ML18360A524
| ML18360A524 | |
| Person / Time | |
|---|---|
| Issue date: | 12/06/2018 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-4025 | |
| Download: ML18360A524 (200) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards Docket Number:
(n/a)
Location:
Rockville, Maryland Date:
Thursday, December 6, 2018 Work Order No.:
NRC-4025 Pages 1-200 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 659TH MEETING 4
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5
(ACRS) 6
+ + + + +
7 THURSDAY 8
DECEMBER 6, 2018 9
+ + + + +
10 ROCKVILLE, MARYLAND 11
+ + + + +
12 The Advisory Committee met at the Nuclear 13 Regulatory Commission, Three White Flint North, Room 14 1C3 & 1C5, 11601 Landsdown Street, at 1:00 p.m.,
15 Michael L. Corradini, Chairman, presiding.
16 17 COMMITTEE MEMBERS:
18 MICHAEL L. CORRADINI, Chairman 19 PETER RICCARDELLA, Vice Chairman 20 RONALD G. BALLINGER, Member 21 DENNIS C. BLEY, Member 22 CHARLES H. BROWN, JR. Member 23 MARGARET SZE-TAI Y. CHU, Member 24 VESNA B. DIMITRIJEVIC, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2 WALTER L. KIRCHNER, Member 1
JOSE MARCH-LEUBA, Member 2
HAROLD B. RAY, Member 3
JOY L. REMPE, Member 4
GORDON R. SKILLMAN, Member 5
MATTHEW W. SUNSERI, Member 6
7 ACRS CONSULTANT:
8 STEPHEN SCHULTZ 9
10 DESIGNATED FEDERAL OFFICIALS:
11 QUYNH NGUYEN 12 KENT HOWARD 13 14 ALSO PRESENT:
15 KENNETH BROWNE, NextEra 16 WILLIAM BURTON, NRR 17 ANDY CAMPBELL, NRO 18 EDWARD CARLEY, NextEra 19 MICHAEL COLLINS, NextEra 20 JOSEPH DONOGHUE, NRR 21 ALLEN FETTER, NRO 22 RUDY GIL, NextEra 23 MICHELLE HART, NRO 24 ALLEN HISER, NRR 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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3 ARCHIE MANOHARAN, Tennessee Valley Authority 1
ERIC McCARTNEY, NextEra 2
BRUCE MUSICO, NSIR 3
ERIC OESTERLE, NRR 4
RAYMOND SCHIELE, Tennessee Valley Authority 5
DANIEL STOUT, Tennessee Valley Authority 7
ALEX YOUNG, Tennessee Valley Authority 9
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 C-O-N-T-E-N-T-S 1
Clinch River Early Site Permit 5
2 Seabrook License Renewal Application
...... 97 3
Adjourn....................
123 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5 P R O C E E D I N G S 1
(1:00 p.m.)
2 CHAIRMAN CORRADINI: Okay. The meeting 3
will come to order. This is the first day of the 4
659th meeting of the Advisory Committee on Reactor 5
Safeguards.
6 During today's meeting the Committee will 7
consider the following. Clinch River early site 8
permit, Seabrook License Renewal Application, and then 9
preparation of ACRS reports.
10 The ACRS was established by statute, and 11 is governed by the Federal Advisory Committee Act, or 12 FACA. As such, this meeting is being conducted in 13 accordance with the provisions of FACA. That means 14 that the Committee can only speak through its 15 published letter reports.
16 We hold meetings to gather information to 17 support our deliberations. Interested parties who 18 wish to provide comments can contact our offices 19 requesting time after the Federal Register notice 20 describing the meeting as published.
21 That said, we also set aside ten minutes 22 for extemporaneous comments from members of the public 23 attending or listening to our meetings. Written 24 comments are also welcome.
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6 Mr. Quynh Nguyen is the designated Federal 1
Official for the initial portion of the Meeting. The 2
ACRS section of the US NRC public website provides our 3
charter, by-laws, letter reports, and full transcripts 4
of all our full and Subcommittee meetings, including 5
all the slides presented at those meetings.
6 At this time we've not received any 7
written comments, or requests to make oral statements 8
from members of the public regarding today's session.
9 There will be phone bridge line. To preclude 10 interruption of the meeting the phone will placed in 11 a listen in only mode during the presentation of the 12 Committee discussion.
13 Also, a transcript of portions of the 14 meeting is being kept, and it is requested that 15 speakers use one of the microphones, identify 16 themselves, and speak with sufficient clarity and 17 volume so they can be readily heard.
18 So, at this time I'll just remind 19 everybody, take all your things and turn them off, or 20 put them in mute, so we don't have to hear buzzing or 21 beeping. And with that I'll turn to Member Kirchner 22 to lead us through the first topic.
23 MEMBER KIRCHNER: Thank you, Chairman.
24 Apologies for the slight delay in arriving. We have 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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7 heard from the applicant and staff over the course of 1
the last year.
2 We had a informational briefing on 3
November 15th of last year. Then we had four 4
additional informative meetings with both parties.
5 So, with that I'm ready to turn it over to the staff, 6
to Andy Campbell to proceed, please.
7 MR. CAMPBELL: If I can remember how to 8
turn these things on. I'm Andy Campbell. I'm the 9
Deputy Director of the Division of Licensing, Siting, 10 and Environmental Analysis in the Office of New 11 Reactors at the NRC.
12 Mr. Chairman, it is a great pleasure to be 13 here today for the full Committee meeting on the 14 Clinch River Nuclear site, early site permit, what 15 we'll call the SP, application safety review submitted 16 to the NRC May 26, 2016.
17 This submittal is the first ESP for a 18 small modular reactor plant design. And it was prior 19 to staff's work on the small modular reactor and other 20 new technologies rulemaking. Accordingly, the 21 application and the review of the application by the 22 staff is based on current regulations and guidance.
23 Staff has presented a series of ACRS 24 Subcommittee meetings on the staff's safety review of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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8 the application. And today staff will be presenting 1
our final overview, with no open items, for the Clinch 2
River ESP safety evaluation report.
3 The ESP review has been progressing 4
consistent with the schedule, and completion of 5
today's full Committee now puts the project ahead of 6
schedule.
7 For example, staff provided an overview to 8
ACRS in November 15, 2017, a little over a year ago.
9 Previous staff presentations for the relevant SER 10 chapters to several ACR Subcommittee meetings, from 11 May 15 of this year, 2018, to November 14, 2018.
12 The NRC staff safety review of the 13 application included the execution and completion of 14 five audits and one inspection, and the issuances of 15 12 RAIs comprising 50 questions.
16 The staff completed all the advance safety 17 evaluation with no open items.
18 Staff's presentation, and then the 19 applicant's presentations today are, we're going to 20 focus on, the staff will focus on the EPZ, with an 21 overview of the other Subcommittee presentations.
22 One key point is, if the exemptions are 23 approved for the ESP, the COL applicant can adopt 24 these exemptions if it shows that a COLA PEPE EPZ 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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9 source term release to the atmosphere are bounded by 1
the non-design specific plant parameter source term 2
information developed for the ESP.
3 A future COL application featuring an SMR 4
design that fits within the plant parameter envelope 5
established in the ESP could apply the approved 6
methodology to the design selected, to determine the 7
appropriate PEP EPZ, and for the site, and also to 8
demonstrate whether the conditions for either of the 9
two sets of exemptions have been met.
10 Also in the audience today, besides NRC 11 staff and applicant staff are representatives from the 12 Federal Emergency Management
- Agency, FEMA, 13 Technological Hazards Division. And representatives 14 from Tennessee Emergency Management Agency are on the 15 conference bridge. So, now I'm going to turn it back 16 to you.
17 MEMBER KIRCHNER: So, thank you. I think 18 we're going to turn to the applicant at this point.
19 Okay. Dan, please proceed.
20 MR. STOUT: Thank you. Good afternoon.
21 I want to start by expressing our appreciation for the 22 flexibility to adjust the schedule, and get this done.
23 We took advantage of the opportunity and got to go pay 24 our respects at the Capital yesterday morning early.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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10 And so, win, win.
1 So, I'm Dan Stout. I'm the Director of 2
Nuclear Technology and Innovation for the Tennessee 3
Valley Authority, managing this small module reactor 4
activity, particularly the early site permanent 5
application.
6 I'll be kicking off an introduction, 7
talking about the site and the SMR program. And then 8
I'm going to turn it over to Ray Schiele, Licensing 9
Manager, who's going to cover the specifics of the 10 early site permanent application itself. And then, as 11 requested, Archie Manoharan will be doing a deeper 12 dive into the emergency preparedness portion of the 13 application.
14 So, I'd like to acknowledge the Department 15 of Energy, who has been an integral partner in 16 supporting the SMR activities that TVA is undertaking, 17 particularly with financial assistance. However, the 18 views expressed are TVA's alone.
19 So, on Slide 5, I'll remind everyone that 20 Tennessee Valley Authority's mission is broader than 21 just making electricity. It's also important to be a 22 good steward of the environmental resources, and to be 23 a partner in economic development.
