MEMBER BLEY: -- open up those valves, and 20 everything works fine. For long-term cooling 21 otherwise, if you use this system --
22 MR. LU: Right.
23 MEMBER BLEY: -- are there requirements on 24 when you open the RRVs and the RVVs?
25 MR. LU: Right.
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276 1 Okay, actually that's part of Chapter 15 2 zero scenario for non-LOCA.
3 MEMBER BLEY: That's part of Chapter 19.
4 MR. LU: Yes, also.
5 MEMBER BLEY: What's the design scenario?
6 MR. LU: Okay. All right. So there are 7 quite different scenarios that will lead to ECCS 8 actuation. One case that definitely is a LOCA, you 9 have, you know, a serious accidental line break. As 10 we just discussed, about that you had dumped enough 11 water into the containment, the water level flood of 12 containment, and that will trigger the ECCS actuation.
13 There was one point here I think somebody 14 asked a question about how is IAB is going to function 15 if you have very quick blowdown, the system pressure 16 quickly goes down below the setpoint of IAB, for 17 example, 1,100 pounds. Then IAB does not do anything, 18 valve opens -- all five ECCS valves are supposed to 19 open, but of course, you can assume certain valve 20 failures.
21 But in reality, they all need just two RVV 22 and one RRV. So once you blast it open, you dump the 23 inventory from the vessel to the containment, then you 24 establish the natural circulation.
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277 1 beautiful --
2 MEMBER BLEY: They did a good job on the 3 LOCA one, I don't remember --
4 MR. LU: So for the non-LOCA side, for 5 example, if there is an event and then I think we can 6 talk a bit more tomorrow. And then, for example, 7 let's assume the DC power everywhere -- we would lose 8 AC power 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later. It's actually -- there's a 9 timer that actuates the ECCS.
10 If at that time, the system pressure is 11 higher than the 1,100 pounds of pressure, IAB would 12 block all five of them.
13 MEMBER BLEY: All five of them.
14 MR. LU: It can add a blast of water into 15 the containment. However, during 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, normally 16 we expect the DHRS would start to function and start 17 to cool them down. And then by the time you really 18 reach 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -- depending on scenario -- you analyze 19 -- then there is a possibility most likely your 20 pressure will be lower than the 1,100 pounds setpoint, 21 IAB will not block your blast open all five valves at 22 that point. So that's for the long-term cooling.
23 Okay, any questions for this slide?
24 (No audible response.)
25 MR. LU: Okay, we're good. Let's go --
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278 1 the regulatory basis -- so I'm going through the 2 top-down approach. So I'll start from regulatory 3 basis and that I think you guys are all familiar with 4 that one. So I'm not going to spend time to explain 5 one by one.
6 MR. LU: But I'm going to summarize, the 7 GDC 2 and GDC 4 have been covered by the environmental 8 review of Chapter 3, the sharing of structure system 9 and components as covered by the pools, how many 10 modules, you know, are shared. It's not a big deal.
11 The electrical power system requirements 12 and then that's the part you probably want to pay 13 attention to. But I think the electrical branch, I've 14 already covered that one because this one does not 15 have the safety grade of DC power.
16 Neutron poison addition and the 17 appropriate shutdown margin for stuck rods. I think 18 that's part where lead to one of the uncommon open 19 items here. Because if it does not have Neutron 20 poison addition capability and resolving the re-21 critical so there's no -- if it reaches the re-22 critical reason sometimes there's no shutdown margin 23 for stuck rods.
24 So that's the part the staff has been 25 working with NuScale and back and forth for a while.
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279 1 And also I think the other staff from the entry system 2 has already briefed the meeting about that.
3 All right. Can we just touch a few more 4 bullets there? Long-term cooling, okay, 50.46(b),
5 long-term cooling, although we did not and, you know, 6 have any specific issue related to rule or regulation, 7 but to really make sure it works, the system can 8 remove not only decayed but also if you have re-9 critical, lots of power, how are you going to remove 10 the power? That's the issue there.
11 Related to the testing, and I think Tom 12 already covered that we are, you know, observing and 13 auditing the NuScale's ECCS performance testing 14 program. So it's ongoing. Actually, it's going 15 according to the plan and then we are going to go back 16 to re-observe the test.
17 Okay. Next slide. This part I don't want 18 to touch too much because that's a typical DSRS for 19 623. We modify that one based on -- back to 2014 20 there is a white paper about the specific NuScale ECCS 21 system. So based on that, we revised the DSRS. We'll 22 follow that one.
23 And the TMI action item, one of them 24 related to boron, boron dilution from precipitation, 25 so that's a part that I'm going to cover a little bit.
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280 1 There are two exemption requests here.
2 The first one actually as I mention it because the 3 NuScale ECCS system design is passive, simple and what 4 they claim there is a lot of margin because two 5 freezable level is always above the core for the 6 largest design basis LOCA. So, therefore, the core 7 never experienced a reflood dry out. So those are the 8 issues back to 2014.
9 There was a white paper submitted by 10 NuScale to the staff. We reviewed that. But 11 basically, we claim at that time was there was no IAB 12 design so basically they said because of this over a 13 combination of the walls and the containment concept 14 and that those are the items they are going to seek 15 the exemption request from us as part of 50.46, 16 Appendix K requirements.
17 And we actually at that time, you know, we 18 made a decision very quickly is that, yes, sounds 19 good.
20 So it's not a final approval, but we 21 basically gave them our feeling because of, you know, 22 our recommendation or our comments, yes, looks good to 23 us. Because we had exposure to their testing program.
24 We had exposure at that time whether they predict a 25 two freezable level.
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281 1 As long as they can support that claim 2 that at the end for the most limiting case, they do 3 not have core uncovered, I think this part makes 4 sense. Of course, there is no pump. And a pump 5 needed to be there.
6 So this partial exemption at this point as 7 part of the ECCS performance evaluation and we feel 8 comfortable and we recommended the approval of this 9 exemption request. And we are going to cover a little 10 more tomorrow as part of a LOCA analysis and LOCA 11 topical analysis and summary.
12 GDC 27, okay. We haven't talked about the 13 detail analysis of this one, especially we have a 14 proprietary session tomorrow to talk about the staff 15 calculations, evaluations.
16 And as I mentioned right at the beginning, 17 the unique features of this ECCS system, it does not 18 have neutron poison injection addition capability in 19 comparison with all other PWRs in the world.
20 So what is going to be the impact on this 21 design that they already told us it's going to become 22 a re-critical. And on top of re-criticality, they 23 felt that they are not going to change the CHF. So 24 staff is evaluating their space safeguard analysis to 25 see whether it's sufficient enough for us to prove NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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282 1 this exemption or not.
2 So that's where we have uncommon open 3 items. All right. Now --
4 MEMBER BLEY: And that one is really 5 breaking new ground.
6 MR. LU: That's their side. Yes.
7 MEMBER BLEY: Yes.
8 MR. LU: That's the reason it's not 9 something --
10 MEMBER BLEY: For you, too, I mean, you've 11 never proved something quite like this.
12 MR. LU: We were aware about that issue --
13 MEMBER BLEY: Oh, you knew it was coming, 14 no?
15 MR. LU: -- three months before this DCD 16 submittal. So we asked --
17 MEMBER BLEY: But you never issued a 18 license allowing a condition like this?
19 MR. LU: We in the past history, never.
20 MEMBER BLEY: Never, ever.
21 MR. LU: Never. And actually one of the 22 -- for the Fukushima event, I think all the world 23 regulatory agency grouped together and developed two 24 common positions. Although we have to deal with our 25 kinds of different governments differences, but there NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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283 1 was only two. Remove decayed heat and maintain core 2 subcritical during downtime cooling.
3 So I'm not so sure. But I'm not only a 4 reviewer. I think also, you know, it's another 5 reviewer's responsibility to call the final shot on 6 that one. But that's my review on the ECCS system.
7 So I'm going to go through from a top down 8 perspective the pieces which can give you a color of 9 the puzzle of what sections impact ECCS performance.
10 So the first one, reactor coolant 11 boundary, as Tom has already covered that one. And so 12 open item is 3.96-1 related to ECCS performance 13 testing needed. So we are doing that. And actually 14 there is a lot of good activity.
15 And then the conceptual design last 16 testing last year went well. And so we are expecting 17 hopefully this one goes through easily, too. But it's 18 ongoing so we cannot say one way or another at this 19 point.
20 All right. Low temperature over pressure 21 protection. In Chapter 5 of the presentation, HR has 22 asked a question. I was not present. And the 23 reviewer passed me the question.
24 So I think I can answer that one. And 25 then the concern at that time was, oh, you have IAB.
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284 1 Are you going to block it at the ECCS actuation to 2 protection their low temperature over pressure.
3 And then we checked with NuScale and they 4 provided the actual protection curves. And then the 5 LTOP through ECCS system is enabled only below 750.
6 So therefore IAB should not interfere during the 7 start-up. So we are still having some conversation 8 with them over the come down with the power. So that 9 might be the case.
10 MEMBER BLEY: That's all automatic here.
11 It's not --
12 MR. LU: Yes. It's automatic. It's not 13 menu. It's not menu. But there is also a menu 14 operation to main they are supposed to follow to 15 protect.
16 All right. All right. Core, you know, I 17 think the open Item 6.3, that's wrong. It's the core 18 cooling related to inadvertent opening of ECCS valves.
19 And then it's right now it's pending on the LOCA 20 topical report at 50.6 review, Open Item 6.3-1. Core 21 cooling secondary --
22 MEMBER CORRADINI: I don't think I 23 understand that.
24 MR. LU: Okay.
25 MEMBER CORRADINI: I'm sorry.
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285 1 MR. LU: Sure. Go ahead, Mike.
2 MEMBER CORRADINI: Is your tracking as an 3 open item because it's a matter of the modeling tool 4 has yet to be upgraded because it's related to 15.02-5 2, which is essentially the NRELAP Version 1.4 that is 6 yet to be reviewed. Is that not correct?
7 MR. LU: You're correct.
8 MEMBER CORRADINI: Okay.
9 MR. LU: And actually that's the ECCS 10 performance valuation. At 6.3, it's a summary of all 11 those sections, supporting sections and then the 12 performance evaluation satisfies all those basic 13 criteria.
14 MEMBER BLEY: Okay. Well, and the first 15 half of this --
16 MR. LU: Right.
17 MEMBER CORRADINI: -- is this inadvertent 18 ECCS, is that one of the LOCAs that's analyzed in the 19 technical report?
20 MR. LU: Yes, sure.
21 MEMBER CORRADINI: It is, isn't it?
22 MR. LU: Yes. It's one of the -- they 23 claim that the -- the NuScale claim is one of the 24 transit.
25 MEMBER CORRADINI: There is --
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286 1 MR. LU: Non-LOCA transit.
2 MEMBER CORRADINI: -- we're going to go 3 back to it tomorrow. But there are six open items 4 related to the model, the tool. And 15.02-2 is one of 5 those in terms of essentially there's a new version of 6 -- if I understand correctly, there's a new version of 7 NRELAP that staff is still reviewing compared from 1.3 8 to 1.4 and that affects essentially all the accident 9 analyses. Thanks.
10 MR. LU: Okay. All right. That's a 11 similar issue related to non-LOCA because during non-12 LOCA the ECCS system at a certain point will be 13 actuated to maybe the non-LOCA. So therefore that 14 part is open.
15 Those open items are what we consider with 16 a clear path forward. Okay. And core cooling loss of 17 coolant accident also we consider with a clear path 18 forward. It is related to the Item 3.62-1 and also 19 15.0 on 2-2. It's a LOCA topical report.
20 Okay. The core cooling long-term cooling 21 return to power and boron transport. That's the 22 section of 15.06 and staff is going to cover that in 23 detail. That's what we considered an uncommon open 24 item.
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287 1 move forward. Next slide.
2 MEMBER BLEY: Are you going to get to that 3 one tomorrow or that will be later?
4 MR. LU: Actually, the staff will prepare 5 a lot of material for tomorrow --
6 MEMBER BLEY: All right. Okay.
7 MR. LU: -- as part 15.0.6 and actually my 8 management had asked me to talk about the boron 9 transport, too. So I will be back here tomorrow, too.
10 MEMBER BLEY: Okay, good.
11 MR. LU: So if you have any question, you 12 can ask me at that time, too. All right. Closed.
13 And then power requirements related to non-safety 14 graded EC power system. And I think there is a pass 15 forward there already.
16 Instrumentation, Jose, you asked a 17 question about the level and initiating our LOCA 18 review. And you asked a question how you did you 19 measure the level and then how did you trigger your 20 protection system?
21 And the initial that they had a vessel 22 with level sensors there. So our questioning at that 23 time was through the Digital I&C Chapter 7 guises.
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288 1 of detector would ever function in that vessel.
2 CO-CHAIR MARCH-LEUBA: My comment was more 3 related to the possibility of digital common cause 4 failures, which we always worry because we cannot 5 quantify them easily.
6 MR. LU: Yes.
7 CO-CHAIR MARCH-LEUBA: They go undetected 8 for two years because you're not using the sensor.
9 MR. LU: Yes. I think that's the similar 10 aspect. So later that they switched to all 11 containment. And that particular issue, I recall that 12 I was talking to the digital I&C guys -- at that time 13 it's part of inspection, the maintenance and the in-14 service testing.
15 So how they're going to do that -- I think 16 at this point Chapter 7. Okay.
17 CO-CHAIR MARCH-LEUBA: Yes. We keep it in 18 mind for Chapter 7 and maybe we will repeat tomorrow 19 again.
20 MR. LU: Okay. System boundary closed.
21 And when NuScale presented that one, we had the trip 22 set valve and then the reset valve all enclosed as a 23 part of containment boundary.
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289 1 the process to address a lot of issues related to 2 testing.
3 The environmental requirements is closed 4 as I mentioned earlier and actually the Chapter 3 guys 5 presented on that one.
6 System reliability, failure mode and the 7 effects analysis also related to the task. And we 8 want to make sure what we, you know, suspect are the 9 issues. For example, water hammer, what's going to 10 happen there? So that's the part of a testing we're 11 going through this week and next week.
12 Single failure and it's an open item right 13 now. And we are going to follow Commission's 14 decision.
15 Technical specification, closed. Chapter 16 7 --
17 MEMBER CORRADINI: Let's stop there for a 18 second. So that is still pending?
19 MR. LU: I have not seen the Commission's 20 final decision paper yet.
21 MEMBER CORRADINI: Thank you.
22 MR. LU: Okay. Any questions on this 23 page? Next slide.
24 All right. Let's talk about now that's 25 the top down view of all the issues inter-related to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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290 1 ECCS evaluation. And now as part of DCD 6.3, they 2 submitted the in-vessel downstream effects as part of 3 the DCD submittal.
4 Actually, they had a lot of good details 5 inside of that Section .3 write-up. So we reviewed 6 that one. As it was, you know, discussed this morning 7 that they only used the reflector metallic 8 installations. And there is no fiber, no insulation 9 nor a particular insulation, no chemical buffer.
10 So the fiber and particulate and all those 11 are intentional and some are insulation within the 12 zoning infinites are excluded. So, therefore, all 13 they have is they slosh the containment through the 14 cleaner process and then it's unlikely there is only 15 going to be even latent debris. But they are digging 16 post their COL item to ensure the cleanness. I think 17 that's sufficient.
18 Debris transport for the given assumed 19 latent debris, we assume 100 percent transportable 20 from containment to the RRV to the inlet of the core.
21 The good news here is that they are using 22 AREVA fuel. So the AREVA provided them actually our 23 reference testing results. And so the results show 24 there was no problem for the given debris loading they 25 are expecting.
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291 1 If you really pay attention to the number, 2 5.6 gram per assembly fiber, it's lower than 15 gram.
3 We are approving the operating fleet.
4 So they only have 5.6 for the given COL 5 action item, which is already conservative from my 6 perspective. So therefore on top of that they assume 7 very conservative chemical participants although they 8 do not have chemical buffer.
9 So where does it come from? They just 10 want to make sure it's conservative and staff has no 11 more questions. I did not have more questions after 12 reading. That's why it's approval. So I don't have 13 a problem with that one. So that part is done.
14 MEMBER CORRADINI: What is this -- I think 15 I understand what you're saying here in terms of their 16 assumptions. But how do they arrive at these values?
17 MR. LU: Okay. Arrive at the values as 18 part of the -- they just assume for the certain amount 19 that's part of cleaning program there is a -- I think 20 there is a merit about a specific number.
21 And then checking certain amount of latent 22 debris they could even find in the containment and 23 then so they just justified that they do not have a 24 certain amount of total number of fiber in particular.
25 That's part of what the cleaning program would do.
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292 1 And then you divide the total number by the total 2 number of assemblies. That's what you get for the 3 fiber.
4 CO-CHAIR MARCH-LEUBA: Is it a speck on 5 the pool that the pool water's cleanliness? We talked 6 this morning about algae growing on certain pools.
7 MR. LU: Okay.
8 CO-CHAIR MARCH-LEUBA: Because the only 9 way that debris can make it after refueling --
10 MR. LU: Right.
11 CO-CHAIR MARCH-LEUBA: -- is if it comes 12 with the water in the pool and then you evaporate the 13 water and leave the debris behind, correct?
14 MR. LU: Yes.
15 CO-CHAIR MARCH-LEUBA: Because all I was 16 going to see how --
17 MR. LU: Okay. Yes. I think my 18 understanding, when I was reviewing that part --
19 that's a good question. Thank you.
20 They can drain the water through the --
21 they have to drain the water eventually, right? So by 22 the time they really drain that one, and I'm not so 23 sure there's any, you know, significant amount of, you 24 know, debris would be there anymore.
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293 1 at the bottom. They are pushing from the top and 2 eventually they evaporate.
3 MR. LU: Right. Right. Yes, but --
4 CO-CHAIR MARCH-LEUBA: So any solids will 5 settle down.
6 MR. LU: Yes. That's what I was thinking 7 about since they are moving the entire containment 8 from one side to another side. By the time it gets 9 sloshed all around, they stir up all the dirt if there 10 is any.
11 CO-CHAIR MARCH-LEUBA: Yes.
12 MR. LU: And then they have clean-up 13 system continuously. So that part you can also 14 assume, but with the margins we have we're talking 15 about -- let's say if we're down more area, they only 16 can because they have 5.6 gram per assembly fiber.
17 That's three times. That means your cleaner's 18 program, let's see, if you really have this scenario, 19 you can still survive.
20 CO-CHAIR MARCH-LEUBA: I'm pretty sure 21 that if you start growing a green algae on top of the 22 pool that --.
23 MR. LU: Yes. I would say that nobody 24 wants to see muddy water, you know, in a containment 25 pool. All right. Next slide.
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294 1 Water hammer, okay. It's part of the 2 requirement that we review the water hammer for the 3 ECCS system. But in a standard ECCS system, you have 4 large pipe. You have a 90 degree bend. When you have 5 water coming in, there's a possibility you have water 6 hammer.
7 For this one, the people were asking me 8 why do ask questions? And then that was because 9 between the actuator line and the IAB you have a very 10 long line. And from RRV to RVV to that the trip set 11 valve there is a long line.
12 MEMBER CORRADINI: It's a sampling line.
13 It's a small line.
14 MR. LU: It's a small line. That's 15 exactly the case. So we asked as part of Tom's 16 performance testing requirement. And we actually 17 imposed that it was specific to maintain the pressure 18 of the trip actuator line, trip set off line, make 19 sure that it's high temperature, high pressure 20 temperature up to 400 degree Fahrenheit.
21 So when trip valve opens, boom. We have 22 a critical flow going through. So what's going to 23 happen to those vents? So the good news at least at 24 this point we observed one of the tests that did not 25 show water hammer after they put quite a few vents NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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295 1 there. So we are waiting for these testing results 2 and then waiting for their analysis coming back to 3 show, yes, it's not a problem.
4 So this particular item, and it's part of 5 open item associated with Tom's performance testing, 6 and I think we have a clear pass forward. So next 7 slide.
8 Conclusion. This particular section and 9 I'm not surprised because that's brand new in the 10 latest design so we have open items there. But we 11 have seven open items.
12 And right now we have three open items of 13 what we consider uncommon. But actually the second 14 one, the second paper, we got an IAB single filter 15 where we cannot discuss tomorrow. We're going to 16 follow the Commission's direction.
17 So ECCS demonstration testing, although we 18 are still saying it's not done yet, but we hope this 19 testing will go well and that we observe that one.
20 And it looks like NuScale already responded to staff's 21 requests. They did what we asked during the audit 22 process in March and then they assembled the testing 23 program well. And so we hope that will be okay.
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296 1 tomorrow. We have detailed slides and discussion 2 about that.
3 So that's what I got. Any other 4 questions?
5 CO-CHAIR MARCH-LEUBA: Maybe I missed the 6 presentation, but on the reliability of the valves, 7 there are three, five valves in the system?
8 MR. LU: Okay.
9 CO-CHAIR MARCH-LEUBA: And they're each 10 kind of mobile design?
11 MEMBER BLEY: Yes.
12 CO-CHAIR MARCH-LEUBA: I mean, complicated 13 enough so when I see a drawing I don't see how it 14 works.
15 MR. LU: That's a perfect question. I 16 think that will be covered by our closed session.
17 CO-CHAIR MARCH-LEUBA: The closed session.
18 But you --
19 MR. LU: Because until you see the 20 details.
21 CO-CHAIR MARCH-LEUBA: Can you tell us the 22 conclusions in the open session? I mean, don't give 23 us the details, but what are the conclusions?
24 MR. LU: The conclusions, the reason I'm 25 still in the open item, actually it's Tommy that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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297 1 still only in open item stages because we cannot give 2 a person a conclusion until we see something docketed 3 from NuScale.
4 CO-CHAIR MARCH-LEUBA: Fair enough.
5 MR. LU: Any other questions?
6 MEMBER CORRADINI: Other questions?
7 Turning to 6.4.
8 MS. HART: Yes. I'm Michelle Hart. I'm 9 with radiation protection and accident consequences.
10 And I'll be talking about the control room 11 radiological habitability.
12 The main regulation that we look at is GDC 13 19 which has a design criterion of 5 rem TEDE for the 14 duration of the accident.
15 However, I do have an open item 6.4-1 16 because NuScale did request exemption from the control 17 room design criteria in GDC 19 to instead propose a 18 NuScale principle design criterion 19.
19 The effect of that proposed exemption does 20 not change the dose requirements. It does make a 21 chance to the remote equipment outside the control 22 room that would refer to being able to take it to safe 23 shutdown instead of cold shutdown and hot shutdown.
24 Although there is no change to the dose 25 requirement, I am tracking it as an open item in 6.4 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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298 1 for the radiological habitability in that I would have 2 to refer to PDC 19 if we were referring to that 3 instead of GDC 19.
4 Next slide, please. The next topic is 5 that the dose to control room operators from accidents 6 is analyzed in DCA, the FSAR Chapter 15. The 7 specifics of the radiological consequence analysis are 8 presented in Chapter 15.
9 Those do include several design basis 10 accidents and also includes direct dose contribution 11 from the core damage event from filters and outside 12 radiation cloud.
13 Open Item 6.4-3 is that there are 14 revisions to the dose analyses that were provided in 15 conjunction with Revision 3 to the accident source 16 term methodology topical report, which was just 17 received in late April 2019.
18 At the time of writing of this Phase 2 19 draft SER we did not have that information in hand.
20 And so that review is certainly not complete. We've 21 just gotten it. There are not a lot of changes from 22 what was seen before. But the review is just not 23 complete at this time.
24 So to go to the big topic in control room 25 radiological habitability, one of the things that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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299 1 NuScale did propose is that the post-accident control 2 room habitability is not a safety-related function.
3 And that is because there are no operator actions that 4 are required or credited to mitigate the consequences 5 of design basis accidents. And also there are no 6 post-accident long-term monitoring that is necessary 7 to be taken from the control room.
8 Open Item 6.4-2, this relates to a 9 discussion that Anne-Marie had brought up earlier.
10 Because they are required to monitor oxygen and 11 hydrogen concentrations in the containment, they do 12 have to set up that monitoring pathway. They have to 13 reopen the containment to be able to have those 14 monitors take those concentrations.
15 It is not the same thing as sampling.
16 It's in the same system as sampling. They do have an 17 exemption request from taking samples altogether.
18 MEMBER CORRADINI: So I don't think I 19 appreciate that subtlety. Can you try that one on me 20 again?
21 MS. HART: So a TMI related action item 22 requirement in 10 CFR 34(f)(2), I think it's (viii),
23 would have you take samples quickly after an accident.
24 And they have requested an exemption from that 25 regulation. We have not completed the review of that.
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300 1 And so that's the sampling requirement.
2 And so that may be sampling for chemical purposes and 3 also may include combustible gas sampling.
4 However, for the requirement for 50.44(c),
5 they have to monitor combustible gasses and that's 6 continuously monitor. It's in the same pathway that 7 they have in their systems. It's in the process 8 sampling system and it's online monitor.
9 So we have some questions about whether 10 there are actions they have to take from the control 11 room for 6.4, the open item. Are there actions that 12 they would have to take from the control room to be 13 able to set up that pathway to be able to do 14 combustion gas monitoring?
15 MEMBER CORRADINI: So they're not taking 16 a sample. They're going to have to use the same flow 17 pathway for some sort of monitoring post-accident?
18 MS. HART: That's correct. And so we 19 still have concerns about -- in Chapter 12, we still 20 have some concerns about if there are local actions 21 they have to take, is there an impact to the dose --
22 MEMBER CORRADINI: Okay.
23 MS. HART: -- to the operator to take 24 those actions?
25 MEMBER CORRADINI: I remember that from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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301 1 Chapter 12.
2 MS. HART: Right.
3 MEMBER CORRADINI: Okay.
4 MS. HART: But from Chapter 6 perspective, 5 are there actions they have to take in the control 6 room to be able to do this? So this goes directly to 7 that discussion as to whether control room 8 habitability is a safety related function or not.
9 MEMBER BROWN: Before you leave --
10 MS. HART: Mm-hmm.
11 MEMBER BROWN: -- put aside the sampling 12 thing. What does it mean to the control room to be 13 either safety related or not safety related? I'm 14 having a hard time --
15 MS. HART: Right.
16 MEMBER BROWN: -- figuring out why the 17 operation of the control room -- regardless of what 18 operator actions are required or not required is not 19 a safety-related --
20 MS. HART: Right.
21 MEMBER BROWN: -- need to me. I don't 22 understand the difference, okay? What does it mean to 23 be non-safety related? Does that mean they can go 24 home and have a beer or abandon --
25 MS. HART: What I can tell you is that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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302 1 NuScale has stated that within the context of having 2 to have control room radiological habitability be 3 supplied by safety related equipment that that post-4 accident radiological habitability in the control room 5 is not a safety function because there are no operator 6 actions that are required for the control room to 7 keep.
8 MEMBER BROWN: They can leave because it 9 doesn't have to be a radiation -- you don't have to be 10 able to stay in there during the accident. That's the 11 way I read your --
12 MS. HART: They do have systems to provide 13 that. They do not need to be safety related to assure 14 it to that higher mandate that a safety-related 15 function would have you take. I know it's a very fine 16 point. We've had a lot of discussions about this.
17 MEMBER BROWN: Well, there's a lot of fine 18 points. I mean, here we can stay critical for days 19 after a plant shuts down without having to -- I mean 20 stack on another one. We don't have to be 21 radiologically secure in the control room.
22 MS. HART: Right. Certainly, their plan 23 is not to evacuate the control room. Their plan is to 24 stay in --
25 MEMBER BROWN: I understand the plan, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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303 1 --
2 MS. HART: Right. There are no actions 3 that are necessary from the control room. So if they 4 did have to evacuate --
5 MEMBER CORRADINI: So let me turn 6 Charlie's question around a different way. If it were 7 safety related, what would be required that they're 8 not providing already now?
9 MS. HART: Safety related ventilation 10 systems.
11 CO-CHAIR MARCH-LEUBA: And maybe you want 12 to back up. I mean you have to have two systems. And 13 right now you have only one, right?
14 MS. HART: You may only need one system.
15 But it needs to be safety related.
16 CO-CHAIR MARCH-LEUBA: Do they have --
17 MS. HART: It's like all the other plants 18 have one safety related --
19 CO-CHAIR MARCH-LEUBA: Typically, if you 20 leave the room --
21 (Simultaneous speaking.)
22 MS. HART: -- control room habitability 23 system. It's a --
24 CO-CHAIR MARCH-LEUBA: Requires.
25 MS. HART: -- filtration system or like in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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304 1 a thousands of cases --
2 MEMBER CORRADINI: The filtration and air 3 supply system they have is provided. It's just not 4 safety related.
5 MS. HART: That is correct.
6 CO-CHAIR MARCH-LEUBA: I think it's an 7 excellent solution that NuScale has proposed to have 8 compressed air and feed them -- the operators as if 9 they were diving under water. I mean, you just use 10 your compressed air, which we know is clean and it 11 hasn't been affected by the accident. The question is 12 do you want it classified as safety related or not?
13 MS. HART: Correct. So their systems look 14 similar to other systems we've seen in other 15 facilities. You know, we do have bottled air systems 16 in some currently operating plants.
17 The AP1000 also has a similar safety 18 related, in their case, bottled air system that they 19 rely on for control room habitability.
20 CO-CHAIR MARCH-LEUBA: But my concern of 21 your Open Item 6.2 is -- I'm not concerned about 22 control room habitability. They have air, and they 23 have a system.
24 But if you decide that something is a 25 safety-related action, then you need a minimum DC NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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305 1 power and operability of those valves and operability 2 of your instruments. So if you decide that is a 3 safety-related action, it's a cascade of events.