24 TVA has been focused on the Clinch River 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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11 site in Oak Ridge, Tennessee. It's a 1,200 acre site.
1 The project is confined to 335 acres on that 1,200 2
acre reservation.
3 And it is a good site. Has access to both 4
500 and 161 KV transmission, which cut through the 5
site. It is a neighbor to the Department of Energy, 6
a customer that is interested in the output from this 7
project.
8 The site was disturbed back in the 1970s 9
and '80s. It was the site of the former Clinch River 10 breeder reactors.
So, there's some basic 11 infrastructure, roads, storm water retention, things 12 like that.
13 The community of Oak Ridge, you couldn't 14 ask for a better place to want to do something 15 nuclear. Not only is there strong community support, 16 but there's an abundant and skilled nuclear workforce 17 there. And it's a site that's within TVA's ownership 18 and control. So, it makes proceeding rather easy.
19 Next.
So, the early site permit 20 application itself consists of site safety analysis 21 report, environmental report, Part 5 emergency plans.
22 And we actually submitted two different emergency 23 plans, one for site boundary, one for two mile.
24 Archie will get into those details. And a consistent 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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12 set of exemptions that go along with those emergency 1
plans.
2 Our early site permit application is based 3
upon a plant parameter envelope that was informed by 4
the designs of the four U.S. light water reactors that 5
were under development over the previous few years.
6 That includes the B&W mPower, Holtec, NuScale, and 7
8 The application was developed, and the 9
plant parameter envelope was developed based upon NE 10 1001 guidance. It assume that two or more reactors of 11 the same design deployed, and a maximum of 800 12 megawatts thermal for an individual reactor, and a 13 maximum of 2,420 megawatts thermal for the site.
14 Next. So, the schedule, we're here 15 focused on the safety element, which is the, kind of 16 the top row. There's the other track, environmental, 17 and then the hearings.
18 So, on the safety side the NRC schedule 19 calls for issuance of the final safety evaluation 20 report in August. We're hopeful that we're ahead of 21 that schedule.
22 The environmental, the staff issued the 23 draft environmental impact statement in April. And it 24 looks like we're on track to be ahead of the June 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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13 schedule goal.
1 On the hearing side there were four 2
contentions filed. Two were admitted. In July the 3
Atomic Safety and Licensing Board dismissed all 4
outstanding contentions, and terminated the contested 5
hearing. Subsequently, the Commission indicated that 6
it's their intent to run the mandatory hearing.
7 Next. So, I'd like to hit some highlights 8
of the early site permit application, and the review 9
process itself. The NRC commenced the review in the 10 very beginning of 2017. The application as originally 11 submitted had about 8,000 pages, supported by about 12 80,000 pages of technical information.
13 One of the highlights I'd like to point 14 out is the efficient use of audits. The staff did a 15 great job of preparing well in advance, and listing 16 out all of their questions, all of the information 17 needs, well in advance of the audit.
18 So then, when the audit occurred we were 19 able to prepare responses to all of those open items 20 well, all of those information needs well in advance, 21 so that when they were there face to face there was 22 meaningful discussion on the challenges.
23 By the end of the audits we had clarity on 24 how to resolve all the issues. That manifested itself 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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14 in very few RAIs. As Andy mentioned, it's about a 1
dozen, as compared to hundreds for prior applications.
2 And I'm going to attribute a lot of that 3
success to very
- frequent, clear, and candid 4
communication. We, both staff and the applicant, we 5
identified issues early, and we escalated them, put 6
the resources on those issues early.
7 So, next. So, I'd like to turn it over to 8
Ray Schiele now to talk about the early site permit 9
application.
10 MR. SCHIELE: Thank you, Dan. Good 11 afternoon. I'm Ray Schiele, currently the Licensing 12 Manager for the Clinch River Nuclear Early Site Permit 13 Application. I have 44 years in this industry, 14 primarily operations and licensing. And since 2016 15 the Licensing Manager for the Clinch River project.
16 Quick overview of the organization of the 17 application. The Clinch River application contains 18 the information required by 10 C.F.R. 52.17, contents 19 of applications for an early site permit. And was 20 submitted in accordance with NRC guidance on 21 electronic submittals.
22 Part 1, administrative information. This 23 section contains an overview of the early site permit 24 application, a general description of the format, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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15 content of the application, and corporate information, 1
including ownership, management, and Board of 2
Directors.
3 Part 2, the SSAR, includes a discussion of 4
the site description, safety assessment, quality 5
assurance, general location of the site, site 6
suitability, design parameters postulated for the CRN 7
site, population profiles, and an assessment of site 8
features that may affect the design chosen for the 9
facility.
10 Part 3, environmental report. The ER 11 addresses the environmental impacts associated with 12 construction and operation of new SMRs.
13 Part 4, site redress plan. TVA is not 14 pursuing a limited work authorization with this 15 application. Therefore, there is no redress plan.
16 Part 5, emergency planning information.
17 The emergency planning information includes major 18 features of the emergency plan. And there will be 19 more information with Archie.
20 Part 6, exemptions and departures. This 21 part lists applicant requested exemptions that are 22 authorized by law, would not endanger life, property, 23 or common defense and security, and are otherwise in 24 the public interest. A discussion and justification 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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16 for each of the requests is included in this part.
1 There were no departures requested in Part 6.
2 Part 7, withheld information. This part 3
contains information redacted from other parts of the 4
application due to sensitive or proprietary nature of 5
the information.
6 And last, Part 8, enclosures. All 7
enclosures submitted with the early site permit 8
application are provided in Part 8.
9 ESPA development, the regulatory bases.
10 This slide illustrates the regulatory bases for the 11 development of both the SSAR and ER. The regulatory 12 bases consist of various regulations, standard review 13 plans, reg guides, and review standards.
14 NRC interactions. Prior to the ESPA 15 submittal in May of 2016 the NRC performed pre-16 application site visits, alternative site visits, and 17 pre-application readiness review.
18 After submittal the NRC performed three 19 major audits in the spring and summer of 2017, 20 supporting hydrology, ground water, seismic, geotech, 21 and environmental.
22 In addition, a comprehensive four month EP 23 audit not listed on this slide commenced in the fall 24 of 2017, and was supplemented by an additional audit 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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17 in the spring of 2018. In the spring of 2018 the NRC 1
conducted a QA inspection, covering Chapter 17.5 of 2
the SSAR.
3 Community timeline. In 2018 the ACRS 4
Committee met in May, August, October, and November to 5
review selected SSAR sections, as shown on the slide.
6 And today, as the slide illustrates, we're here for 7
the final full Committee meeting.
8 TVA was asked to provide additional 9
information associated with the approach to emergency 10 preparedness. I would now like to introduce Archie to 11 discuss the EP. Archie.
12 MS. MANOHARAN: Thank you, Ray. Good 13 afternoon. Thank you for the opportunity to present 14 today. As we mentioned I'm Archie Manoharan. I've 15 been working in the nuclear industry for the last ten 16 years, joined the licensing team at Clinch River in 17 2017.
18 And I would like to begin with the layout 19 of the emergency preparedness approach in the 20 application. To fully understand the emergency 21 preparedness approach for Clinch River it's important 22 to consider the information in three parts of the 23 application.
24 Part 2, SSAR Section 13.3 in Section, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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18 emergency preparedness, describes a dose based 1
consequence or entered methodology for determining a 2
plume exposure pathway EPZ for the site.
3 We have not selected a reactor design for 4
the site. So, in this section the application is only 5
seeking approval to use the methodology at a later 6
stage, with design specific information, say in a 7
COLA.
8 This methodology, along with the SMR 9
design features is sort of the basis for the emergency 10 preparedness approach described in the application.
11 Based on the methodology Part 5 of the application has 12 two distinct emergency plans.
13 Part 5 Alpha has major features of an 14 emergency plan for a site boundary EPZ. And Part 5 15 Bravo has major features of an emergency plan for a 16 two mile EPZ. Again, only major features are 17 discussed in Part 5. There is no design specific 18 information.
19 At a COLA, once the reactor design has 20 been selected, and the dose based methodology that's 21 described in 13.3 is adequately demonstrated, we would 22 pick one of the emergency plans described in Part 5.
23 For example, if the selected reactor 24 design meets the dose criteria at site boundary, we 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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19 would go ahead and use Part 5 Alpha to create a 1
integrated and complete emergency plan and COLA. If 2
the reactor design meets the dose criteria at two mile 3
EPZ, then Part 5 Bravo would be used.
4 The information in Part 5 meets the 5
regulatory requirements if you consider it with the 6
exemption requests described in Part 6. In Part 6 of 7
the application two sets of exemption requests have 8
been described, one to support the site boundary EPZ, 9
and the other for the two mile.
10 Next slide. We're on Slide 17. And the 11 dose based methodology described in Section 13.3 is 12 consistent with the sizing rationale described in 13 NUREG 0396. The NUREG introduced the concept of a 14 generic EPZ, and recommends that a spectrum of 15 accidents be addressed for the EPZ sizing.