4 MS. HART: That is correct.
5 CO-CHAIR MARCH-LEUBA: It is not just the 6 bottled air.
7 MS. HART: And that's why it's still under 8 discussion at this point.
9 CO-CHAIR MARCH-LEUBA: It's a big 10 decision.
11 MS. HART: Correct.
12 DR. SCHULTZ: Michelle, you said earlier, 13 at least I heard earlier and then we're going to talk 14 about it tomorrow, the expectation or the requirement 15 is still that the control room dose will be less than 16 5 rem.
17 MS. HART: That is correct.
18 DR. SCHULTZ: Okay. So we'll talk more 19 about that.
20 MS. HART: Okay. So to get to that point 21 that we were just talking about, the NuScale design 22 does have two systems that they use for ventilation 23 and/or habitability in their control room. And the 24 credit they've taken in those dose analyses is for 25 these systems that are not engineered safety features.
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306 1 And so in effect the control ventilation 2 system and the control room habitability system, the 3 normal ventilation system and the bottled air system 4 are backup systems for each other. They're both 5 designed to be reliable and capable of operation 6 during accident conditions. And in this case the 7 accident duration is 30 days as for all other plants 8 so far.
9 They are independent and diverse systems.
10 The bottled air system does not use the same injection 11 points as the duct work for the ventilation system.
12 They both have automatic initiation with 13 different signals based on the radiation. And for the 14 habitability system, the bottled air system as they 15 described earlier, they do have initiation signals on 16 that based on power loss. There's a couple different 17 flavors of that and also there's a backup manual for 18 those systems.
19 So in the dose analyses, they analyzed two 20 different dose analysis cases. For the one case, they 21 assume that the bottled air system operates for 72 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />. Then the control room ventilation system in 23 the supplemental filtration mode comes on after that 24 and operates through the rest of the duration of the 25 accident. And that's 30 days.
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307 1 In the second case, they assume that the 2 normal ventilation system stays in supplemental 3 filtration mode for the duration of the event. And in 4 both cases, they meet the dose criteria for all 5 accidents analyzed, all those DBAs in Chapter 15.
6 So we did have some concerns about the 7 apparent reliance on both systems over the duration of 8 the accident to meet the control room requirements in 9 that first case that they analyzed where the --
10 MEMBER CORRADINI: It's only the first 11 case that --
12 MS. HART: That's correct, yes. The 13 second case it's only the one system that they're 14 relying on.
15 For that first case, the control room 16 habitability system does operate for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. And 17 then the control room ventilation will come in 18 filtration mode. So we asked them some questions 19 about the sensitivity if those systems would fail to 20 operate as expected.
21 And so they provided us some sensitivity 22 analyses where the control habitability system, the 23 bottled air system, operates for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 24 and then the control room ventilation system fails and 25 the control room habitability system does not come NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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308 1 back in. So there's no ventilation filtration after 2 the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3 They also provided a more extreme case 4 where neither the bottled air system nor the control 5 room ventilation supplemental filtration system 6 operate.
7 And all of those show that they meet the 8 dose criteria. They do have higher doses than are 9 reported in the DCD, of course. But they still do 10 show doses that are less than 5 rem criterion.
11 DR. SCHULTZ: Michelle, where are these 12 analyses reported? Was that in response to an RAI?
13 MS. HART: It's in response to an RAI, 14 yes. And unfortunately, I don't have the RAI 15 responses on me.
16 MEMBER BROWN: When you say they're 17 higher, what do you mean they're higher?
18 MS. HART: They're higher than 2.14 rem.
19 MEMBER BROWN: But what does that mean?
20 Are they still considered --
21 MS. HART: They're still less --
22 MEMBER BROWN: Are they 25?
23 MS. HART: They're less than 5 rem.
24 MEMBER BROWN: They're less than 5.
25 MS. HART: But they're higher than what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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309 1 are reported in the design certification documents.
2 So therefore what's reported in their FSAR is not the 3 result of their sensitivity analysis is what I'm 4 trying to get across.
5 DR. SCHULTZ: The 5 rem is accumulated 6 over the 30 day period?
7 MS. HART: That's correct. And does 8 include direct dose from the filters as well.
9 MEMBER BROWN: If you could give us 10 reference to that RAI, not now. Tomorrow is fine.
11 MS. HART: Sure.
12 MEMBER BROWN: I presume you'll be back 13 for Chapter 15.
14 MS. HART: You may be disappointed in my 15 Chapter 15 presentation. I will just let you know 16 that right now.
17 MEMBER BROWN: I understand why. Well, 18 tomorrow would be fine.
19 MS. HART: I'm just saying. You may not 20 have as much information as you would hope for.
21 MEMBER CORRADINI: But I think if we could 22 get the RAI, then I think that's what Steve's --
23 MS. HART: Yes. Perhaps by the end --
24 DR. SCHULTZ: Yes. I would like to see 25 that comparison. It would be interesting.
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310 1 MS. HART: Right.
2 DR. SCHULTZ: There are a lot of different 3 assumptions that are in tomorrow's evaluation.
4 There's a couple --
5 MS. HART: Correct.
6 DR. SCHULTZ: As you say things have 7 changed in the analysis, input --
8 MS. HART: Right.
9 DR. SCHULTZ: -- assumptions have changed.
10 MS. HART: And I myself have done a lot of 11 sensitivity analyses to try to push the boundaries and 12 see -- and, of course, you can push it to where it is 13 much greater than 5 rem if you come up with the 14 correct reasoning. If you make some other assumption 15 changes, especially in the core damage event --
16 DR. SCHULTZ: Okay.
17 MS. HART: -- for any accident other or 18 any analysis other than the core damage event.
19 They're not going to approach the control room dose 20 criterion.
21 DR. SCHULTZ: Thank you.
22 MEMBER CORRADINI: Thank you.
23 MEMBER BROWN: Are there requirements for 24 the core damage event?
25 MS. HART: I'm sorry. Say that again.
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311 1 MEMBER BROWN: You said they could be 2 higher than 5 if you had a core damage event.
3 MS. HART: No, if there were changes to 4 the dose analysis, if there were -- because we are 5 still evaluating the control room -- I mean, excuse 6 me, the core damage event, it's not a final.
7 There were some questions that we had 8 about containment leak rate. There were some 9 questions that we had about the particular assumptions 10 of the release rate to the containment, if those would 11 have to be changed because it's 2.14 rem and with 12 those criterion it's 5 rem. There's not a huge amount 13 of margin. So there's a potential that they could 14 increase to be over the top of that.
15 MEMBER BROWN: Over the 5.
16 MS. HART: Correct. But right now, I 17 don't think that that's the case. And their dose 18 analyses and my confirmatory analyses but the current 19 dose analyses do not show that that is the case.
20 MEMBER CORRADINI: Okay. Further 21 questions by the members because I want to go to 22 public comments and then we want to have enough time 23 for our extended closed session. Okay. Any more 24 questions by the members? All right.
25 Why don't I ask then if there are members NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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312 1 of the public that want to make a comment in the room.
2 And then we'll open the phone line. Nobody in the 3 room? The phone line is open. Could we ask anybody 4 on the open phone line for the public to please at 5 least acknowledge that you're out there.
6 MS. FIELDS: This is Sarah Fields so.
7 MEMBER CORRADINI: Okay. Thank you very 8 much.
9 MS. FIELDS: There are listeners out 10 there.
11 MEMBER CORRADINI: Good. Go ahead with 12 your comment, please, Ms. Fields.
13 MS. FIELDS: A couple of comments. The 14 first are concerns. According to the NRC schedule for 15 the review -- there's a lot of background noise here.
16 I don't know what that is.
17 MEMBER CORRADINI: We can barely hear you.
18 You're going to have to speak -- we can barely hear 19 you.
20 MS. FIELDS: There was some loud 21 background noise. Can you hear me better now?
22 MEMBER CORRADINI: I think if you would 23 get closer to the microphone. If you have a receiver, 24 pick up the receiver so we can hear you a little 25 better, please?
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313 1 MS. FIELDS: Is that better?
2 MEMBER CORRADINI: Yes.
3 MS. FIELDS: Okay. I'm talking into my 4 mobile phone. I'm concerned that it appears that the 5 NRC staff intends to commence the design certification 6 rulemaking next December.
7 It's not been officially put on the 8 website part of the schedule. But it was commenced 9 before the ACRS review of the advanced SCR with no 10 open items and before the final SCR with no open 11 items.
12 So I'm concerned that rulemaking will 13 commence before making a public comment before the 14 ACRS completes its review and before the NRC completes 15 its final SCR approval. So that's my concern.
16 MEMBER CORRADINI: Okay.
17 MS. FIELDS: I know you're not going to --
18 MEMBER CORRADINI: May I repeat your 19 comment so I understand it? You're concerned about 20 the time scale of the final rule versus ACRS review.
21 Is that what you're saying?
22 MS. FIELDS: Right.
23 MEMBER CORRADINI: Okay. Fine. I have 24 your comment. Thank you.
25 MS. FIELDS: And then another comment is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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314 1 it's pretty clear there's some internally important 2 open items that have yet to be resolved. And also 3 there are a number of extensions left and assuming 4 that the NRC assumes most of those extensions that I 5 wonder what the NRC is going to do if after a period 6 of operational history of this design, it has no 7 operational history.
8 So if after a period of operational 9 history, the operators in the NRC find that some of 10 these assumptions in these extension requests don't 11 hold true during operations and there may be other 12 things that don't hold true for this design.
13 And I think the NRC should make clear to 14 the public what will happen if and when one or more 15 aspects of this design and one or more aspects of the 16 bases for the exemption requests don't pan out in 17 reality.
18 Those are my comments. Thank you very 19 much.
20 MEMBER CORRADINI: Thank you. Is there 21 anyone else on the line that wants to make a comment?
22 Okay. Hearing none, why don't we close the public 23 line, please? And at this point, we're going to go 24 into closed session. So I'm going to thank the staff 25 --
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315 1 MR. TABATABAI: Before we go into the 2 closed session, I just want to -- I think Michelle has 3 an update for an RAI number that Mr. Schultz asked.
4 MEMBER CORRADINI: Okay.
5 MR. TABATABAI: And also I just want to 6 make sure that you're aware that the staff, for the 7 purpose of the Chapter 6 presentation, we are not 8 going to present any slides during the closed session.
9 We don't have any proprietary slides to show. But 10 we'll be in the audience to answer any questions.
11 MEMBER CORRADINI: That's fine. So what's 12 the information about the open item that you wanted to 13 give us?
14 MS. HART: Okay. So you were asking about 15 the RAI where I ask about the sensitivity analysis for 16 the dose analysis. And it was RAI 9534 and it was 17 specifically Question 06.04-4. And there were several 18 responses to that.
19 MEMBER CORRADINI: What's the question 20 again, please?
21 MS. HART: It was a question about the 22 sensitivity analysis if you --
23 MEMBER CORRADINI: But the number again?
24 MS. HART: I'm sorry, 06.04-4 and that was 25 in RAI 9534.
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316 1 MEMBER CORRADINI: Okay. Thank you.
2 MS. HART: Yes.
3 MEMBER CORRADINI: All right. So with 4 that I'm going to ask the staff to please -- and 5 NuScale to make sure the room is closed so that we 6 have only the appropriate people in the room. And 7 we'll take a few minutes to kind of do that. And then 8 can we close the public line? And then the -- I 9 assume NuScale's private line to its subject matter 10 experts will remain open.
11 (Whereupon, the above-entitled matter went 12 off the record at 4:15 p.m.)
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LO-0619-65896 June 11, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS NuScale Subcommittee Presentation: FSAR Chapter 3, Design of Structures, Systems, Components and Equipment, PM-0619-65894, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on June 18, 2019. The materials support NuScales presentation of Chapter 3, Design of Structures, Systems, Components and Equipment, of the NuScale Design Certification Application.
The enclosure to this letter is the nonproprietary version of the presentation titled ACRS NuScale Subcommittee Presentation: FSAR Chapter 3, Design of Structures, Systems, Components and Equipment, PM-0619-65894, Revision 0.
If you have any questions, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-8H12 Michael Snodderly, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Marieliz Vera, NRC, OWFN-8H12
Enclosure:
ACRS NuScale Subcommittee Presentation: FSAR Chapter 3, Design of Structures, Systems, Components and Equipment, PM-0619-65894, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0619-65896
Enclosure:
ACRS NuScale Subcommittee Presentation: FSAR Chapter 3, Design of Structures, Systems, Components and Equipment, PM-0619-65894, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Nonproprietary ACRS NuScale Subcommittee Presentation NuScale FSAR Chapter 3 Design of Structures, Systems, Components and Equipment June 18, 2019 1
PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presenter Marty Bryan Licensing Project Manager Patrick Conley Programs Engineer, Section 3.2, 3.10, 3.11 J. J. Arthur Manager, Structures and Design Analysis, Sections 3.9, 3.12, 3.13 Josh Parker Supervisor, Civil/Structural Analysis, Sections 3.3, 3.4, 3.5, 3.7, 3.8 Storm Kauffman Consultant, Section 3.6 2
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Purpose Provide an overview of FSAR Chapter 3 to the ACRS NuScale Subcommittee 3
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3.2 - Classification of Systems, Structures, and Components
- SSC classified according to
- Seismic category (RG 1.29 R5)
- Quality group (RG 1.26 R4)
- Radwaste classification (RG 1.143 R2)
- SSC classified as
- A1 is safety-related and risk significant
- A2 is safety-related and not risk-significant
- B1 is nonsafety-related and risk-significant
- B2 is nonsafety-related and not risk-significant
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3.10 - Seismic and Dynamic Qualifications of Mechanical and Electrical Equipment
- Addresses seismic and dynamic qualifications of SC-I mechanical and electrical equipment and supports
- Methods and procedures meet RG 1.100 R3 and IEEE 344-2004
- Three COL Items to develop site-specific seismic and dynamic qualification program and equipment qualification database 5
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3.11 - Environmental Qualification of Mechanical and Electrical Equipment
- EQ program complies with DSRS 3.11.
- EQ program includes equipment in 10 CFR 50.49 scope, certain PAM equipment specified in RG 1.97, and active mechanical equipment with safety-related function
- Meets GDC 1, 2, 4, and 23
- Table 3.11-1: harsh equipment list subject to EQ
- Environmental conditions considered include AOOs and normal, accident, and post-accident (See Appendix 3C)
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3.9.1 - Mechanical Systems and Components
- Addresses analysis methods for SC-I components and supports
- Operating condition categories ASME Service Level A through D and Test Conditions apply to Class 1 & 2 components, CNV, supports, RVI, piping and valves inside and outside containment
- Dynamic and static analyses used ANSYS, AutoPIPE, NRELAP5, RspMatch2009, SAP2000, SASSI2010, SHAKE2000, EMDAC, and Simulink
- NuScale technical reports for NPM Seismic Analysis, CVAP, and Short-Term Transient Analysis reviewed in May 16, 2019 ACRS meeting 7
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3.9.3 - ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures
- ASME Code Class 1, 2, and 3 Components are designed and constructed in accordance with Section III the 2013 edition of the ASME Boiler & Pressure Vessel Code.
- Section 3.9.3 defines loading combinations, system operating transients, and stress limits for these components.
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3.9.4 - Control Rod Drive System
- The control rod drive system (CRDS) is composed of:
- a pressure-retaining housing, a control rod drive shaft with a coupling for attaching to the control rod assembly (CRA) hub,
- external electromagnetic coils with cooling loop heat exchangers
- The RCPB pressure boundary parts of the CRDS applicable requirements of ASME BPVC, 2013 Edition,Section III Subsection NB
- The control rod drive mechanisms (CRDMs) provide:
- Means for CRA insertion and rod position indication to the module control system
- Unique CRDS features subject to prototype testing described by Section 4.2.4 9
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3.9.5 - Reactor Vessel Internals
- Reactor Vessel Internals (RVI) composed of subassemblies which:
- Support and align the reactor core system including the CRAs, and include the guide tubes
- RVI channels the reactor coolant from the reactor core to the steam generator (SG) and back to the reactor core
- RVI core support structures and internal structures comply with ASME BPVC Section III, Division 1, Subsection NG.
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3.9.6 - Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints
- ASME OM-2012 Code Edition was used to develop inservice testing requirements.
- Pursuant to 10 CFR 50.55a(z), an Alternate Authorization was used to apply Mandatory Appendix IV (AOV Testing) of the ASME OM-2017 Code Edition for performance assessment testing of power operated valves.
- The IST program contains 39 valves per NPM. This includes 26 HOVs (22 are CIVs), 5 ECCS valves, 2 AOVs, 4 PRVs, and 2 CKVs.
[FSAR Table 3.9-16]
- The Augmented Valve Testing Program (Augmented IST) contains 12 valves per NPM. These valves do not meet the criteria of ISTA-1100, but do have an augmented quality function. [FSAR Table 3.9-17]
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COL Items in Section 3.9
- 12 COL Items to assure compatibility of design with site-specific conditions and equipment, describe programs, and provide test procedures 12 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.12 - ASME Code Class 1, 2, 3 Piping Systems, Piping Components and Associated Supports
- The NuScale design includes less ASME Section III piping than traditional LWRs.
- The NPM does not include large reactor coolant piping. ASME Class 1 piping systems are specified as NPS 2 or smaller.
- Maximum pipe size in the NPM is NPS 12 (Main Steam Lines)
- Snubbers are not used within the NPM.
- Stress analysis has been performed for high-energy lines (>NPS 1) within the NPM to support Section 3.6 (i.e., postulate HELBs)
- NPM piping screened for thermal stratification and thermal oscillations using criteria developed by EPRI.
- The DHRS condensate branch connection to the feedwater line was identified as being potentially susceptible. CFD analysis was performed to demonstrate that stratification does not occur and that the temperature fluctuations in the DHRS line and the associated containment penetration cause thermal stresses that are below the endurance limit for the materials of the piping, welds, and CNV 13 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.12 - ASME Code Class 1, 2, 3 Piping Systems, Piping Components and Associated Supports (contd)
NPM Piping Outside the CNV NPM Piping NPM Inside the CNV 14 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.13 - Threaded Fasteners (ASME Code Class 1, 2, and 3)
- Meet 10 CFR 50.55a
- No code cases per RG 1.84 R36 used
- Meet GDC 1, 4, 14, 30, and 31
- ISI meets 10 CFR 50.55a
- COL Item to provide ISI program for threaded fasteners 15 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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NuScale Plant Structures RWB is SC-II and RW-IIa RXB and CRB are SC-I 16 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.3 - Wind and Tornado Loadings
- Wind Loads
- Design basis severe wind, 3-second gust = 145 mph (ASCE/SEI 7-05)
- Importance factor of 1.15 for RXB, CRB, RWB
- Exposure Category C
- Tornado and Hurricane Loads
- Design basis tornado: 230 mph (RG 1.76 R1)
- Design basis hurricane: 260 mph (RG 1.221 R0)
- Design complies with GDC 2 and 4
- COL Item confirms site specific structures will not collapse and adversely affect RXB or SC-I portion of CRB 17 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.4 - Water level (Flood) Design
- Flooding analysis is conducted on a level-by-level and room-by-room basis for the RXB and CRB for postulated flooding events
- The RXB and CRB flooding analysis consists of the following steps:
- Identification of potential flooding sources
- Identification of rooms/areas that contain equipment subject to flood protection
- Estimation of flood depth in the identified rooms/areas
- Determination of the need for protection and mitigation measures for rooms 18 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.4 - Water Level (Flood) Design (contd)
- Protection from external sources
- Probable Maximum Flood
- 1 foot below grade elevation
- Max groundwater elevation: 2 feet below grade elevation
- There are no dynamic flood loads on the structures given the below grade flood elevation
- The ground water and soil pressure are included in the static and dynamic loads
- Precipitation
- Rainfall rate: 19.4 in/hr and 6.3 inches for 5 minutes
- Roof snow load: 50 psf
- Extreme snow load: 75 psf
- Bounding rain and snow loads are considered in the structural analysis 19 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.4 - Water Level (Flood) Design (contd)
- Interaction of non-SC-I structures with SC-I structures assessed/analyzed to ensure no credible potential
- RXB and CRB flooding analyses protection of safety-related equipment
- Seven COL Items confirm site-specific conditions, programs, locations, and no adverse impact to RXB and SC-I portion of CRB 20 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.5 - Missile Protection
- Design meets GDC 2 and 4
- SC-II RWB also classified RW-IIa (RG 1.143 R2) and designed for same missiles as SC-I structures
- Potential missiles considered
- Internally-generated missiles
- Tornado/extreme winds missiles
- Site proximity missiles
- Aircraft hazards 21 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.5 - Missile Protection (contd)
- No credible missiles inside containment
- Four COL Items to confirm site-specific missile analyses 22 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.5.1.3 Turbine Missile Protection
- NuScale plant designed per RG 1.115 using barrier approach
- Three different missiles analyzed
- Turbine blade, 32.6 pound
- Turbine blade with rotor fragment, 52.6 pound
- Half of last stage turbine rotor disk, ~3,000 pound
- Velocity of missiles based on 3600 revolution per minute turbine speed
- Varying overspeed conditions considered- up to 160% in accordance with RG 1.115 23 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.5.1.3 Turbine Missile Protection (contd)
- Barriers designed per SRP 3.5.3
- Local missile effects
- Penetration
- Scabbing
- Perforation
- Global missile effects
- Reactor Building and Control Building structures analyzed as turbine missile barriers;
- Provide adequate protecting for essential SSCs housed within buildings 24 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design
- Design meets GDC 2 and 10 CFR 50, Appendix S
- CSDRS is design basis for SC-I SSC
- SC-I structures also evaluated using CSDRS-HF with more high frequency content than CSDRS CSDRS CSDRS-HF 1.00 1.00 Acceleration (g) 0.10 0.10 0.01 0.01 0.10 1.00 10.00 100.00 0.10 1.00 10.00 100.00 Frequency (Hz) Frequency (Hz)
Vertical CSDRS Horizontal CSDRS 25 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- NuScale spectra higher and broader than previous design certifications 26 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- Time Histories
- Five seed time histories compatible with the CSDRS
- Separate seed time history compatible with the CSDRS-HF
- Soil Profiles
- Four soil profiles - Soft Soil [Type 11], Firm Soil/Soft Rock [Type 8],
Rock [Type 7], Hard Rock [Type 9]
- Only two SC-I structures: RXB and CRB
- RWB is SC-II and also RW-IIa (high hazard) (RG 1.143 R2)
- Buildings Analyzed using SAP2000, SASSI2010, and ANSYS 27 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- SAP2000
- Used to develop the RXB and CRB finite element models for static analyses
- Consider both uncracked and fully cracked material properties
- NPM submodel is input based on ANSYS model
- Single building models and triple building models developed 28 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- SASSI2010
- The SAP2000 model is converted into SASSI model
- The models have the same node numbers, node coordinates, and finite elements in terms of types, numbers, sectional, and material properties
- Models use the extended subtraction method to perform soil structure interaction analysis
- Interaction nodes are nodes on the seven planes (i.e., six sides [east, west, north, south, top, and bottom])
7P Interaction Nodes (in Red) and the middle plane, of the excavated Cross-Section through RXB soil model 29 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- ANSYS
- Necessary since neither SAP2000 nor SASSI have explicit fluid element formulations
- Used to determine the hydrodynamic pressures on the reactor pool walls and foundation from a Fluid-Structure Interaction analysis
- Used to determine the sloshing wave height
- Used for RXB and CRB stability analyses
- Used to develop thermal and stress analysis models of the RXB
- Obtained bounding rebar and pool liner strains from load combinations that involve thermal and HELB loads 30 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- Subsystem analysis
- Four subsystems evaluated in the DCA
- NPM- Described in the NPM Seismic Technical Report
- RBC- Described in Section 9.1.5
- Fuel Racks- Described in Section 9.1.2
- Bioshield Air flow
- Bioshield
- a nonsafety related, not risk significant SC-II component
- Major functions include fire protection, radiation protection, ventilation, and support personnel access Bioshield Vertical Face Cross Section 31 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.7 - Seismic Design (contd)
- Sensitivity studies
- Multiple sensitivity studies were evaluated such as:
- Empty dry dock
- Less than 12 modules
- Soil separation
- Extended subtraction method to direct method
- Non-vertically propagating shear waves
- Results were shown to be bounded or design basis was modified
- Compared in-structure response spectra and design forces and moments
- Fifteen COL Items to confirm that the site-specific design is bounded by the DCA evaluation 32 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.8.2 - Steel Containment
- The NuScale containment is a steel pressure vessel designed to the requirements of ASME III, Division 1, Subsection NB.
- The CNV has an upper and lower section connected with an approximate 218-inch diameter bolted flange.
- The CNV is housed in the reactor pool within the RXB
- The internal design pressure of 1,050 psia bounds all service level pressures except for hydrostatic test conditions. The external design pressure is 60 psia. The design temperature is 550 deg-F.
- An analysis conforming to the guidance provided in Appendix A of NUREG/CR-6906 was performed to determine the ultimate pressure capacity of the CNV (1240 psia)
- Documented by Technical report TR-0917-56119, CNV Ultimate Pressure Integrity 33 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.8.4 - Design of Category I Structures
- Designed for normal loads, severe environmental loads, extreme environmental loads, and abnormal loads (e.g., high energy pipe break)
- Six COL Items to confirm acceptability of RXB and CRB, describe program for monitoring and maintaining SC-I structures, and evaluate construction elements 34 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.8.4 - Other Seismic Category I Structures (contd)
- 15 Critical Sections for the RXB, and 7 for the CRB
- Perform a safety-critical function
- Are subjected to large stress demands
- Is a feature that is difficult to design or construct
- Are considered to be representative of the structural design
- Selection of walls, slabs, beams, buttresses, pilasters, NPM bay wall, and NPM supports
- Design Evaluation of Critical Sections demonstrate the acceptability of the design
- Compare the demand on the section to available capacity 35 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.8.5 - Foundations
- Stability (i.e., sliding and overturning) is checked per SRP 3.8.5
- Nonlinear sliding analyses were performed for both RXB and CRB.
- Nonlinear overturning also performed for CRB.
- Results demonstrate that the effect of sliding and uplift is small and acceptable for deeply embedded structures 36 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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3.6 - Protection Against Pipe Rupture Effects Key NuScale Differentiators
- Piping small diameter and short compared to typical LWR
- HELB/MELB response is passive
- Containment immersed in pool of water
- CNV designed/fabricated to ASME Class 1 (not a building)
- Containment at a vacuum
- No containment spray or emergency make-up
- Piping inside containment is not insulated
- Insulation outside CNV cannot clog ECCS
- MSS & FWS piping inside containment satisfies LBB
- MSS & FWS piping in containment penetration area is RCS-like
- Essential SSCs inside CNVs qualified to 1050 psia saturated steam
- Congested arrangements
- CIVs and DHRS actuation valves outside CNVs
- NPM moved during refueling / disconnect flanges
- Main control room in building from high energy piping
- Up to 12 NPMs operating in proximity 37 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Locations of Possible Line Breaks
- Analyses divided into 3 regions because of different:
- Environment
- Systems that are high or moderate energy
- Potential target SSCs
- The regions are
- Inside CNV: specific locations and arrangements
- Main steam (MSS) and feedwater (FWS) piping designed to satisfy LBB
- Small amount of other piping, all small bore - no containment penetration area
- Evaluate effects of breaks at terminal ends
- Follow BTP 3-4 B.A.(iii) guidance for determining break locations
- In NPM bay under bioshield: specific locations and arrangements
- Containment penetration area out to weld between last valve & pipe
- Non-mechanistic break of MSS or FWS
- In RXB where piping present: generic break and target locations
- Bound possible breaks
- Future finalization of pipe routing will not affect conclusions 38 PM-0619-65896 T Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Non-Mechanistic Breaks
- MSS and FWS piping generally
- Largest, high-energy piping near the containment boundary
- Single CIV outside containment (GDC 57)
- Piping in other plants usually made of less-corrosion-resistant material (carbon or low-alloy steel), which have greater susceptibility to degradation than stainless steel 39 PM-0619-65896 T Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Break Exclusion 40 PM-0619-65896 T Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Blast Wave Summary
- Assumed instantaneous break opening
- Only steam filled lines generate a blast
- No blast if break opening time more than a few milliseconds
- Small piping reduces mass & energy available
- Accelerated material forms region of high density
- Higher density wave speed exceeds ambient speed of sound
- 3D computational fluid dynamics using CFX V18.0 code
- Verification and validation: 8 test problems
- Simplified plant geometry
- 3D needed to capture geometry of wave propagation / reflection
- Force-time history for
- 3 degasification line cases in CNV
- 3 MSS line cases in RXB
- Blast forces very brief and acceptable 41 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Pipe Whip
- Impact energy depends on fluid conditions & length Small pipes have low energy content Congested areas limit length of whip
- Determine If fluid energy sufficient to form plastic hinge (i.e., thrust force)
Motion of whip in plane based on pipe geometry If essential SSCs are too far away to be struck If nearby SSC serves as barrier Impact force if pipe whip impact occurs 42 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Pipe Whip-caused Secondary HELB SRP 3.6.2 - an unrestrained, whipping pipe need not be assumed to cause ruptures or 60 thick through-wall wall cracks in pipes of equal or larger NPS with equal or greater wall thickness.