16 So, consistent with that approach the 17 methodology we are proposing in the application also 18 describes, also addresses a spectrum of accidents.
19 And more importantly, it has the same dose criteria 20 for the plume exposure pathway EPZ as a recommendation 21 in NUREG 0396, which is the one rem total effective 22 dose equivalent, the early phase EPA PAG.
23 Consistent with the NUREG the technical 24 criteria in the dose based methodology can be 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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20 understood as Criteria Alpha, Bravo, and Charlie.
1 Alpha can be understood as the plume exposure EPZ 2
should be of, encompass of those areas where projected 3
dose from design basis accidents could exceed the one 4
5 Bravo the
- same, except for dose 6
consequences from less sever core melt accidents.
7 Criterion Charlie would verify that the plume exposure 8
pathway EPZ is of sufficient size to provide for 9
substantial reduction in early health effects in the 10 case of more severe core melt accidents.
11 Next slide. So, we're on Slide 18. And 12 this slide here describes the steps involved in 13 implementing the methodology. The methodology at a 14 high level contains four steps, starting with accident 15 scenario selection. This is where you would rely on 16 design and site specific information to do the 17 appropriate accident selection.
18 For Criterion Alpha accidents you would 19 rely on the bounding design basis accidents from 20 Chapter 15 of the COLA. For the severe accident 21 scenarios you would rely on the site and design 22 specific PRA. And the criteria is actually shown 23 here.
24 So, sequence. Firstly we'll start with 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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21 sequences with a mean core damage frequency greater 1
than E to the negative eight per reactor year. And 2
then you would further categorize them into criteria.
3 Bravo accident scenarios would be mean 4
core damage frequency greater than E to the negative 5
six, with intact containment.
6 And Charlie, the more severe core melt 7
accidents, would be accidents with mean core damage 8
frequency greater than E to the negative seven, or 9
with containment bypass of the --
10 MEMBER KIRCHNER: Archie, may I interrupt 11 here? So, I think this is mentioned in your 12 application. For rhetorical purposes, if the design 13 you choose, the PRA doesn't show any accidents 14 greater, severe accidents. I'm looking at in 15 particular greater than one E to the minus seven.
16 Then I think you suggest putting in an alternate 17 source term. Is that --
18 MS. MANOHARAN: That is correct. So, for 19 Criterion Bravo, it's not listed on this slide, but 20 there is an additional note in the methodology that 21 even if you pick a reactor design that has no accident 22 screened in for Criterion Bravo you still have to 23 create alternate --
24 MEMBER KIRCHNER: Now, is -- Well, I'll 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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22 get a chance to ask the staff whether they're in 1
agreement with this approach. But then, how would you 2
come about, go about picking that source term?
3 MS. MANOHARAN: I think we can actually 4
explain that during the example analysis.
5 MEMBER KIRCHNER: Okay. I'll wait.
6 MS. MANOHARAN: Which is in the next 7
slide. Because we encountered that exact scenario in 8
the example analysis. So, okay. So, moving on to the 9
next slide, 19.
10 So, you would, after the steps one through 11
-- I apologize. Can we go back to 18? Yes. So, 12 after the accident selection, based on the cut off 13 frequencies described here, Step 2 would be to 14 determine the source term releases from the selected 15 accidents.
16 Step 3 would be to calculate the dose 17 resulting from these accidents at a distance from the 18 plant. Four obviously would be to compare that to the 19 EPA PAG limits to ensure that we are within that one 20 rem limit. Next slide, please.
21 So, Criteria Alpha and Bravo, as I just 22 mentioned, you would compare the dose calculated to 23 one rem, and make sure you're not exceeding that. For 24 Criterion
- Charlie, consistent with NUREG 0396 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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23 approach, you would calculate the distance at which 1
the conditional probability to exceed 200 rem whole 2
body exceeds in the negative three per reactor year.
3 CHAIRMAN CORRADINI: So, with that one, 4
can you tell me how the one in a thousand is computed?
5 I go back to 0396, and I'm lost. Tell me how that's 6
computed. I understand the dose criteria. I don't 7
understand what the frequency represents.
8 MS. MANOHARAN: Okay. I will bring in 9
Alex to --
10 CHAIRMAN CORRADINI: If you want to do it 11 later, that's fine. I, whenever it's suitable. I 12 just want to understand what that is.
13 MS. MANOHARAN: We can do it now.
14 CHAIRMAN CORRADINI: Okay.
15 MR. YOUNG: So, my name's Alex Young. I'm 16 working as a design engineer on the SMR project. Been 17 here since September of 2014. So, the question is, 18 you know, about the Criterion C dose criteria.
19 The conditional probably to exceed 200 rem 20 whole body is one E minus three. So, we look at that.
21 As you go out in distance from the release point, the 22 reactor building, the probability of acquiring a 23 certain dose goes down, based on meteorology.
24 So, we're looking at the distance at which 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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24 the probability of acquiring the 200 rem whole body 1
dose exceeds the one E minus 3.
2 CHAIRMAN CORRADINI: I got that part. I 3
don't understand why -- So, let me, so, here's where 4
I'm confused. I've now got accidents that fit in a 5
range of greater than ten to the minus seven, but less 6
than ten to the minus six. Yes, the frequency is one 7
ten minus three.
8 So, have you subtracted a way, or taken 9
out the initiating even frequency? This, the number 10 sounds high to me, one in a thousand. I'm confused 11 about one in ten to the seventh, ten to the minus 12 seventh, versus ten to the minus three. That's where 13 I'm struggling.
14 MR. YOUNG: Sure. So, for the Criterion 15 C piece, on the previous line we kind of highlight the 16 main CDF greater than one E minus seven per reactor 17 year. So, that's looking at the probability of the 18 event.
19 So, once you have the event, and you have 20 a release, primarily based on meteorology statistics 21 you have the probability changing as you go out in 22 distance for that release. So, it's an additional 23 factor in addition to the screening piece that's added 24 in Criterion C.
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25 MEMBER MARCH-LEUBA: So, in terms I can 1
understand. Sometime we talk about a 500 year flood, 2
100 year flood. This is equivalent to that? So, you 3
have the same, the initial source term. And now you 4
consider the one thousand worst year that can possibly 5
happen? Correct?
6 MR. YOUNG: Yes. That's a good analogy to 7
categorize it.
8 CHAIRMAN CORRADINI: Okay. I'm still not 9
there. Sorry. So, I've taken away the initiating 10 event frequency, and all the estimates. And I've 11 developed the source term. Then I release the source 12 term, and I ask, what's the probability of getting a 13 dose greater than 200 rem at a distance?
14 MR. YOUNG: You find out what distance it 15 is at which the probability of getting that dose is 16 one E minus --
17 MEMBER MARCH-LEUBA: But you run a 18 thousand different years and pick the worst.
19 Basically that's what you do, right? So, you start 20 with a source term. And then, you propagate it, year 21 one, year two, year three, using different winds, 22 rain, different meteorological conditions, and pick 23 the worst in a thousand.
24 MR.
YOUNG:
So, that's where the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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26 meteorology comes into play, is looking at, you know, 1
the meteorology that we have over time, how the 2
statistics play out in that. What are the 3
probabilities of having certain meteorological 4
conditions that, you know, make it, you know, how that 5
disburses.
6 CHAIRMAN CORRADINI: Okay. But if I might 7
just jump in? So, the one in a thousand is due to the 8
meteorology at the site? It's not due to the 9
production of the source term?
10 MR. YOUNG: It's both. It's the 11 combination. Because you have the initial even, which 12 allows the probability of the release.
13 CHAIRMAN CORRADINI: No. That part I got.
14 But once I get the source term, because it sits in 15 this band between ten minus seven and ten to the minus 16 six, now I have a source term. And the one in a 17 thousand is just a meteorological uncertainty, or 18 meteorological distribution?
19 MR. YOUNG: That's the additional factor 20 that is applied to the propagation of the source term.
21 CHAIRMAN CORRADINI: Okay.
22 MEMBER REMPE: So, if I went to the next 23 slide here, and I looked at that number. You call it 24 a probability. But it's got a frequency unit.
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27 MR. YOUNG: Yes.
1 MEMBER REMPE: So, wouldn't it be a 2
probability? Doesn't have a unit -- Why does it have 3
units of frequency if, I mean, earlier you called it 4
a core damage frequency, something per reactor year.
5 Now you're calling this a conditional 6
probability. Shouldn't it just be ten to the minus 7
three, instead of per reactor year? This is kind of 8
a basic question here. But I thought probabilities 9
wouldn't be in per reactor year.
10 MR. YOUNG: Sure. So, we think, and a lot 11 of times we think of, you know, probability. And we 12 tie that to a frequency here. So, we're looking at, 13 you know, the probability that you have that 200 rem 14 dose at what distance for one E minus three per 15 reactor year.
16 MEMBER SKILLMAN: Alex, I'd like to ask 17 this. At least two times, and maybe three, you 18 mentioned the coupling of the probability of the event 19 with meteorology.