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Pipe Whip Summary
- Impact force depends on Thrust force at break exit (NuScale force ~5% of large PWRs)
Short length of moment arm to plastic hinge Distance through which pipe whips
- Know piping arrangements in CNV
- Conservative assumptions in RXB Energy absorbed by pipe on impact
- Two types of targets Metallic components/pipes Concrete in RXB: penetration (Young (Sandia) formula) & spalling
- Conclusions Same or larger size pipe does not break or crack (SRP 3.6.2)
Targets in CNV except RPV, CNV, & CRDMs out of range Concrete wall penetration minor & wall thickness prevents spalling 44 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Jet Impingement
- Impingement pressure energy depends on fluid conditions, spreading of jet, & distance to target Small pipes have low energy content
- Jet behavior phase dependent Steam: limited spread 2-phase: NUREG-2913
- Determine Pressure threshold for SSC damage Zone of influence (ZOI)
Jet spreading half-angle If nearby SSC serves as barrier If essential SSCs are close enough to be above pressure threshold 45 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Jet Shape & Pressure Overview
- Inside CNV (NPS 2 only - no longitudinal breaks)
- ZOI: forward facing hemisphere out to threshold
- Pressure: half-angle of 30 to 5 L/D, then 10 Greater spreading in low pressure ambient More conservative than Modified Moody Model 45 half-angle
- In RXB ZOI: anywhere because piping arrangements not finalized
- Assumes target SSC is 2 L/D from break exit Pressure: full break exit pressure (no expansion) 46 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Jet Dynamic Amplification
- SRP 3.6.2 identifies concern that jet impingement on target can induce resonant condition
- Numerous research papers show resonance requires
- Axisymmetric jet
- A phase lock in boundary shear layer at jet exit
- Occurs in gas jets but not seen in steam or 2-phase jets
- Resonance will not occur due to jet impingement
- Condensable fluid in jet attenuates oscillations
- Distorted exit geometry violates axisymmetry
- Absence of a large, flat impingement surface sufficiently close and perpendicular to the jet axis eliminates return of acoustic energy
- Instability of jet exit separation and angle prevents phase lock
- Presence of obstacles or intersecting flow disrupt jet axisymmetry
- Frequency mismatch with structures 47 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Jet Impingement Summary
- Impingement pressure depends on Thrust force at break exit; NuScale thrust force small Spreading angle of underexpanded jet Distance to target SSCs
- Two types of targets in CNV (no thermal insulation stripping)
Metallic components/pipes Safety-related components (e.g., valves, CRDMs) go to safe position on loss of power or pressure holding them open Cables
- No dynamic amplification
Conclusions:
No safety-related SSCs within range in CNV RXB structure satisfactory by analysis bounding future final design 48 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Pressurization Summary
- Under bioshield: HELBs limited to nonmechanistic breaks of MSS and FWS and to leakage cracks Passive vent to pool room limits pressure and temperature Envelope of results used to create EQ temperature envelope Structural limit of 1 psid met
- In RXB: various HELBs limiting in different areas Vent paths limit room pressures Structural limit of 3 psid met 49 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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PRHA Conclusions
- Break Locations limited by break exclusion and LBB
- Blast - 3D CFD determined loading is low and brief
- Pipe Whip - SSCs out of range or can withstand impact
- Jet Behavior - components and cabling have good resistance to damage, and conservative modeling showed no adverse effects
- Subcompartment Pressurization - GOTHIC analysis used to determine need for venting, to set EQ envelope, and to show structural limits met
- COL items confirm final arrangements satisfy criteria Analyses of the various external effects of HELBs confirmed acceptability of the NuScale design 50 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Leak Before Break
- SA-312 TP304/304L dual-certified stainless steel
- Corrosion, erosion, fatigue, other failure mechanisms evaluated
- Margin of 10 for leak rate and 2 for flaw size
- Leak detection by change in
- 1) CNV pressure
- 2) CES sample vessel condensate level 51 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Acronyms COL combined license NPM NuScale Power Module CNV containment vessel PAM post-accident monitoring CRB Control Building PDC principal design criteria CSDRS certified seismic design response spectra PMF probable maximum flood CSDRS-HF CSDRS - high frequency PMP probable maximum precipitation CVAP Comprehensive Vibration Assessment Program RCP reactor coolant pump DBT design basis tornado RCPB reactor coolant pressure boundary ECCS emergency core cooling system RCS reactor coolant system EQ environmental qualification RG Regulatory Guide FIV flow-induced vibration RPV reactor pressure vessel FSAR Final Safety Analysis Report RVI reactor vessel internals GDC General Design Criteria RWB Radioactive Waste Building HE high energy RXB Reactor Building HELB high-energy line break SC-I Seismic Category I ISI in-service inspection SC-II Seismic Catogory II LOCA loss-of-coolant accident SSC structures, systems, and components ME moderate energy 52 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 53 PM-0619-65896 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Presentation to the ACRS Subcommittee NuScale Design Certification Application Review Safety Evaluation Report Chapter 3: DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS Project Manager: Marieliz Vera June 18, 2019
Staff Review Team Technical Staff Renee Li, Senior Mechanical Engineer Tuan Le, Mechanical Engineer Yiu Law, Mechanical Engineer Alexander Tsirigotis, Mechanical Engineering Yuken Wong, Senior Mechanical Engineer Jason Huang, Mechanical Engineer Nick Hansing, Mechanical Engineer Michael Breach, Mechanical Engineer Kaihwa Hsu, Senior Mechanical Engineer Tom Scarbrough, Senior Mechanical Engineer Jorge Cintron- Rivera, Electrical Engineer Sheila Ray, Senior Electrical Engineer Vaughn Thomas, Structural Engineer NuScale Chapter 3 - Design of Structures, June 18, 2019 2 Systems, Components, and Equipment
Staff Review Team Technical Staff Robert Roche-Rivera, Structural Engineer Sunwoo Park, Structural Engineer Pravin Patel, Structural Engineer Ata Istar, Structural Engineer George Wang, Structural Engineer Maryam Khan, Structural Engineer Manas Chakravorty, Senior Structural Engineer Bhagwat Jain, Senior Structural Engineer John Ma, Senior Structural Engineer Edward Stutzcage, Health Physicist Raul Hernandez, Reactor System Engineer Chang Li, Senior Reactor System Engineer Nicholas Mcmurray, Materials Engineer NuScale Chapter 3 - Design of Structures, June 18, 2019 3 Systems, Components, and Equipment
Staff Review Team Technical Staff John Honcharik, Senior Materials Engineer Eric Reichelt, Senior Materials Engineer Seshagiri Tammara, Physical Scientist Vladimir Graizer, Seismologist John Wu, Mechanical Engineer Jim Strnisha, Mechanical Engineer Project Managers Gregory Cranston, Lead Project Manager Marieliz Vera, Chapter Project Manager NuScale Chapter 3 - Design of Structures, June 18, 2019 4 Systems, Components, and Equipment
Sections with no Open Items 3.2.1 - Seismic Classification 3.2.2 - System Quality Group Classification 3.3.1 - Severe Wind Loads 3.3.2 - Extreme Wind Loads (Tornado and Hurricane Loads) 3.4.1 - Internal Flood Protection for Onsite Equipment Failure 3.4.2 - Analysis Procedures 3.5.1.1 & 3.5.1.2 - Internally Generated Missiles (Outside and Inside Containment) 3.5.1.4 - Missiles Generated by Tornadoes and Extreme Winds 3.5.1.5 - Site Proximity Missiles (Except Aircraft) 3.5.1.6 - Aircraft Hazards 3.5.2 - Structures, Systems, and Components to be Protected from Externally Generated Missiles NuScale Chapter 3 - Design of Structures, June 18, 2019 5 Systems, Components, and Equipment
Sections with no Open Items (Contd) 3.6.1 - Plant Design for Protection against Postulated Piping Failures in Fluid Systems Outside Containment 3.7.1 - Seismic Design Parameters 3.7.4 - Seismic Instrumentation 3.8.5 - Foundations 3.9.1 - Special Topics for Mechanical Components 3.10 - Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 3.12 - ASME BPV Code Class 1, 2, and 3 Piping Systems and Associated Support Design 3.13 - Threaded FastenersASME BPV Code Class 1, 2, and 3 NuScale Chapter 3 - Design of Structures, June 18, 2019 6 Systems, Components, and Equipment
3.6.3 - Leak Before Break Evaluation Procedures Eric Reichelt NuScale Chapter 3 - Design of Structures, June 18, 2019 7 Systems, Components, and Equipment
Section 3.6.3 - Leak Before Break
- NuScale DCD requested to apply Leak Before Break (LBB) methodology to Main Steam Piping and Feedwater Piping Systems.
- Unique aspect for NuScale is curved pipe system, and making sure fabrication (pipe bending) limits cold-working to an acceptable limit.
- The methods and criteria to evaluate LBB are consistent with the guidance in SRP 3.6.3, and NUREG-1061, Volume 3.
NuScale Chapter 3 - Design of Structures, June 18, 2019 8 Systems, Components, and Equipment
Section 3.6.3 - Leak Before Break
- Reviewed applicable NuScale DCD subsections in Section 3.6.3.
- Reviewed DCD references for applicability and use.
- Held public meetings with NuScale staff about technical issues and RAIs leading to proposed DCD markups.
- The technical issues and response by NuScale to RAIs were acceptable and were therefore closed.
- The staff is currently reviewing the methodology and will perform a confirmatory analysis on the FWS proprietary information and data provided by the applicant on June 8, 2019.
- The staff will review the methodology and will perform a confirmatory analysis on the MSS when NuScale provides the proprietary information and data scheduled for July 15, 2019.
- Based on results of the confirmatory analysis, this is a confirmatory item.
NuScale Chapter 3 - Design of Structures, June 18, 2019 9 Systems, Components, and Equipment
Section 3.5.1.3 - Turbine Missile and Section 3.5.3 -
Barrier Design Procedures John Honcharik BP Jain NuScale Chapter 3 - Design of Structures, June 18, 2019 10 Systems, Components, and Equipment
3.5.1.3 Turbine Missiles Regulatory Basis and Use of Barriers
- 10 CFR Part 50, Appendix A, GDC 4, requires SSCs important to safety to be appropriately protected against environmental and dynamic effects, including the effects of missiles that may result from equipment failure.
- Turbine generator rotor shafts are unfavorably oriented such that the RXB and CRB are within the turbine low-trajectory hazard zone.
- To meet the requirements of 10 CFR Part 50, Appendix A, GDC 4, NuScale proposes to use installed or existing structures for protecting essential SSCs that meet the acceptance criteria in DSRS Section 3.5.3.
NuScale Chapter 3 - Design of Structures, June 18, 2019 11 Systems, Components, and Equipment
3.5.1.3 Turbine Missiles Determining Turbine Missile Parameter on Barriers
- Open Item 03.05.01.03-1 : The staff determined that NuScale had not used the full spectrum of turbine missiles (size, weight and speed) which includes up to half of the last stage of the rotor in the barriers analysis.
- NuScale provided additional information in June, and the staff is currently reviewing the information on spectrum of turbine missile parameters.
- Section 3.5.3 addresses the verification of the barriers used for turbine missile protection.
NuScale Chapter 3 - Design of Structures, June 18, 2019 12 Systems, Components, and Equipment
Section 3.5.3 - Turbine Missile Barrier Open Item 03.05.03-1
- Postulated Turbine Missile spectrum - an open item 3.5.1.3-1
- 1. Turbine Blade : Weight: 32.6 lbs ;Velocity 784 mph,
- 2. Turbine Blade with rotor fragment : Weight 52.6 lbs ; velocity 996 mph
- 3. Half of last stage rotor: weight 3079 lbs; velocity 350 mph
- Staff Guidance in SRP 3.5.3 used to review Barrier Design Procedures for the local and overall damage.
- Acceptance Criteria - Concrete barriers should be thick enough to prevent backface scabbing.
- Staff is required to make the following review findings.
The procedures used for barriers design for the impact of design basis turbine missiles are acceptable.
Information presented in the DCD provides a reasonable assurance that Reactor and Control building walls provide adequate protection to essential SSCs from the impact of the design basis turbine missiles.
- Complete and technically sufficient information not available in DCD for staff to make safety findings. NuScale submitted supplementary info in June.
- Staff is reviewing NuScales response to barrier design issues.
NuScale Chapter 3 - Design of Structures, June 18, 2019 13 Systems, Components, and Equipment
Section 3.5.3 - Turbine Missile Barrier Open Item 03.05.03-1
- Overview of Key Barrier Design Issues:
- NuScale used FEM procedure for penetration depth calculation instead of SRP 3.5.3 empirical NDRC formula.
- Local and global damage assessment for the full turbine missile spectrum was not provided
- FEM procedure is not previously reviewed by the staff for high speed turbine missile impact Staff to review Validation and Verification of the FEM approach Computer Program - not reviewed for turbine missiles Benchmarking of computer program against relevant test data required
- Penetration depth from the impact of deformable missile is significantly reduced to almost 1/3 of that from the missile impact (e.g., reduced to 20 from 54.6)
Validation by test and/or analytical results required NuScale Chapter 3 - Design of Structures, June 18, 2019 14 Systems, Components, and Equipment
Section 3.5.3 - Turbine Missile Barrier Conclusions
- Staff will review the barrier design information and conduct audits as required to insure compliance with the regulatory requirements
- NuScale application to demonstrate compliance with regulatory requirements for barrier design NuScale Chapter 3 - Design of Structures, June 18, 2019 15 Systems, Components, and Equipment
Sections 3.7.2 - Seismic System Analysis and 3.7.3 -
Seismic Subsystem Analysis Sunwoo Park NuScale Chapter 3 - Design of Structures, June 18, 2019 16 Systems, Components, and Equipment
Sections 3.7.2 and 3.7.3 -
Overview
- Review Scope
- 3.7.2 Seismic System Analysis - Seismic Category I structures:
Reactor Building, Control Building
- 3.7.3 Seismic Subsystem Analysis - example, Bioshield
- Phase-2 SER had 4 Open Items
- 2 OIs resolved to date
- 2 OIs require additional information and evaluation
- 1 in Section 3.7.2 - to be discussed in the following slides
- 1 in Section 3.7.3 - to be discussed in the following slides NuScale Chapter 3 - Design of Structures, June 18, 2019 17 Systems, Components, and Equipment
Section 3.7.2 - Seismic System Analysis Open Item 03.07.02-1
- Seismic load demands for NPM-RXB interface supports are determined by analysis of the RXB SASSI model and NPM ANSYS model
- Applicant expanded analysis cases to include 130% NPM nominal stiffness
- Applicant adopted new methodology for modeling hydrodynamic mass of pool water
- Applicant provided information from new analysis, which is currently in staff evaluation
- Applicant is to provide information about seismic loads on pool walls from new analysis NuScale Chapter 3 - Design of Structures, June 18, 2019 18 Systems, Components, and Equipment
Section 3.7.3 - Seismic Subsystem Analysis Open Item 03.07.03-1
- Bioshields are Seismic Category II concrete covers placed on top of each NPM as additional radiological barrier
- Bioshield is removed during NPM refueling, and removed Bioshield is stacked on top of adjacent Bioshield
- During December 2018 audit, staff identified issues concerning seismic design of stacked Bioshields
- Applicant revised the design approach to address identified issues
- Applicant provided information on new seismic analysis and design of Bioshields, which is currently in staff evaluation NuScale Chapter 3 - Design of Structures, June 18, 2019 19 Systems, Components, and Equipment
Sections 3.7.2 and 3.7.3 -
Conclusion
- Applicant is undertaking actions for timely resolution of the open items.