20 MR. YOUNG: Yes.
21 MEMBER SKILLMAN: And I'll just tell you, 22 my background was Bellefonte. I was one of the 23 original managers for, or B&W managers for Bellefonte.
24 So, we got well-schooled in the Sequatchie anticline, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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28 and the Lake Guntersville, and the meteorology down in 1
that section of Alabama.
2 But we were interacting with the teams 3
that were doing the other TVA plants at Sequoya and 4
Watts Bar, at Browns Ferry. And so, we got tuned into 5
different meteorologies at different locations.
6 I understand you to say, if you look at 7
the event frequency, and look at the meteorology, you 8
then come up with a probability of someone getting 9
dosed at 200 rem.
10 Does that say that if you put the plant at 11 Clinch River it might have one probability? And if 12 you put the plant at Sequoya or Watts Bar with a 13 different meteorology, that will be a different?
14 Okay. Now, hold that thought. How do you 15 predict that meteorology? Because it sounds to me 16 like you're using a probability riddle for a natural 17 event that, at least in my judgment is very variable.
18 The uncertainty has to be huge.
19 MR. YOUNG: So, the meteorology that we 20 used for this analysis, and for the additional pieces 21 of this are based on data collected from the site, and 22 analyzed over, you know, a period of time, in 23 accordance with, you know, applicable regulatory 24 guidance.
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29 MEMBER SKILLMAN: Over what period of 1
time?
2 MR. YOUNG: So, for SSAR Section 2.3, in 3
accordance with regulatory guidance 1.23, that comes 4
down to a minimum of two years of data.
5 MEMBER SKILLMAN: Why is two years 6
sufficient for a siting decision, when that site will 7
be employed potentially for 60 or 80 years?
8 MR. YOUNG: So, there are additional steps 9
that continue to -- So, in addition with monitoring 10 the site specific data over two years, you have to 11 compare that to historical pieces as well, and 12 different pieces in the area, to make sure that it's 13 representative of the site, and over a period of time.
14 In addition to that, there's also on site 15 monitoring that you continue to do over the life of 16 the plant.
17 MEMBER SKILLMAN: Thank you.
18 MEMBER MARCH-LEUBA: That's scary. You're 19 saying that I build my plant, I pay the money, and now 20 I have to monitor the wind. And if the wind gets off 21 outside you assume I lose my license?
22 MR. YOUNG: So, there's, to that question, 23 what we're looking at is, we have changes in 24 meteorology. We do a lot of analysis to, you know, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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30 show that that meteorology is consistent over a long 1
period of time. And we include abundant margin within 2
that meteorology to account for potential changes like 3
that.
4 MEMBER MARCH-LEUBA: So, you're hoping 5
that your monitoring is large enough that you'll never 6
get caught?
7 MR. YOUNG: Absolutely. Yes.
8 MEMBER MARCH-LEUBA: But you are running 9
the risk?
10 MR. YOUNG: That's an operational risk we 11 take.
12 MS. MANOHARAN: Okay. So --
13 CHAIRMAN CORRADINI: So, let me summarize, 14 since I started this. I want to make sure I am clear.
15 So, the one in a thousand is based on the site 16 meteorology, conditional on the fact that I've had a 17 severe accident of a certain frequency band. And is 18 it all those accidents that, and you look for the 19 worst source term of that grouping of accidents?
20 MR. YOUNG: So, that comes down to the 21 step of, you know, determine source term releases from 22 selected accidents in determining the selected 23 accident, that appropriate evaluation. So, you know, 24 as you go through this you'll come up with the, for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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31 the accidents that screen in you would come up with 1
the, you know, the bounding accident --
2 CHAIRMAN CORRADINI: Okay.
3 MR. YOUNG: -- evaluation.
4 CHAIRMAN CORRADINI: So, you're looking 5
for the bounding source term within that frequency 6
band. You then do the computation on some sort of 7
weather variability. And the weather variability is 8
what essentially the term is a one in a thousand? I 9
want to make sure I'm clear. Have I said it 10 correctly?
11 MR. YOUNG: Yes. The, yes.
12 MEMBER RICCARDELLA: So then, are we 13 really talking like probability to ten to the minus 14 nine? Yes. If we have a event probability of ten to 15 the minus six, and then the, if that event occurs the 16 probability of achieving this dose is --
17 MR. YOUNG: Yes.
18 MEMBER RICCARDELLA: -- ten to the minus 19 third. So, we're talking ten to the minus ninth?
20 PARTICIPANT: No.
21 MR. YOUNG: The essential. So, you would 22 have the even probability, which would be greater than 23 one E minus 7.
24 MEMBER RICCARDELLA: Yes. Somewhere 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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32 between seven, six and seven.
1 MR. YOUNG: And then you would apply the 2
factor to it, based on meteorology. And if the total 3
frequency of the 200 rem dose exceeds one E minus --
4 It has to be, at that distance you have to be within 5
a probability of one E minus three for the 200 rem.
6 MEMBER RICCARDELLA: So --
7 CHAIRMAN CORRADINI: You said it now.
8 MEMBER RICCARDELLA: So then, the real 9
probability of that occurring, of that event 10 occurring, and a person getting that dose is ten is to 11 the minus nine, or somewhere between ten to the minus 12 tenth and ten to the minus ninth, right?
13 MR. YOUNG: Yes. You'd have to have the 14 probability of the event --
15 MEMBER RICCARDELLA: Yes.
16 MR. YOUNG: -- first.
17 MEMBER RICCARDELLA: Yes.
18 MR. YOUNG: And actually --
19 MEMBER MARCH-LEUBA: You would have to 20 integrate --
21 MEMBER RICCARDELLA: It's a condition.
22 Yes.
23 MEMBER MARCH-LEUBA: -- year one through 24 1,000 what the consequences are. So, it's not ten to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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33 the minus nine. It's much, much higher.
1 MEMBER RICCARDELLA: Why?
2 MEMBER MARCH-LEUBA: Well, because --
3 MEMBER RICCARDELLA: Multiple events.
4 MEMBER MARCH-LEUBA: This is the 1,000 5
year methodology. You can have the 500 year 6
methodology, the 100 year methodology. All of those 7
give you those. So, you have to do the interval of 8
all of those to get that average. It's math.
9 MEMBER RICCARDELLA: But regardless, 10 that's a conditional probability, right? So, that 11 only applies if you have the event.
12 MEMBER MARCH-LEUBA: Ten to minus seven 13 you're giving with.
14 MEMBER RICCARDELLA: Yes.
15 MEMBER MARCH-LEUBA: Because that's when 16 you have the event.
17 MEMBER RICCARDELLA: Yes.
18 MEMBER MARCH-LEUBA: Now you're picking 19 the worst possible year in a 1,000 --
20 MEMBER RICCARDELLA: Yes.
21 MEMBER MARCH-LEUBA: -- to propagate it to 22 the end of EPZ. But if you had a better way you will 23 still propagate some dose.
24 CHAIRMAN CORRADINI: But a lower dose.
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34 MEMBER MARCH-LEUBA: It will be a little 1
lower dose with higher probability. So, you will have 2
to do some kind of interval. And I don't know how to 3
write it out right now.
4 MEMBER RICCARDELLA: I'll have Dennis, 5
I'll ask Dennis to explain it to me after the meeting.
6 MS. MANOHARAN: So, back on this slide, 7
this is the example analysis that was conducted as a 8
result of the staff's RAI. So, we use the NuScale 9
design at Clinch River site to do a demonstration, an 10 example demonstration, to show what the dose at site 11 boundary would result from the NuScale design.
12 So, as you can see for Criterion Alpha and 13 Bravo the doses are on, in that table. And they have 14 significant margin to the one rem limit. And there's 15 also additional margin built in within the calculation 16 that resulted in that example analysis.
17 Moving on to next slide, Slide number 20.
18 So as, both Dan and Ray had mentioned earlier, Part 5 19 of the application contains two major feature, two 20 emergency plans, major features of emergency plan.
21 One to support the site boundary EPZ, and the other 22 for the two mile EPZ.
23 Now, what they do is they, both of the 24 address the 16 planning standards of NUREG 0654. Once 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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35 a reactor design is selected for COLA you would do the 1
dose based methodology that Section 13.3 describes to 2
pick your EPZ size.
3 So, if it is site boundary, then you go 4
with 5 Alpha, and you would incorporate design 5
specific information, and create a complete and 6
integrated emergency plan.
7 If the dose is met at two miles you would 8
take the Part 5 Bravo, incorporate the rest of the 9
elements to make a complete and integrated emergency 10 plan.
11 If for some reason you pick a reactor 12 design that doesn't meet either site boundary or two 13 mile, then we would have to come up with a new 14 emergency plan and COLA. Next slide, please.
15 CHAIRMAN CORRADINI:
Just one 16 clarification. The thinking that you guys have come 17 with, with this either or approach is, the two miles 18 is bound to the EAB?
19 MS. MANOHARAN: So, the reason for two 20 emergency plans is, when the plant parameter envelope 21 was being developed at least one of them, we were 22 confident that at least one design would meet site 23 boundary EPZ. So, we pursued the site boundary 24 emergency plan.
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36 CHAIRMAN CORRADINI: But not all of them?
1 MS. MANOHARAN: We were confident that all 2
of them would meet --
3 CHAIRMAN CORRADINI: Okay.
4 MS. MANOHARAN: -- two mile.
5 CHAIRMAN CORRADINI: Okay.
6 MS. MANOHARAN: Therefore, the two mile.
7 MR. STOUT: And two miles was a surrogate 8
for scalable. You know, we, the staff had indicated 9
through SECYs a willingness to consider scalable EPZ.
10 We picked the number that we thought would bound all 11 four designs, and be representative of scalable.
12 CHAIRMAN CORRADINI: Can I torture you one 13 last time? So, did you do any sort -- Well, maybe I 14 should ask the staff this. Somebody should ask 15 someone this question, which is, if I did two years 16 and had the appropriate meteorology, and then I looked 17 back ten years, and I did the same thing, did I see a 18 big difference in the, I'll call it the uncertainty, 19 or the distribution function of the various types of 20 meteorology. Was this done?
21 MR. YOUNG: So, you're asking, so, we did, 22 we collected two years of onsite data. Did we look at 23 how that compared to, you know, a longer period of 24 time? Yes, we did.
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37 We did comparisons, you know, from data 1
that was collected from the breeder reactor project.
2 We also did comparisons to operating fleets, or our 3
operating fleet, data collected in surrounding stuff.
4 CHAIRMAN CORRADINI: Thank you.
5 MS. MANOHARAN: So, moving on to Slide 6
number 21. Part 6 of the application describes the 7
exemption requests that support the emergency 8
preparedness approach in the application.
9 So, if you look at Part 6 there are two 10 sets of exemption requests. One that support the side 11 boundary EPZ, and one that support two mile. As Dan 12 had mentioned, two mile is a surrogate for scalable.
13 And the only real exemption request we're 14 asking for in two mile EPZ is to deviate from the ten 15 mile. We understand that if we go with two mile then 16 there would be a need for formal offsite emergency 17 plans.
18 And for the site boundary, in addition to 19 deviate from the ten mile EPZ, some, various elements 20 of, let's say off site exercises and notifications, 21 evacuation time estimate analysis, we're taking 22 exemption, we're requesting exemptions from that.
23 MEMBER RAY: Excuse me. You said, if we 24 go with two mile. And then I couldn't understand what 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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38 you said after that.
1 MS. MANOHARAN: That there would be a need 2
for formal off site emergency response plans. So, 3
even if it is two mile, and not a ten mile, there 4
would still need to be an off site response structure, 5
if you will.
6 So, the site boundary EPZ is, let's say 7
the most restrictive, and has the most number of 8
exemption requests. And two mile is only asking to 9
deviate from the size of the EPZ.
10 MEMBER RAY: Thank you.
11 MS. MANOHARAN: Next slide, please. So 12 lastly, this is a summary slide that shows the 13 emergency preparedness information in the ESPA, and 14 how each of these pieces will be used in the COLA if 15 at all the COLA is pursued.
16 So, in Section 13.3, as we've been 17 discussing throughout this presentation, there's a 18 dose
- based, consequence oriented methodology 19 described. It's design neutral. It's not specific to 20 any one particular design that informs the PPE. And 21 we're asking approval of the methodology.
22 At COLA, once the reactor design has been 23 selected, we would implement the methodology with 24 design specific implementation, and figure out what 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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39 the EPZ size for that particular reactor design at the 1
site would be.
2 In Part 6 of the ESPA the set of exemption 3
requests that have been requested. And those would be 4
implemented based on the dose based methodology 5
results at COLA. So, at COLA we would seek approval 6
of a design specific plume exposure pathway EPZ size 7
for the reactor design selected.
8 Lastly, the emergency plan, Part 5, two 9
distinct major features of an emergency plan for site 10 boundary and two mile are represented in the ESPA. At 11 COLA, after the dose based methodology is implemented, 12 the final EPZ size has been determined, we would pick 13 the appropriate emergency plan.
14 It could be the site boundary in Part 5 15 Alpha, or Part 5 Bravo, or a new design, a new EP 16 based on the reactor design. And we'll create a 17 complete and integrated plan. And the next one? And 18 that concludes our portion of the presentation. Thank 19 you for the opportunity today.
20 MEMBER KIRCHNER: Thank you. I think what 21 we, what you're hearing from us is, we, since we met 22 last we've been struggling with understanding exactly 23 how -- In NUREG 0396 they have a figure. It's, for 24 the record I'll cite it.
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40 It's Figure I-11, Page I-38, which is how 1
that task force actually came to the recommendation 2
for the ten mile EPZ for the larger fleet of reactors 3
that existed. And this was shortly after WASH-1400, 4
the Reactor Safety Study.
5 So, it appears to us that this is an 6
integrated curve, as Member Rempe is pointing out.
7 It's giving us probability, but not per reactor year.
8 It takes a probability based on a conditional core 9
melt of, on the order of ten to the minus five at the 10 time.
11 And
- then, with that source
- term, 12 propagates with, in this case they use straight line 13 plume trajectories for the weather. And then we're 14 able to come up with isoclines, so to speak, of dose 15 versus distance.
16 So, that's the historical basis and 17 background for the current ten mile. What has been 18 puzzling us is, and what you're proposing, how you go 19 through the calculation once you have a given, either 20 a class of accidents that are severe, or even a 21 dominant accident.
22 It's clear to us how you used meteorology 23 to propagate dose. But it's not clear how this 24 probability of ten to the minus third is arrived at.
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41 So, perhaps that's a question we also take up with the 1
staff. Okay. So, that's the concern. I hope I've 2
summarized well enough why we're puzzling collectively 3
here.
4 The methodology in principle makes sense.
5 But we're, we have been puzzling over just why this is 6
probability per reactor year. As Member Riccardella 7
pointed out, a simplistic approach might be to 8
multiply the two together and get very low numbers, 9
not what are reason -- what are indicated as a fairly 10 high number, one in a thousand.
11 CHAIRMAN CORRADINI: We're engineers. We 12 want to get the mechanics right.
13 MEMBER KIRCHNER: Thank you.
14 MEMBER REMPE: Actually, again, because we 15 were chatting, and trying to figure out what was going 16 on, and I have not attended all your Subcommittee 17 meetings.
18 But you mentioned at the beginning what if 19 the, or someone asked, what if they don't even have a 20 source term at something ten to the minus eight? They 21 can't get something out. And you said that there was 22 some sort of example you were going to show us.
23 And, was I distracted, and I missed what 24 you were going to do if they have no source term, for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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42 example?
1 MS. MANOHARAN: So, what I was mentioning 2
for that question is that for severe accidents if, 3
what if you pick a reactor design that does not have 4
a accident that screens in -- Can you go to one?
5 MEMBER REMPE: Yes. What if it's like ten 6
to the minus ten?
7 CHAIRMAN CORRADINI: Screen it out.
8 MEMBER REMPE: Yes.
9 CHAIRMAN CORRADINI: Then it's not there.
10 MEMBER REMPE: If you totally --
11 MS. MANOHARAN: Yes.
12 MEMBER REMPE: And what, you're going to 13 force them, you said earlier, to come up with 14 something.
15 MS. MANOHARAN: Yes. So, there is an 16 additional --
17 (Off microphone comments.)
18 MS. MANOHARAN: Yes. So, for Criterion 19 Bravo there is an additional note in our methodology 20 that says that even if there are accidents that are, 21 that screen in based on your reactor design, you still 22 have to create a source term, an alternative source 23 term to analyze the severe accident. So --
24 MEMBER REMPE: And again, I haven't 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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43 attended your Subcommittee meeting, but remind me what 1
are you going to do for how they created? Or that's 2
to be determined, on how they're going to generate 3
something?
4 MS. MANOHARAN: So, I think Alex can speak 5
a little bit on that. But I will say, for example --
6 let's go to the next one. Sorry to keep jumping. So, 7
this is the NuScale example, as I was mentioning.
8 It's just an example to show how the methodology would 9
be implemented.
10 So, Criterion Alpha would be the design 11 basis accidents from NuScale's Chapter 16 analysis.
12 And then Bravo would be the severe, less severe core 13 melt accidents.
14 So, I will walk through the example, and 15 what accidents screened in, and why it would make 16 sense to have an alternative source. So, if we pick 17 a reactor design that doesn't have screening.
18 MEMBER REMPE: So, with your example, 19 which you claim is associated with NuScale, they 20 generated an alternate source term? And that's really 21 beyond your methodology. You don't know how they did 22 it. They just came up with something that was their 23 alternate source term?
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44 don't, for example, the accidents that resulted in 1
this example analysis for Criterion A would be the 2
design based accidents, which is a combination of 3
what, their LOCA and other accidents.
4 So, it's not just the design based 5
accidents. So, it's more representative of their 6
Criteria Bravo also. And then, Bravo was their most 7
probable accident, which is the loss of DC power 8
sequence, the most probably accident.