- NuScale DCA demonstrates compliance with regulatory requirements for seismic system and seismic subsystem analysis.
NuScale Chapter 3 - Design of Structures, June 18, 2019 20 Systems, Components, and Equipment
Sections 3.8.4 - Other Seismic Category I Structures Robert Roche-Rivera NuScale Chapter 3 - Design of Structures, June 18, 2019 21 Systems, Components, and Equipment
Section 3.8.4 - Overview
- Review Scope
- DCA Tier 2 Section 3.8.4, Other Seismic Category I Structures
- DCA Tier 2 Appendix 3B, Design Reports and Critical Section Details
- Reviewed DCA Tier 2 Section 3.8.4, associated Appendix 3B, and the listed tables and figures in accordance with the DSRS Section 3.8.4 acceptance criteria.
- Held bi-weekly public meetings with the applicant to discuss technical issues and resolutions to RAIs
- Conducted audit of design reports supporting the information provided in the DCA
- Phase 2 SER had 5 Open Items
- All OIs have been resolved NuScale Chapter 3 - Design of Structures, June 18, 2019 22 Systems, Components, and Equipment
Section 3.8.4 - Other Seismic Category I Structures
- Reviewed and compared the applicants design procedures and associated results to the applicable code and or standard acceptance criteria and allowable
- Loads and load combinations considered and structural capacity determinations are in accordance with the applicable code and or standard
- Displacement and strain results meet the applicable code and or standard allowable
- Structural capacities are greater than design basis demands (i.e. D/C ratios less than 1)
- Concluded that the applicants methods for demonstrating the design adequacy of the structures are consistent with the NRCs regulatory requirements NuScale Chapter 3 - Design of Structures, June 18, 2019 23 Systems, Components, and Equipment
Section 3.8.4 - Other Seismic Category I Structures (Contd)
- Open Item 3.8.4-1: RAI 8971, Question 3.8.4-13
- Staff finding:
- Design evaluation for temperature (operating and accident) and accident pressure demands not provided.
- Resolution:
- Applicant performed design evaluations of temperature and accident pressure demands and other concurrent loads for the RXB
- The results demonstrated that the concrete, rebar, and liner strains are below the strain limits NuScale Chapter 3 - Design of Structures, June 18, 2019 24 Systems, Components, and Equipment
Section 3.9.4 - Control Rod Drive Systems Nicholas Hansing NuScale Chapter 3 - Design of Structures, June 18, 2019 25 Systems, Components, and Equipment
Section 3.9.4 Key Design Considerations and Features
- Pressure housing/ electromagnetic components very similar to existing fleet (Pressure housings designed & constructed to ASME BPV Code Class 1 requirements)
- Long drive shaft and remote disconnect mechanism are unique
- Design standards and testing programs were emphasized in the review Drive Shaft
- Additional design requirements are specified in the DCA, such as Service Level loading combinations Remote Disconnect Mechanism
- Applicant confirmed the remote disconnect coil is always deenergized during normal operations and remains in this state during a reactor trip
- Mechanism was tested during Key Feature Mock-Up Testing for its full design life of 150 cycles (5x estimated cycles expected) with satisfactory performance NuScale Chapter 3 - Design of Structures, June 18, 2019 26 Systems, Components, and Equipment
Section 3.9.4 Drop Testing
- Appendix B compliant testing with prototypical components
- Staff independently verified dimensions of design documents to as-built dimensions of test facility - important dimensions like diametrical gap were consistent
- 1 misalignment possible at each interface- testing bounded manufacturing tolerances &
seismic displacements
- 14 configurations used, number of variables reduced using fabrication constraints &
testing worst-case scenarios - tested displacement exceeded the maximum expected displacement by more than a factor of two
- The most limiting drop (maximum displacement and longest drop time) was bounded by the performance assumed in the safety analysis for control rod drop time.
Operability Assurance Program
- Includes performance testing, stability testing, endurance testing, and production testing.
- Will be completed by COL applicant,
- DC applicant has provided overview of program in DC application with a proposed COL Item to implement the program and provide a summary of the testing program and results.
NuScale Chapter 3 - Design of Structures, June 18, 2019 27 Systems, Components, and Equipment
Section 3.9.6 - Functional Design and Qualification, and Preservice Testing and Inservice Testing of Pumps, Valves, and Dynamic Restraints Thomas Scarbrough NuScale Chapter 3 - Design of Structures, June 18, 2019 28 Systems, Components, and Equipment
Section 3.9.6
- NRC staff reviewed functional design and qualification, and preservice testing (PST) and inservice testing (IST) program description for NuScale safety-related valves.
- DCA provides full description of PST/IST programs with SER Confirmatory Items.
- First-of-a-kind (FOAK) Emergency Core Cooling System (ECCS)
Valves and Containment Isolation Valves (CIVs).
- ECCS Valve Design Demonstration Testing in 2019.
NuScale Chapter 3 - Design of Structures, June 18, 2019 29 Systems, Components, and Equipment
ECCS Valve Design
- 3 Reactor Vent Valves (RVVs) and 2 Reactor Recirculation Valves (RRVs) allow natural circulation for emergency core cooling.
- Each RVV and RRV has FOAK design arrangement of a main valve, inadvertent actuation block (IAB) valve, solenoid trip valve, and solenoid reset valve connected by hydraulic tubing.
- Proprietary design and operation of IAB valve can be discussed in a closed session.
- ECCS Valve Design Demonstration Testing to satisfy 10 CFR 50.43(e) initiated June 2019.
- Ongoing NRC audit.
NuScale Chapter 3 - Design of Structures, June 18, 2019 30 Systems, Components, and Equipment
Functional Design and Qualification
- DCA specifies ASME Standard QME-1-2007 for qualification of safety-related valves as accepted with conditions in RG 1.100 (Revision 3).
- ITAAC acceptance criteria for functional qualification of safety-related valves require Qualification Report (specified in QME-1 standard).
- Safety-Related Valve Design Specification audits conducted in 2017 and 2018 with follow-up items for close-out in 2019.
- No safety-related motor-operated valves, pumps, or snubbers.
NuScale Chapter 3 - Design of Structures, June 18, 2019 31 Systems, Components, and Equipment
Containment Isolation Valve Design
- Hydraulic-operated actuators with ball valves.
- 16 Primary System Containment Isolation Valves (PSCIVs) have FOAK design with two actuators and ball valves for same valve body.
- 6 Secondary System Containment Isolation Valves (SSCIVs) have one actuator and ball valve.
- NRC audit in 2018 with follow-up items for close-out in 2019.
NuScale Chapter 3 - Design of Structures, June 18, 2019 32 Systems, Components, and Equipment
Section 3.6.2 - Introduction Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping Renee Li NuScale Chapter 3 - Design of Structures, June 18, 2019 33 Systems, Components, and Equipment
Section 3.6.2
- This review is to ensure that NuScales design provides adequate protection against the effects of the postulated pipe ruptures and meet the GDC 4 requirement.
- The review concentrated on those areas that were outside the guidance of BTP 3-4
- Two isolation valves outside containment for penetration piping
- Bolted connection of the ECCS Valves to the RV head
- Reviewed TR-0818-61384-P, Revision 1 that addresses applicants pipe rupture hazards analysis and the results.
NuScale Chapter 3 - Design of Structures, June 18, 2019 34 Systems, Components, and Equipment
Break Exclusion Areas for NuScale Plant Design
- NuScale two isolation valves outside containment configuration deviates from the BTP 3-4 containment penetration area.
- Evaluation will be provided in the Chapter 6 presentation.
- Applicability of break exclusion to the containment penetration areas.
- 1. Design stress and fatigue limit criteria for the piping segments within the NuScales break exclusion area are consistent with BTP 3-4 guidelines and the associated calculation results are within the relevant BTP 3-4 stress and fatigue limits for break exclusion.
- 2. Augmented 100 percent volumetric in-service examination requirement for all the welds within the break exclusion areas is consistent with BTP 3-4 staff guidelines.
- The NRC staff found the above applicants justification for its application of break exclusion areas acceptable.
NuScale Chapter 3 - Design of Structures, June 18, 2019 35 Systems, Components, and Equipment
Break Exclusion of RVV and RRV Bolted Connection
- Applicability of break exclusion to the ECCS Valve connection to the RPV:
- 1. Bolting stress design criteria per ASME BPV Section III, NB-3230 meet the intent of BTP 3-4 stress acceptance guidance for typical piping system
- 2. Cumulative usage factor of 1.0 is acceptable ASME Section III, NB-3230.3(c) fatigue criteria applies a strength reduction factor of no less than 4.
Bolted connections are not susceptible to the phenomena (e.g.,
unexpected modes of operation vibration, stress corrosion cracking, water hammer,etc.) that might adversely affect fatigue evaluations.
- Basis for recommending CUF of 0.1 in BTP 3-4.
NuScale Chapter 3 - Design of Structures, June 18, 2019 36 Systems, Components, and Equipment
Break Exclusion of RVV/RRV Bolted Connection (Cont.)
- 3. Augmented fabrication and in-service examination requirements.
In-service inspection, Ultrasonic examination of the bolt at least once every 10 years.
- 4. Bolting design adopted NUREG-1339 guidelines with highly SCC resistant material (Alloy 718)
- 5. Highly sensitive leakage detection system (sensitive to leak rate as low as 0.001 gallon per minute)
- NRC staff found that the applicants justification provides reasonable assurance
- NRC audit of the applicants stress and fatigue calculations to close this SER Open Item.
NuScale Chapter 3 - Design of Structures, June 18, 2019 37 Systems, Components, and Equipment
3.8.2 - Steel Containment, 3.9.3 - ASME BPV Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3.9.5 - Reactor Pressure Vessel Internals Alexander Tsirigotis NuScale Chapter 3 - Design of Structures, June 18, 2019 38 Systems, Components, and Equipment
Design Stress/Fatigue Analyses
- During a Design Specification audit, stress analyses and fatigue evaluations of the RPV, CNV and RVI were not available.
- SECY 90-377 establishes position that this level of detail should be complete at DC stage.
- Open Item
- Stress analyses are being updated.
- NuScale evaluated fatigue locations
- Selected most critical locations to conduct ASME BPV Code,Section III fatigue analysis
- All locations scheduled to be available for audit by end of July NuScale Chapter 3 - Design of Structures, June 18, 2019 Systems, Components, and Equipment 39
RCS/CNV Pressure Boundary
- ASME BPV Code,Section III requires surface examination of valve body
- NRC staff questioned adequate level of quality for design
- NRO office accepting ASME BPV Code,Section III as acceptable level of quality NuScale Chapter 3 - Design of Structures, June 18, 2019 40 Systems, Components, and Equipment
ECCS Trip/Reset Solenoid Penetration NuScale Chapter 3 - Design of Structures, June 18, 2019 41 Systems, Components, and Equipment
Abbreviations ACRS Advisory Committee on Reactor Safeguards AOV Air-Operated Valve ASME American Society of Mechanical Engineers BPV Boiler & Pressure Vessel BTP Branch Technical Position CIV Containment Isolation Valve COF Coefficient of Friction COL Combined License CNV Containment Vessel CRB Control Building DC Design Certification DSRS Design Structure Response Spectra DCA Design Certification Application ECCS Emergency Core Cooling System FEM Finite Element Method FWS Feedwater Piping NuScale Chapter 3 - Design of Structures, June 18, 2019 42 Systems, Components, and Equipment
Abbreviations FOAK First of a Kind FSAR Final Safety Analysis Report HOV Hydraulic-Operated Valve IAB Inadvertent Actuation Block IST Inservice Testing ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LBB Leak Before Break MSS Main Steam Piping MOV Motor-Operated Valve NDRC National Defense Research Committee NRO NRC Office of New Reactors OM Operation and Maintenance PST Preservice Testing QA Quality Assurance RAI Request for Additional Information RG Regulatory Guide RCS Reactor Coolant System NuScale Chapter 3 - Design of Structures, June 18, 2019 43 Systems, Components, and Equipment
Abbreviations RPV Reactor Pressure Vessel RRV Reactor Recirculation Valve RVV Reactor Vent Valve RXB Reactor Building SER Safety Evaluation Report SRP Standard Review Plan SSCs Structures, Systems, and Components NuScale Chapter 3 - Design of Structures, June 18, 2019 44 Systems, Components, and Equipment
LO-0619-65928 June 14, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation, NuScale FSAR Chapter 6, Engineered Safety Features, PM-0619-65926, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee meeting on June 18, 2019. The materials support NuScales presentation of Chapter 6, Engineered Safety Features, of the NuScale Design Certification Application.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Presentation, NuScale FSAR Chapter 6, Engineered Safety Features, PM-0619-65926, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Rebecca Norris at 541-602-1260 or at rnorris@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-7H4 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Omid Tabatabai, NRC, OWFN-8H12
Enclosure:
ACRS Subcommittee Presentation, NuScale FSAR Chapter 6, Engineered Safety Features, PM-0619-65926, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-0619-65928
Enclosure:
ACRS Subcommittee Presentation, NuScale FSAR Chapter 6, Engineered Safety Features, PM-0619-65926, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Nonproprietary ACRS Subcommittee Presentation NuScale FSAR Chapter 6 Engineered Safety Features June 18, 2019 1
PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
Template #: 0000-21727-F01 R5
Presentation Team Daniel Lassiter Engineer, Mechanical Design Zack Houghton Manager, Mechanical Design Engineering Rebecca Norris Supervisor, Licensing 2
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Chapter 6 - Engineered Safety Features Section Title 6.1 Engineered Safety Feature Materials Containment System 6.2
- Performance analysis results presented with FSAR Ch15 Emergency Core Cooling System 6.3
- Additional figures will be provided in closed session 6.4 Control Room Habitability 6.5 Fission Product Removal and Control Systems Inservice Inspection and Testing of Class 2 and 3 Systems and 6.6 Components 6.7 Main Steamline Isolation Valve Leakage Control System 3
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Objective Provide a technical summary of NuScale engineered safety features. All containment peak pressure and temperature analyses related to 6.2.1 and 6.2.2 will be presented with Chapter 15 on Wednesday.
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6.1 - Engineered Safety Feature Materials
- Components compatible with environmental conditions of normal operation, maintenance, testing, and accidents
- 60-yr component design life
- Grade SA-965, FXM-19 material provided in the core vicinity
- Demonstrates good resistance to neutron embrittlement
- No protective coating allowed on the containment vessel (CNV) or any components within CNV 5
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6.2.1 - Containment Functional Design
- CNV Primary Functions
- Enclosure of the RPV, RCS and associated components
- Containment of fission product releases
- Containment of the postulated mass and energy releases inside containment
- Operation of the ECCS by the retention of reactor coolant and the transfer of sensible and core heat to the UHS
- Heat removal capability to maintain peak P/T less than design (GDC-50)
- Reduction of peak P/T to < 50% within 24 hr (GDC-38)
- CNV peak P/T analysis discussed with Chapter 15 6
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NuScale Containment Overview
- Designed and stamped as Class 1 ASME vessel
- 1050 psia design pressure
- 550 F design temperature
- Located directly within the ultimate heat sink
- SA-508 Grade 3, Class 2 upper containment material
- FXM-19 lower containment material
- Piping and electrical penetrations located above pool level
- ~15 shell diameter 7
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Containment Class Breaks 8
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6.2.1 - COL Items Item Number Description COL Item 6.2-3 Perform an analysis that, in consideration of the as-built containment internal free volume, demonstrates that containment design pressure and temperature bounds containment peak accident pressure and temperature. The evaluation value for containment internal free volume must include margin to address the complex shapes of internal structures and components and manufacturing processes.
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6.2.2 - Containment Heat Removal
- Pool water level just below CNV upper head during normal operation (> 30 days cooling volume)
- CNV steel wall allows direct, passive heat transfer to pool
- For limiting peak pressure case, CNV pressure reduced to < 50% of peak within 2 hr
- Exemption from GDC 40 requested - testing of containment heat removal system justified in FSAR Section 3.1.4
- CNV removes accident heat from NPM as accident progresses; no operator action required 10 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.4 - Containment Isolation System
57 with the following exemptions requested:
- Both automatic CIVs located outside of the CNV for GDC 55 & 56 penetrations
- Closed system piping both inside and outside of the CNV for GDC 57 penetrations (DHRS)
- Piping between CNV and CIV designed to break exclusion zone criteria
- Primary system containment isolation valves (PSCIVs) constructed and stamped as ASME Class 1 valves
- Secondary system containment isolation valves (SSCIVs) constructed and stamped as ASME Class 2 valves
- CIVs hydraulically operated, fail closed 11 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Primary CIV Arrangement 12 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.4 - Containment Isolation System (Cont)
- Each module has two separate dedicated hydraulic control skids, located in separate areas of the Reactor Building (RXB)
- Each hydraulic control skid operates one division of valves 13 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Actuator Hydraulic Schematic 14 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.5 - Combustible Gas Control
- Containment integrity analyzed for effects of complete hydrogen burn and detonations
- Exemption requested from10 CFR 50.44(c)(2) requirement for inerted atmosphere or limited hydrogen concentration
- Design provides a mixed CNV volume during design basis events (DBEs) and beyond design basis events (BDBEs)
- No sub-compartments: design relies on natural convection 15 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.5 - Combustible Gas Control (cont)
- Equipment provided for monitoring in-containment hydrogen concentration - capable of continuously monitoring hydrogen gas concentrations after DBEs or BDBEs
- Hydrogen and oxygen analyzers provided in the process sampling system
- Monitoring equipment performance requirements still under discussion with NRC staff
- Design includes connections for cleanup equipment to scrub CNV atmosphere prior to release 16 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.6 - Containment Leakage Testing
- Follows 10 CFR 50, Appendix J for Type B and Type C testing
- Design supports exemption from GDC 52 (Type A ILRT) testing (see TR-1116-51962, Rev 0 and FSAR 6.2.6.1)
- Fabrication and testing similar to reactor pressure vessel (RPV) requirements
- All known leakage pathways tested through local leak rate testing
- Comprehensive inservice inspection (ISI) meets ASME Class 1 requirements
- CNV hydrostatically tested 17 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.6 - Containment Leakage Testing (cont)
- NuScale commitments made to support the GDC 52 exemption include:
- CNV preservice design pressure leakage test described by FSAR Section 6.2.6.5.2 (ITAAC #23)
- Demonstration that the as-built containment maintains flange contact pressure at accident temperature concurrent with peak pressure conditions (COL Item 6.2-2)
- Containment flange bolting preload verification 18 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.2.6 - COL Items Item Number Description COL Item 6.2-1 Develop a containment leakage rate testing program that will identify which option is to be implemented under 10 CFR 50, Appendix J.
Option A defines a prescriptive-based testing approach whereas Option B defines a performance-based testing program.
COL Item 6.2-2 Verify that the final design of the containment vessel meets the design basis requirement to maintain flange contact pressure at accident temperature, concurrent with peak accident pressure.
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6.2.7 - Fracture Prevention of CNV
- CNV meets relevant parts of GDC 1, 16, 51 (thus, ASME Section III, Subsection NB)
- Ferritic CNV pressure boundary meet ASME Section III, Subsection NB fracture toughness requirements
- Lower CNV made of austenitic stainless steel (grade FXM-19) which is not subject to neutron embrittlement at CNV fluence levels 20 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.3 - Emergency Core Cooling System (ECCS)
- Cools core when it cannot be cooled by other means, such as loss of coolant accidents
- Main components are pilot-actuated valves
- 3 reactor vent valves (RVV)
- 2 reactor recirculation valves (RRV)
- Actuation signals CNV Level
- high CNV level Setpoint
- loss of AC power for 24 hr 21 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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EmergencyNuScales ECCS Core Cooling System Operation (contd)
ECCS valves open Coolant is vented into containment 22 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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EmergencyNuScales Core Cooling ECCS System Operation(contd)
Containment vessel is surrounded by UHS Steam condenses and liquid collects in containment vessel This continues until As RPV level lowers in the containment vessel level downcomer region, rises above RRVs containment vessel level rises 23 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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EmergencyNuScales ECCS Core Cooling System Operation (contd)
Coolant flows from Coolant level the containment remains above the vessel into the RPV core, and containment level Flows to bottom of remains above the RPV and back to the RRVs core Heated by the core Heat is transferred
- Core to coolant Flows up to top of
- Coolant to RPV containment vessel Coolant again exits
- Containment through RVVs vessel to UHS 24 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Emergency Core Cooling System (contd)
- Only 2 RVVs and 1 RRV required for operation
- Core stays covered for all design basis events
- ECCS does not require operator action or support from non-safety systems. Operators can manually actuate ECCS from control room.
- Operator alerted if a valve fails to move to its safety position
- Engineered safety features actuation system (ESFAS) design prohibits manual override of ESFAS signal
- Components form part of reactor coolant pressure boundary (RCPB),
ASME Class 1 components
- Auxiliary ECCS Function
- Low temperature overpressure (LTOP) protection for RPV
- Module protection system (MPS) actuation logic based on pressure-temperature limits for RPV 25 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Emergency Core Cooling System (contd)
- Simplified main valve and pilot actuator assembly configuration Containment Reactor Vessel Vessel S Reset P IAB S
-Discharge to containment (RVV)
-Condensate return from containment (RRV) 26 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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ECCS Valve Operation
- Removing power to the solenoid actuators actuates the main valves which operate using RCS pressure
- Spring ensures the ECCS valves stay open after depressurization
- Each ECCS main valve includes an inadvertent actuation block (IAB) device
- Operates solely on differential pressure between reactor vessel and containment
- The IAB is normally not engaged, only engages on high differential pressure
- The IAB does not actuate for any design basis ECCS demands
- Prevents inadvertent opening of an ECCS main valve until the differential pressure between the CNV and RCS reaches a predetermined range 27 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Emergency Core Cooling System (contd)
- Debris generation is limited by restricting use of insulation, paint, and coatings in containment
- Debris effects evaluated, see FSAR Section 6.3.3.1
- COL Item 6.3-1 requires COL applicant to implement cleanliness program to limit debris within the CNV
- Support debris evaluation assumptions
- ECCS capable of post-accident, extended long-term cooling beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (PDC 35 and GDCs 36 and 37) 28 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.3 - ECCS COL Item Item Number Description COL Item 6.3-1 Describe a containment cleanliness program that limits debris within containment. The program should contain the following elements:
- Foreign material exclusion controls to limit the introduction of foreign material and debris sources into containment.
- Maintenance activity controls, including temporary changes, that confirm the emergency core cooling system function is not reduced by changes to analytical inputs or assumptions or other activities that could introduce debris or potential debris sources into containment.
- Controls that limit the introduction of coating materials into containment.
- An inspection program to confirm containment vessel cleanliness prior to closing for normal power operation.
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6.4 - Control Room Habitability
- The control room habitability system (CRHS) is a passive system in standby during normal plant operation
- Provides breathable air and positive pressure to the control room if normal control room area ventilation is unavailable (FSAR 9.4.1)
- Supports 20 personnel for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without electrical power
- Air inventory monitored by pressure and temperature instrumentation
- Automatic valves and manual actuation line provided
- CRHS stored air and supply components are Seismic Category I
- Design includes connection to offsite air supply source if needed past 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 30 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.4 - Control Room Habitability
- Simplified CRHS diagram Control Room Envelope (CRE)
Air Bottles manual actuation S
S S
S automatic actuation balancing and pressure relief Compressor/Dryer M M M M Normal control room HVAC supply/return 31 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Control Room Habitability (contd)
- Classified nonsafety-related, however, augmented quality requirements are applied
- COL Item 6.4-5 requires applicant periodically test and inspect system
- CRHS automatic actuation upon
- Loss of AC to both normal control room HVAC system (CRVS) air handler units for 10 minutes, or
- High rad downstream of CRVS air filtration unit, or
- Loss of AC to all four EDSS-C battery chargers 32 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.4 - COL Items Item Number Description COL Item 6.4-1 Comply with Regulatory Guide 1.78 Revision 1, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."
COL Item 6.4-5 Specify testing and inspection requirements for the control room habitability system and control room envelope integrity testing as specified in Table 6.4-4.
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6.5 - Fission Product Removal and Control
- Reactor building and associated HVAC not credited in dose analysis
- No containment spray system 34 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.6 - Inservice Inspection and Testing
- ISI program for Class 2 & 3 systems/components based on 10 CFR 50.55a(g)(3)
- Preservice inspections and ISIs are performed on ASME BPVC Class 2 and 3 components in accordance with ASME BPVC Section XI
- Inspections performed per Class 1 requirements
- ISI Program includes augmented volumetric and surface inspections to protect against postulated piping failures
- High energy fluid system piping
- Break exclusion zone piping 35 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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6.6 - COL Items Item Number Description COL Item 6.6-1 Implement an inservice testing program in accordance with 10 CFR 50.55a(f).
COL Item 6.6-2 Develop preservice inspection and in-service inspection program plans in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), and establish the implementation milestones for the program. The COL applicant will identify the applicable edition of the ASME BPVC used in the program plan consistent with the requirements of 10 CFR 50.55a. The COL applicant will, if needed, address the use of a single in-service inspection program for multiple NuScale Power Modules, including any alternative to the code that may be necessary to implement such an in-service inspection program.
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Acronyms (page 1 of 3)
ANSI - American National Standards Institute ASME - American Society of Mechanical Engineers BDBE - beyond design basis event BPVC - Boiler and Pressure Vessel Code CIV - containment isolation valve CNV - containment vessel CNTS - containment system COL - combined license CRDS - control rod drive system CRE - control room envelope CRHS - control room habitability system CRVS - normal control room HVAC system DBE - design basis event DHR(S) - decay heat removal (system)
ECCS - emergency core cooling system ESFAS - engineered safety features actuation system 37 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Acronyms (page 2 of 3)
FSAR - Final Safety Analysis Report GDC - General Design Criteria HVAC - heating ventilation and air conditioning IAB - inadvertent actuation block ILRT - integrated leak rate testing ISI - inservice inspection ITAAC - Inspections, Tests, Analyses, and Acceptance Criteria LTOP - low temperature overpressure MPS - module protection system NPM - NuScale Power Module NPS - nominal pipe size PDC - principal design criteria PSCIV - primary system containment isolation valve P/T - pressure/temperature PZR - pressurizer RCPB - reactor coolant pressure boundary 38 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Acronyms (page 3 of 3)
RCS - reactor coolant system RG - Regulatory Guide RPV - reactor pressure vessel RRV - reactor recirculation valve RVV - reactor vent valve RXB - Reactor Building SGS - steam generator system SSCIV - secondary system containment isolation valve UHS - ultimate heat sink 39 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541.360.0500 971.371.1592 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541.360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1E 5BH Rockville, MD 20852 United Kingdom 301.770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www.nuscalepower.com Twitter: @NuScale_Power 40 PM-0619-65926 Revision: 0 Copyright 2019 by NuScale Power, LLC.
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Safety Evaluation with Open Items:
Chapter 6, Engineered Safety Features NuScale Design Certification Application ACRS Subcommittee Meeting June 18, 2019
Agenda
- NRC Staff Review Team
- Overview of the NRC Staffs Presentation
- Section 6.2.4
- Sections 6.2.5 & 6.2.6
- Section 6.3
- Section 6.4
- Abbreviations June 18, 2019 2
NRC Staff Review Team
- NRC Key Technical Reviewers
- Clint Ashley, NRO
- Ryan Nolan, NRO
- Shanlai Lu, NRO
- Michelle Hart, NRO
- Nan (Danny) Chien, NRO
- Boyce Travis, NRO
- Gregory Makar, NRO
- Syed Haider, NRO
- Tom Scarbrough, NRO
- Nick Hansing, NRO
- Carl Thurston, NRO
- Peter Lien, RES
- Project Management
- Omid Tabatabai, Senior Project Manager
- Greg Cranston, Lead Project Manager June 18, 2019 3
Overview of the NRC Staffs Presentation
- The focus of staffs presentation today is on SER Sections that involve Open Items and/or Exemption Requests, namely, Sections 6.2, Containment Systems, 6.3, Emergency Core Cooling System, and 6.4, Control Room Habitability.
- The NRC staffs will present its evaluation of FSAR Sections 6.2.1, Containment Structure, and 6.2.2, Containment Heat Removal, as part of Chapter 15 presentations June 19-20
- There are several Open Issues in Section 6.3 that do not have a path forward to resolution. These OIs are directly tied to open issues in Chapter 15.
- For Chapter 6, the staff issued 34 RAIs (114 Questions). We have received responses to 111 Questions.
June 18, 2019 4
Staffs Evaluation of Section 6.2.4, Containment Isolation System Clint Ashley Reactor Systems Engineer, NRO June 18, 2019 5
Regulatory Basis
- GDC 1 - Quality standards and records
- GDC 2 - Design bases for protection against natural phenomena
- GDC 4 - Environmental and dynamic effects design bases
- GDC 5 - Sharing of structures, systems and components
- GDC 16 - Containment design (essentially leak-tight barrier)
- 10 CFR 52.47(a)(8) - Technically relevant items in 10 CFR 50.34(f)
- 10 CFR 50.63 - Loss of all alternating current power
- 10 CFR 52.7 - Specific exemptions (refers to 10 CFR 50.12)
June 18, 2019 6
Review Guidance
- NuScale Design Specific Review Standard (DSRS)
- Section 6.2.4, Containment Isolation System.