9 MEMBER REMPE: So, they didn't have to go 10 to some -- Or did they tell you what the frequency was 11 for those type of events?
12 MS. MANOHARAN: I think we know the 13 answer.
14 MR. YOUNG: So, we do know what the 15 frequencies are associated with those sequences that 16 informed their design basis accident analysis. But 17 those are proprietary to NuScale.
18 MEMBER REMPE: Okay. So, let's ask it in 19 a way that -- They basically picked something below 20 ten to the minus eight, and they went ahead and moved 21 it up.
22 So, basically you're kind of forcing them.
23 So, I'm glad I brought this up, even though I may have 24 missed some of the details.
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45 MS. MANOHARAN: You may, yes.
1 MEMBER REMPE: But they basically agreed 2
to just take a hit --
3 MS. MANOHARAN: It's several magnitudes 4
lower. So --
5 MEMBER REMPE: Yes. So, they basically 6
agreed to take a hit, just so that they could do 7
something.
8 MS. MANOHARAN: Because of the note in our 9
methodology that you have to do the analysis.
10 MEMBER REMPE: Okay. And they were okay 11 with that? Okay. Thank you.
12 MS. MANOHARAN: And that information, I do 13 want to just, that information is in an RAI response 14 to the staff. So, the staff has seen that analysis.
15 MEMBER REMPE: Thank you.
16 MEMBER RICCARDELLA: Could I ask why Row 17 B has a higher dose, site boundary dose, than Row A?
18 And it's less severe?
19 MR. YOUNG: So, Row A is based on 20 NuScale's Chapter 15 design basis accident analysis.
21 And Criterion A, the accident or sequences that were 22 evaluated for that are based on several accidents, you 23 know, happening. Criterion B is just looking at one 24 of those accident sequences that informs A, which is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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46 a --
1 MEMBER RICCARDELLA: Okay.
2 MR. YOUNG: It's a more severe accident.
3 MEMBER BLEY: Design basis accidents 4
aren't core melt accidents.
5 MEMBER RICCARDELLA: Okay.
6 MEMBER BLEY: Are not. So, the next one 7
is more severe.
8 MEMBER KIRCHNER: Okay. Well, at this 9
point then, if there are no further questions of the 10 applicant from the members at this point? Okay.
11 Well, let's change then to your team. Andy, please.
12 (Pause.)
13 CHAIRMAN CORRADINI: Okay. Mallecia, 14 Allen? Who's going to lead off?
15 MR. FETTER: I'm going to start. Just 16 getting us started here.
17 (Off microphone comments.)
18 MR. FETTER: Okay. Just my screen looked 19 a little different. So, I was a little confused 20 there. Good afternoon. I'm Allen Fetter. Mallecia 21 Sutton and I are the safety project managers for the 22 Clinch River nuclear
- site, early site permit 23 application.
24 And I will be presenting an overview of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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47 the staff's findings and recommendations, which were 1
discussed at the four previous ACRS Subcommittee 2
meetings. The technical reviewers are also here to 3
address questions in their technical areas that, any 4
questions you have during the presentation.
5 TVA submitted an early site permit 6
application for the Clinch River nuclear site on May 7
26, 2016. The application was accepted for detailed 8
technical review and docketing on December 30th, 2016.
9 TVA requested a permit approval for a 20 10 year term, along with approval for a plume exposure 11 pathway, or PEP, emergency planning zone, sizing 12 methodology, two major features, on site emergency 13 plans and exemption requests for site boundary and two 14 mile PEP EPZs. The plant perimeter envelope was based 15 on four small modular reactor designs.
16 A staff overview presentation to ACRS on 17 the Clinch River ESP was given on November 15th, 2017.
18 The NRC staff's safety review of the application 19 included execution of five audits, and one inspection, 20 and issuance of 12 RAIs, comprising 50 questions.
21 The staff completed all advanced safety 22 evaluations with no open items, and presented their 23 findings at four ACRS Subcommittee meetings between 24 May 15th, 2018 and November 14th, 2018. The advanced 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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48 safety evaluations include 42 COL action items and 1
eight permit conditions.
2 Staff cooperated with the U.S. Army Corps 3
of Engineers, consulted with the Federal Emergency 4
Management Agency, and engaged with the Department of 5
Energy, the Tennessee Department of Environment and 6
Conservation, and the U.S. Geological Survey, and the 7
Tennessee Emergency Management Agency.
8 So, an early site permit plant parameter 9
envelope values can bound a variety of reactor 10 technologies, rather than one specific technology, an 11 amalgam of values representing a surrogate nuclear 12 plant.
13 The PPE values are bounding criteria used 14 by staff to determine the suitability of an ESP site 15 for construction and operation of a nuclear plant.
16 In the combined license application, when 17 a specific technology is identified the PPE values are 18 compared to those of the selected technology.
19 If design parameters of the selected 20 technology exceed bounding ESP PPE values additional 21 reviews are conducted to ensure that the site remains 22 suitable from a safety and environmental standpoint 23 for the construction and operation of the selected 24 nuclear plant technology.
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49 MEMBER KIRCHNER: Allen?
1 MR. FETTER: Yes.
2 MEMBER KIRCHNER: I'm going to interrupt 3
at this point. I was going to ask this later in your 4
presentation. Maybe I'll just put this down. And 5
maybe you can address it later.
6 The, one of your permit conditions that 7
you're going to share with us is the use of the Table 8
13.3-1, which is the PPE set of source terms by 9
isotopes. And what if there's a variance in that?
10 Or are you confident that, maybe it's a 11 question of the applicant as well, that if something 12 in the fuel cycle that is used, we know that they're 13 using LWR derivative fuel in most of the concepts that 14 are under consideration.
15 But what if there's a variance in that 16 table, that they exceeded one of these radionuclide 17 amounts with the concept that they chose to go forward 18 with, that COL point? What happens then?
19 MS. SUTTON: So, during the exemption and 20 presentation Michelle will discuss that --
21 MEMBER KIRCHNER: Okay.
22 MS. SUTTON: -- in more detail.
23 MEMBER KIRCHNER: Excellent. Okay.
24 MS. SUTTON: Thank you.
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50 MEMBER KIRCHNER: Thank you.
1 MS. SUTTON: You're welcome.
2 MR. FETTER: Okay. As stated before, the 3
plant parameter envelope is based on four modular 4
reactor
- designs, mPower,
- NuScale, Holtec, and 5
Westinghouse. TVA's PPA is based on construction and 6
operation of two or more SMRs at the Clinch River 7
nuclear site, with a generating capacity of 2,420 8
megawatts thermal, or 800 megawatts electric.
9 Okay. This slide is for ACRS records. It 10 depicts all of the advanced safety evaluations, and 11 their associated accession numbers in ADAMS, that were 12 provided for all the ACRS Subcommittee meetings.
13 MEMBER RAY: There's no assumption at this 14 point as to the number of units that might be affected 15 by any of the events described, right? It could be 16 one. It could be all. Is that correct?
17 MS. SUTTON: So, during the exemption 18 presentation -- This is just a overview of the staff's 19 safety evaluation. We will address all those in 20 details for you. I promise. So, hold that thought.
21 MEMBER RAY: So, I will indeed. But so, 22 we're going to find, the answer is different for each 23 of these. Is that what I just heard you say?
24 MS. SUTTON: No. That's not what I said, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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51 sir. So, Michelle, do you want to --
1 (Off microphone comments.)
2 MS. SUTTON: Okay. Go ahead. Ask the 3
question one more time.
4 MEMBER RAY: Is there any assumption in 5
what we're reviewing here, in a multi module site, 6
that only one of the modules will be affected at a 7
time? I'm looking at events here that include 8
vibratory ground motion, for example.
9 MS. SUTTON: Does any of the staff like to 10 address the question?
11 MR. CAMPBELL: Well, let me address that.
12 This is a plant parameter envelope for an ESP. There 13 are a variety of assumptions that are put in by both 14 the applicant and the staff in its review.
15 And with that said, we're looking at a 16 number of different scenarios within that plant 17 parameter envelope. So, that's how it's developed.
18 And the plant parameter envelope encompasses all the 19 designs. So, I don't know if TVA wants to 20 specifically --
21 MR. FETTER: It looks like Alex --
22 MR. CAMPBELL: -- address that question.
23 MR. FETTER: -- wants to come to the --
24 MEMBER RAY: Before they do, let me just 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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52 say I would interpret what you just said to be that 1
no, there's no assumption in what we're reviewing now 2
that only one module would be affected by an event.
3 That's how I interpret what you just said.
4 MR. CAMPBELL: In some scenarios that go 5
into the plant parameter envelope, and someone who's 6
actually an expert in this can correct me if I'm 7
wrong.
8 There are scenarios where there's one 9
module. There are scenarios where there's more than 10 one module, if it makes sense. And, you know, the 11 frequency of occurrence of more than one module is 12 within that range that should be considered.
13 MEMBER RAY: Okay. Well, I think that 14 you've answered the question. I'm at least going to 15 understand it to be that we aren't limiting 16 consideration to only a single module being affected 17 in what we're discussing now. But that's my 18 understanding of what you just said.
19 MR. CAMPBELL: And we'll confirm that.
20 MEMBER RAY: Thank you.
21 MR. CAMPBELL: Okay.
22 MEMBER REMPE: Actually, there was a guy 23 from TVA who might be able to confirm it now.
24 MR. CAMPBELL: Yes.
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53 MEMBER REMPE: And since we're doing our 1
letter in the next --
2 MR. CAMPBELL: Yes.
3 MEMBER REMPE: -- few hours --
4 MEMBER RAY: Thank you.
5 MEMBER REMPE: Yes. I'd like to hear --
6 MEMBER RAY: Yes.
7 MEMBER REMPE: -- his response --
8 MEMBER RAY: We would too.
9 MEMBER REMPE: -- to my --
10 MR. YOUNG: Sure. So, my name's Alex 11 Young. So, the question was about multi module 12 accidents for the ESPA. Currently the way we've 13 assessed the ESPA, based on the plant parameter 14 envelope, the inputs that we have do not assume any 15 multi module accidents. They're all based on single 16 unit accidents, or single units events.
17 At the COLA stage, depending on the design 18 selected, that's something that would have to be 19 evaluated based on the design. But currently the 20 assumption for the ESPA is only single module events.
21 MEMBER RAY: But what's the basis for 22 that?
23 MR. YOUNG: The basis for that is based on 24 the design information that we have available at the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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54 time for input into the ESPA for the PPE. We don't 1
project, or believe that there are going to be multi 2
module events that have to be considered for this.
3 MEMBER RAY: But that's a belief, as you 4
express it, that I don't understand the basis for.
5 MEMBER RICCARDELLA: But if your plant 6
parameter -- Excuse me, Hal, I'm sorry. If your plant 7
parameter envelope is based on 800 megawatts, then 8
doesn't that automatically address, doesn't that 9
automatically cover multi-unit accidents for the 10 smaller module, for the smaller units?
11 MEMBER RAY: Well, I don't know if you're 12 asking me or not --
13 MEMBER RICCARDELLA: No. I'm asking TVA.
14 MR. YOUNG: So, part of the piece here is 15 design basis accidents versus beyond design basis 16 accidents. So, there's Chapter 15 analysis, design 17 basis accidents, which for the information we have 18 right now doesn't consider those multi module 19 accidents, based on the design information that we 20 have currently, when we developed this.
21 For the EPZ portion, you know, we have to 22 consider those multi module accidents. And at COLA we 23 still have to, you know, go and consider the 24 possibility of those multi module accidents for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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55 Chapter 15 as well. So --
1 MEMBER RAY: Yes. I understand it at the 2
COLA period. Whether or not the early site permit 3
parameters fit within what is then being proposed of 4
the COLA is one of the issues that is, necessarily has 5
to be addressed at that time.
6 But I guess it's just, you said, based on 7
your understanding of the plants, this -- What's being 8
described here isn't just a early site permit boundary 9
based on a limiting size accident. You're actually 10 talking about multiple units.
11 And now you're saying that the assumption 12 is based on an understanding which is not part of this 13 process. That only one of them at a time will be 14 affected. And I just want to be clear that that's 15 what's going on.
16 MR. CAMPBELL: At the stage of the COL the 17 applicant would have to, with the specific design.
18 Because the applicant here for the ESP looked at a 19 range of different designs.
20 At the COL stage, when you have a specific 21 design, then you can do that type of analysis, and 22 establish what that is. And if that exceeds the 23 parameter, then they would have to take a deviation or 24 exemption.
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56 MEMBER REMPE: So --
1 MEMBER RAY: Although the applicant did 2
what you described, it doesn't sound like we did what 3
you describe.
4 MR. CAMPBELL: our review is based upon 5
the ESP, not on what will be done at the COL stage.
6 MEMBER RAY: I guess I'm asking, why is 7
it, why are we even talking about multiple units, only 8
one of which has an accident at a time? Why is that 9
part of the discussion at this point?
10 I mean, I understand establishing an ESP.
11 I don't understand talking about multiple units, only 12 one of which is assumed to have an accident at a time, 13 based on information that isn't part of this 14 application.
15 I mean, I know that when the COL comes up 16 this can be addressed. I grant that. But I don't 17 understand why we're doing what we're doing at this 18 point, relative to limiting the assumed event to one 19 of several units that will be at the site that we're 20 talking about.
21 And with that, I guess we ought to just 22 leave it there and move on. I just don't understand 23 it. At least we ought to be clear that that's what's 24 happening.
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57 MEMBER REMPE: If we could have TVA come 1
back up for a second to the mic? I had a question, 2
and I didn't get to get it in in the discussion.
3 Okay. So, you did, as we talked about, you came up 4
with some alternate source term based on a
5 hypothetical 6
What if you learn more about one of these 7
plants, and they determine that multiple modules are 8
involved. How do you think that would affect your 9
process you've developed here?
10 MR. YOUNG: So, our process does, you 11 know, this is specifically talking about the EPZ 12 methodology. So, that methodology does require us to 13 look at those multi module events. And we'll look at 14 those.
15 In our example analysis, for instance, for 16 one of the events that we did look at as we were going 17 through the screening criteria for Criterion C, one of 18 those was considered a multi module event. It's a 19 beyond design basis event. But it was required to be 20 considered based on our methodology in the initial 21 screening.
22 The second screening portion, so we have 23 the E to the minus eight screening, and then we have 24 the second screening for, at a greater frequency. It 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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58 was then excluded from that, because it was only a 1
single event. So, it didn't meet the second screening 2
criteria.
3 MEMBER REMPE: So, basically if you, if 4
they learn something new about their plant, and not 5
picking on any particular one, and they decide 6
suddenly, well, both modules or all 12 modules are 7
going to be impacted by an event at a much higher 8
frequency, your process could accommodate it?
9 MR. YOUNG: Yes. We have to consider 10 that, yes.
11 MEMBER REMPE: Okay.
12 MEMBER SKILLMAN: Can you accommodate it 13 without shopping for new meteorology? Yes. I'm 14 pulling your leg. But I'm serious on the question.
15 MR. YOUNG: So, from what we know about 16 the example analysis we considered, we would be able 17 to meet our Criterion C dose requirements based on 18 that, you know, assumed, if we assume that that multi 19 module accident screened in. That would meet the dose 20 criteria. It depends on the accident that would 21 screen in. So --
22 MEMBER SKILLMAN: Okay. Thank you.
23 MEMBER RAY: Well, I just want to talk to 24 the NRC at this point, not to the applicant. I just 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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59 think it needs to be really clear what our 1
understanding is of what the applicant is assuming in 2
connection with this ESP. Because it was not 3
something that I thought was explicit or clear at all.
4 MEMBER KIRCHNER: If I might summarize my 5
understanding at this point? It's going in, the 6
applicant has bounded the source term up to a single 7
unit of 800 megawatts thermal.
8 And they've deferred on the multi-unit, 9
say common cause, common mode failure kind of concerns 10 until the COLA application, the COL application. And 11 a PRA that would then have to be examined to see 12 whether a multi-unit failure of some kind, or accident 13 sequence would then lead to a source term that would 14 exceed what they're currently asking for, as an 15 exemption for either the one mile or, not one mile, 16 the site boundary or two mile boundary.
17 If they come in at that point, and don't 18 screen out multi-unit failures, and find that the dose 19 exceeds the envelope, then they are not going to be in 20 a position to get this exemption. Of course --
21 MR. CAMPBELL: They would have to develop 22 additional information at the COL stage to demonstrate 23 what that boundary, what the size of the EPZ would be, 24 given those considerations. There's a full blown PRA 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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60 done at the COL stage for a specific design.
1 Part of the issue here is, we're not 2
approving results on the basis of only one specific 3
design. What we're approving is a methodology. As I 4
said in my opening, this is an approval of a 5
methodology that can then be applied at the COL stage.
6 And the exemptions are to the, essentially 7
the requirements with respect to the ten mile EPZ.
8 That doesn't mean we're automatically approving either 9
a site boundary or a two mile EPZ for a COL applicant.
10 They have to make their case.
11 MEMBER RAY: Well, it's, I'm not sure that 12 the issue of multi-unit failure isn't going to be 13 addressed through the DCD, much less, not necessarily 14 in the COL stage. But in any event, all I'm trying to 15 do is figure out why, what we're assuming, and why 16 we're assuming it. So that it's clear.
17 MEMBER DIMITRIJEVIC: Can we go to Slide 18 number 5? Because it will be clear what we are 19 asking. Because it says in that, PPEs based on 20 construction and operation of two or more SMRs at the 21 Clinch River site.
22 MEMBER RAY: Where, what are you saying, 23 Vesna? I'm sorry.
24 MEMBER DIMITRIJEVIC: That in the last 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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61 paragraph, says that this PPE, the plan parameter 1
envelope is based on construction and operation of two 2
or more SMRs.
3 MEMBER RAY: Yes. So let's --
4 MEMBER DIMITRIJEVIC: So, why are we 5
talking two or more?