- Regulatory Guide (RG)
- RG 1.141, Containment Isolation Provisions for Fluid Systems.
- RG 1.151, Station Blackout.
- U.S. Nuclear Regulatory Commission technical report designation (NUREG)
- NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident.
- NUREG-0737, Clarification of TMI Action Plan Requirements.
June 18, 2019 7
- NuScales DCA (Part 7) contains exemption requests associated with the following containment isolation requirements:
- GDC 55
- GDC 56
- GDC 57
June 18, 2019 8
10 CFR 50.12 - Specific Exemptions
- Pursuant to 10 CFR 50.12 in part, the Commission may grant exemptions when special circumstances are present.
- According to 10 CFR 50.12(a)(2)(ii), special circumstances are present whenever, application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.
June 18, 2019 9
Underlying Purpose
- The underlying purpose of GDCs 55, 56, and 57 is to provide containment isolation capability that supports the safety function of containment to provide a barrier to the uncontrolled release of radioactivity to the environment.
- The underlying purpose of 50.34(f)(2)(xiv)(E) is to limit radiological releases by ensuring containment isolation for systems that provide paths to the environs.
June 18, 2019 10
Staffs Review
- For GDC 55, 56 and 57, the staff finds that the NuScale design accomplishes the containment safety function by providing a containment isolation capability comparable to that required by the GDCs and therefore, the underlying purpose of the GDCs is met.
- The staff finds NuScales exemption request meets the requirements for an exemption as defined in 10 CFR 50.12.
June 18, 2019 11
Staffs Review (Contd)
- For 50.34(f)(2)(xiv)(E), the staff finds that the NuScale design isolates systems that provide a path to the environs before core damage or degradation occurs, preventing significant releases from the containment, and therefore, the underlying purpose of the rule is met.
- The staff finds NuScales exemption request meets the requirements for an exemption as defined in 10 CFR 50.12.
June 18, 2019 12
Staffs Evaluation of Section 6.2.5, Combustible Gas Control in Containment Anne-Marie Grady Reactor Systems Engineer, NRO June 18, 2019 13
Regulatory Basis
- GDC 41, 42, 43 - Containment atmosphere cleanup (and inspection and testing), requires control of hydrogen or oxygen that may be released into containment following postulated accidents to ensure containment integrity is maintained and supporting SSC safety functions can be accomplished
- 10 CFR 50.44(c)(2) - requires that the plant accommodate hydrogen generation up to 100 percent fuel clad-coolant reaction while limiting containment hydrogen to less than 10 percent and maintain containment structural integrity and other accident mitigation features June 18, 2019 14
Key Design Considerations and Features
- The NuScale containment is kept at a very low pressure during normal operation, limiting initial oxygen inventory
- The NuScale containment is designed to accommodate a bounding combustion event resulting from hydrogen generation at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without a loss of integrity or loss of supporting SSC functions
- NuScale has requested an exemption from 50.44(c)(2), which would require the design provide a hydrogen control system to limit hydrogen concentrations below 10 percent June 18, 2019 15
- Staff review focused on:
- Hydrogen conditions in the CNV during the postulated accident
- Demonstration of adequate containment mixing
- Equipment survivability in the event of the postulated hydrogen combustion
- NuScale analyses demonstrated the containment would survive a bounding combustion event inside the containment at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- Staff confirms the calculation for a limiting pressure pulse inside the containment resulting from a hydrogen combustion event at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for a very short duration, and containment integrity would be maintained.
- Staff recommends granting the Exemption Request
- Providing a system to control the hydrogen concentration is not necessary to serve the underlying purpose of 10 CFR 50.44(c)(2), which is to prevent a loss of containment integrity June 18, 2019 16
Open item
- To fulfill the requirement to provide hydrogen and oxygen monitoring that is functional and reliable, NuScale uses the process sampling system.
- The system is isolated on receipt of a containment isolation signal (would occur post-accident), but the applicant states that CNV isolation valves for CES and CFDS could be opened in the event of a need for sampling once containment pressure is low enough.
- Separate exemption request (post-accident sampling) raised questions related to this configuration.
- RAI 9682 has been issued to determine that this configuration could be safely established in the context of a beyond-DBA in order to meet the requirements associated with monitoring for 10 CFR 50.44(c)(4). This is Open Item 6.2.5-1.
June 18, 2019 17
Staffs Evaluation of Section 6.2.6, Containment Leakage Testing Anne-Marie Grady Reactor Systems Engineer, NRO June 18, 2019 18
Exemption Request Regarding Containment Leakage Testing
- NuScale requested an exemption from the following regulations:
- 10 CFR 50, App. A, GDC 52, Capability for Containment Leakage Rate Testing, requires the capability to perform containment periodic integrated leakage rate testing (ILRT) (Type A) at containment design pressure.
- 10 CFR 50, App. JPrimary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, requires Type A tests preoperationally and periodically. thereafter.
- NuScale asserts that the CNV ASME design, analysis for leak tightness, 100% vessel inspectability, and pre service design pressure hydrostatic leakage test would satisfy the underlying goal of demonstrating CNV allowable leakage.
June 18, 2019 19
Exemption Request Regarding Containment Leakage Testing (Contd)
- NuScale CNV is an ASME Section III, Class 1 pressure vessel
- CNV leakage analysis, design specifications, capability for 100% vessel inspection, examination, and testing, provide assurance that the leakage integrity of the CNV is maintained.
- NuScale analyzed the CNV bolt design for the flanged openings, using ANSYS, based on the seal design and specification.
- NuScale calculated flange contact pressures and corresponding flange gaps, based on CNV internal accident pressure and temperatures.
- Staff reviewed these calculations during an audit.
June 18, 2019 20
Exemption Request Regarding Containment Leakage Testing (Contd)
- NuScale asserts that the CNV design combined with the Types B and C test results are sufficiently representative of accident conditions to demonstrate that the TS leak rate, La, would not be exceeded.
- In addition to testing required by ASME, NuScale proposes a preservice design pressure test to confirm the expected performance of the CNV design. This would be verified by ITAAC.
- The staff recommends approving the exemption request to not require Type A testing nor to require design capability for ILRT as required by GDC 52.
June 18, 2019 21
Staffs Evaluation of Section 6.3, Emergency Core Cooling System Shanlai Lu Senior Reactor Systems Engineer, NRO June 18, 2019 22
NuScale ECCS Design Features
- 3 Reactor Vent Valves (RVV)
- 2 Reactor Recirculation Valves (RRV)
- Each ECCS valve has its own IAB, trip valve, and trip reset valve
- Containment functions as part of ECCS June 18, 2019 23
Regulatory Basis
- GDC 2 - Design bases for protection against natural phenomena
- GDC 4 - Environmental and dynamic effects design bases
- GDC 5 - Sharing of structures, systems and components
- GDC 17 - Electrical power system requirements
- GDC 27 - Neutron poison addition and appropriate shutdown margin for stuck rods
- GDC 36 - Permit appropriate periodic inspection of important components
- GDC 37 - Permit appropriate periodic pressure and functional testing
- 10 CFR 50.46(b)(5) - Long-term cooling
- 10 CFR 52.47(b)(1) - Inspection, Tests and Analyses and Acceptance Criteria
- 10 CFR 52.80(a) - COL application contain the proposed inspection, tests and analyses
- 10 CFR 50.34(f) - Reactor coolant system safety and relief valves June 18, 2019 24
Review Guidance
- NuScale Design Specific Review Standard (DSRS)
- Section 6.3, Emergency Core Cooling System
- Regulatory Guide (RG)
- NUREG-0737, Clarification of TMI Action Plan Requirements June 18, 2019 25
- 1. NuScale requested partial exemptions from certain requirements in 10 CFR 50.46 and 10 CFR Part 50, Appendix K, specifically, from the following requirements:
- a. Swelling and rupture of the cladding and fuel rod thermal parameters
- b. Pump modeling
- c. Refill and reflood heat transfer
- d. BWR specific items
- 2. GDC 27 requires neutron poison addition via the ECCS. Current NuScale ECCS design does not have the capability to inject additional neutron poison.
- The staffs evaluation of exemption requests are ongoing. Detailed discussions will be provided during Chapter 15 ACRS presentations.
June 18, 2019 26
Technical Evaluation Sub-section Title Issue Description Resolution Status Reactor coolant boundary ECCS performance testing needed Open Item 3.9.6-1 Low Temperature Over ECCS provides protection up to Closed. Section 5.2 Pressure Protection 750 psig (LTOP)
Core Cooling - Contingent on LOCA TR and Open Item 6.3-1 Inadvertent ECCS 15.6.6 review Actuation Core Cooling - Secondary Contingent on Non-LOCA TR and Open Item 15.0.2-4 System Pipe Break Inside 15.2.8 review Containment Core Cooling - Loss of LOCA TR and ECCS Valve Flange Open Item 3.6.2-1 Coolant Accident Design Open Item 15.0.2-2 Core Cooling - Long Term Return to power and boron Open Item 15.0.6-5 Cooling transport. Section 15.0.6 Shared Systems Only the reactor pool is shared Closed June 18, 2019 27
Technical Evaluation (Contd)
Sub-section Title Issue Description Resolution Status Power Requirements Use of non-safety graded DC system Open Item 15.0.0.5-1 Instrumentation Valve position and solenoid power Closed. Chapter 7.
indication in the control room System Boundary No direct fluid mass exchange with the Closed reactor pool Testing, Inspection and Demonstration tests are on going. Open Item 3.9.6-1 Qualification Environmental Requirements EQ, ASME compliance, Seismic Closed (Section 7.2.2, qualification and missile protection 3.2, 3.3, 3.6, 3.10 and 9.5)
System Reliability Failure mode and effects analysis Open Item 3.9.6-1 Single Failure IAB single failure assumption Open Item 15.0.0.5-1 Technical Specifications ECCS Specific Items Closed. Chapter 16 June 18, 2019 28
Technical Evaluation (Contd)
In-vessel Debris Downstream Effects Evaluation Debris Generation:
Only Reflective Metallic Insulation is used in the containment and there is no chemical buffer. Only latent debris is considered to be present Debris Transport:
100% of latent debris is assumed to arrive at the core inlet Fuel Bundle Head Loss:
Only 5.6 g/assembly fiber, 33.4 g/assembly particulate are estimated to arrive at the inlet of the core. Without a pH buffer, conservative amount of chemical precipitates is assumed.
Based on relevant AREVA fuel bundle head loss testing and LOCA deposition model, it is concluded that there would be no adverse impact on long term cooling due to limited amount of latent debris June 18, 2019 29
Water Hammer Evaluation
- During the ECCS actuation, the ECCS actuator lines and trip set valves experience two-phase choke flow and sudden depressurization.
- Staff issued RAI 9469 (Question 31517, ADAMS Accession No. ML18162A351) to better understand the capability of ECCS actuator hydraulic line design, analysis, and tests, and to evaluate the ability of ECCS to withstand potential water hammer during actuation.
- Additional tests are planned by NuScale to assess the likelihood of water hammer phenomenon. The test facility includes a long actuator hydraulic line with several 90 degree bends and pressure sensors.
- This part of the staffs evaluation is being tracked and documented in Chapter 3 (Open Item 03.09.06-1).
June 18, 2019 30
Conclusions
- Seven Open Items have been identified by the staff
- The NRC staff and NuScale are actively engaged to develop a path toward resolution for the following three Open Items. More discussions during Chapter 15 presentation.
- Open Item 03.09.06-1: ECCS demonstration testing
- Open Item 15.0.0.5-1: SECY paper regarding IAB single failure assumption
- Open Item 15.0.6-5: Boron dilution during long term cooling June 18, 2019 31
Staffs Evaluation of Section 6.4, Control Room Habitability Michelle Hart Senior Reactor Engineer, NRO June 18, 2019 32
Control Room Radiological Habitability
- GDC 19 dose criterion
- 5 rem TEDE for duration of accident
- Open Item 6.4-1:
- NuScale requested exemption from control room design criteria in GDC 19 to instead have NuScale principal design criterion (PDC) 19
- Effectively no change to dose requirement
- Review is not complete June 18, 2019 33
Control Room Radiological Habitability (Contd)
- Includes direct dose contribution from core damage event
- Open Item 6.4-3:
- Revisions to dose analyses in DCA FSAR Ch. 15 and referenced accident source term methodology topical report received in late April 2019
- Review is not complete June 18, 2019 34
Control Room Radiological Habitability (Contd)
- Post-accident control room habitability is not a safety related function
- No operator actions are required or credited to mitigate the consequences of design basis events
- No post-accident long-term monitoring from the control room is necessary
- Open Item 6.4-2:
- Staff to evaluate questions on potential operator actions from control room to achieve post-accident monitoring of containment atmosphere hydrogen and oxygen concentration June 18, 2019 35
Control Room Radiological Habitability (Contd)
- Credit in dose analyses for control room ventilation and habitability systems that are not engineered safety features:
- CRVS and CRHS backup systems for each other
- Both designed to be reliable and capable of operation during accident conditions
- Assumed accident duration is 30 days
- Independent and diverse systems
- Both have automatic initiation with different signals for each system June 18, 2019 36
Control Room Radiological Habitability (Contd)
- CRHS operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then CRVS in supplemental filtration mode for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> through 30 days
- CRVS operation in supplemental filtration mode for entire 30 days
- Both cases meet dose criteria for all accidents analyzed June 18, 2019 37
Control Room Radiological Habitability (Contd)
- Resolution of staffs concerns about apparent reliance on both CRHS and CRVS over the duration of the accident to meet control room requirements
- NuScale sensitivity analyses met dose criterion
- CRHS operates for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> only, then CRVS fails
- Staff confirmatory assessment of sensitivity analyses shows similar results
- Staff confirmed NuScale design does not need to rely on CRVS operation after CRHS is exhausted June 18, 2019 38
Review Status/Conclusion
- Review not complete
- Staff cannot yet reach a conclusion whether the control room radiological habitability requirements are met until resolution of Open Items 6.4-1 through 6.4-3 June 18, 2019 39
Abbreviations ACRS Advisory Committee on Reactor Safeguards ASME American Society of Mechanical Engineers CES Containment Evacuation System CFDS Containment Flooding and Drain System CFR Code of Federal Regulations CNV Containment Vessel CRHS Control Room Habitability System CRVS Control Room Ventilation System COL Combined License DBA Design-Basis Accident DCA Design Certification Application DSRS Design Specific Review Standard FSAR Final Safety Analysis Report GDC General Design Criteria GL Generic Letter ILRT Integrated Leakage Rate Testing ITAAC Inspections, Tests, Analyses, and Acceptance Criteria MSS Main Steam System NRC Nuclear Regulatory Commission NRO NRC Office of New Reactors PDC Principal Design Criteria RG Regulatory Guide SER Safety Evaluation Report SSC Structures, Systems, and Components TEDE Total Effective Dose Equivalent June 18, 2019 40