6 MEMBER RAY: That's why I'm asking the 7
question is, whether or not we assumed only one of 8
these, or more, suffered a release that, is what we're 9
talking about here in setting a boundary. And if we 10 only assumed one, why?
11 (Off microphone comments.)
12 MS. HART: All right. This is Michelle 13 Hart, from the staff. I didn't do the Chapter 15 14 analysis. But I understand the Chapter 15 analysis.
15 So, in general terms, the plan parameter 16 envelope is developed based on current information, 17 and does include consideration of one unit at a time, 18 because we are, there's a presumption that GDCs 2, 4, 19 and 5 will be complied with, so that you won't have 20 common cause failures.
21 That you won't have, you know, much like 22 you don't look at siting for more than one unit, at 23 the currently operating plants we thought that that 24 would also apply to a multi module site, until told 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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62 differently from the specific design.
1 MEMBER RAY: Well, I understand. And I've 2
operated a multi-unit site. An I know exactly what 3
you're talking about. But it's also why I'm asking 4
the question. Because it's not a resolved issue. And 5
the only thing at the end of the day I'm seeking, is 6
for us to be clear about what we're doing.
7 And I don't want anybody later to believe 8
that what we have done here is agree that only a 9
single unit in a multi-unit site need be assumed to 10 fail. Notwithstanding multi-unit sites today that 11 exist today elsewhere. I understand that very well.
12 MS. HART: I think the thing is that the 13 information that we have, Chapter 15 was based on a 14 non multi module unit. And so, the single unit was 15 bounding --
16 MEMBER RAY: Exactly. That's right.
17 MS. HART: And so, it's, I hope it's clear 18 that that's what we did. But there's no prevention of 19 saying that if something came in for the COLA to use 20 this ESP, if it doesn't fit within that PPE, whether 21 it's a single unit or multi module event, that they 22 would have to take a variance, and have to describe it 23 more clearly.
24 MEMBER RAY: Yes. I mean, we think about 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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63 this in the DCD world also. And so, it's not just 1
when a COLA comes in on an ESP, for a given site, I 2
mean. Anyway, I think we've taken enough time here.
3 Again, my goal isn't to try and change what's 4
happened. I just wanted to be really clear about the 5
basis for what I --
6 MEMBER RICCARDELLA: But isn't it fair to 7
say we're approving a methodology to set the EPZ based 8
on probabilities of various events? And when you get 9
to either the DCD stage or for the COL stage, you're 10 going to have a PRA that talks about the probability 11 of single unit --
12 MEMBER KIRCHNER: And multi-unit.
13 MEMBER RICCARDELLA: -- and multi-unit 14 events. And if any of those multi-unit events trigger 15 these probability limits, they're going to have to be 16 considered, right?
17 MEMBER RAY: Well, you're not going to be 18 able to do that given the way the DCDs are envisioned 19 today, as the design certification being approved.
20 You're not going to have the information you're 21 talking about.
22 MEMBER RICCARDELLA: Yes. So, maybe it's 23 the COL stage. But at some stage you have to --
24 MEMBER KIRCHNER: I think it is the COL 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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64 stage.
1 MR. CAMPBELL: Yes. That is correct.
2 It's at the COL stage when they have to do a full 3
blown PRA.
4 MEMBER RICCARDELLA: Yes.
5 MR. CAMPBELL: If the frequencies of a 6
multi-unit failure at a site are low enough that they 7
don't have to be considered, they aren't considered.
8 But if they're high enough, for a variety of reasons 9
that may not be apparent at this stage, when we don't 10 really have --
11 We have designs. But we have designs 12 with, that really aren't solid, not necessarily 13 approved at this point in time. In fact, we have no 14 approved design at this point.
15 When you get to that, that's where you 16 apply this detailed look at multi-unit failures that 17 could exceed the cut off likelihood in terms of CDF.
18 That's where this is done. It's done at the COL.
19 There are a lot of COL action items within 20 an ESP that are simply saying, this is not an item we 21 can make a decision on at this time, because we just 22 simply don't have a design. We have a range of 23 designs we're considering. And that's the way we've 24 been doing ESPs now for five ESP permits so far. And 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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65 we're in, we are consistent with that approach.
1 There are a lot of things that don't have 2
all the information for at this time. But we have 3
enough information to establish what the methodology 4
is, and enough information to establish that one could 5
come in with a design that might meet the site 6
boundary, or two mile, or some other EPZ distance.
7 It might not be two miles. It might be 8
three, or it might be one. But if it goes beyond the 9
site boundary -- So, all of those things are covered 10 in the COL.
11 And they're, I don't know the exact number 12 from the SEs. But there are a large number of COL 13 action items that we'll notify the COL applicant, you 14 have to deal with this.
15 MEMBER RAY: You know, having sought an 16 ESP I do understand and agree. What I was trying to 17 understand is why we were going beyond what I think 18 you said, to talk about multi module plants, implying, 19 I thought, that we were only going to assume one of 20 the modules have an event at a time.
21 And it was the additional small modular 22 concept that I was questioning, not that the ESP goes 23 beyond where it has traditionally gone in the past.
24 MEMBER BLEY: I kind of like everything 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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66 you said. What we're so, the one thing I would 1
mention twice, you've said at the COL stage there's a 2
full blown PRA. So far no COL applicant has performed 3
a full blown PRA. They've deferred a lot of the 4
detailed issues until just before fuel load.
5 MEMBER RICCARDELLA: Getting later and 6
later. What do you do if you get to that stage and 7
you say, woops. The zone has to be three miles, not 8
two miles. That would probably be problematic.
9 MEMBER KIRCHNER: Harold, thank you. The 10 clarity is needed. Let us address that when we 11 deliberate over our letter on this matter and move on 12 in the interest of time, Allen, if you could.
13 MR. FETTER: Yes. And in the interest of 14 time I'm going to go over the next few slides rather 15 quickly so that we can get to the staff's review of 16 13.3 and the exemption request. With that being said, 17 if ACRS has any questions, please don't hesitate to 18 interrupt.
19 Okay. For geography and demography, the 20 staff review is based on information provided by the 21 Applicant and the staff's independent confirmatory 22 evaluation. Staff found that information to be 23 acceptable. It meets the requirements of 10 C.F.R.
24 100.20.
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67 For Section 2.2, nearby industrial 1
transportation and military facilities, based on the 2
information provided -- oops, we're still on that --
3 by the Applicant and staff's independent confirmatory 4
evaluation, the staff found the information to be 5
acceptable as information meets the guidance provided 6
in NUREG 0800, Section 2.2.1 to 2.2.2.
7 Meteorology, discuss the site-specific 8
information related to regional climatology, local 9
meteorology, onsite meteorological monitoring, and 10 long and short term atmospheric dispersion estimates.
11 As noted on the
- slide, site 12 characteristics related to extreme weather were found 13 to be acceptable for the Clinch River site. The 14 onsite meteorological monitoring system was found to 15 provide adequate data to represent the meteorological 16 dispersing conditions at the site.
17 Site characteristics related to short term 18 and long term atmospheric dispersion estimates were 19 found to be acceptable. Based on information provided 20 by the Applicant, the staff found all regulatory 21 requirements to have been satisfied with no open 22 items.
23 Okay, Slide 10, short term or accident 24 atmospheric dispersion factors are X/Q. Estimates 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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68 were developed for the exclusion area boundary and 1
outer boundary of the Low Population Zone.
2 The exclusion area is defined in 10 C.F.R.
3 50.2 as that area surrounding the reactor in which the 4
reactor licensee has the authority to determine all 5
activities, including exclusion or removal of 6
personnel and property from the area.
7 10 C.F.R. 50.2 also defines the Low 8
Population Zone as the area immediately surrounding 9
the exclusion area which contains residents, the total 10 number and density of which are such that there is 11 reasonable probability that the appropriate protective 12 measures can be taken on their behalf in the event of 13 a serious accident.
14 TVA used the NRC endorsed PAVAN 15 Atmospheric Dispersion Model to estimate X/Q values 16 for the zero to two-hour timeframe at the exclusion 17 area boundary as well as the longer timeframes noted 18 on the slide for the outer boundary of the Low 19 Population Zone.
20 These X/Q values are intended to represent 21 dispersion conditions that exceed no more than five 22 percent of the time for the Clinch River site. The 23 X/Q values, in conjunction with the estimated source 24 term discussed in Chapter 15, are used to demonstrate 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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69 compliance with 10 C.F.R. 5217 dose guidelines for 1
design basis accidents.
2 Those dose guidelines include 25 rem TEDE 3
for any individual located at the exclusion area 4
boundary for two hours and 25 rem TEDE for any 5
individual located at the outer boundary of the Low 6
Population Zone for 30 days. I will now turn it over 7
to Mallecia.
8 MS. SUTTON: For Slide 11, for Section 9
2.4, hydrologic engineering, TVA proposed adequate 10 site characteristics and boundary design parameters 11 for the inclusion in the early site permit. Design 12 basis flood and maximum groundwater levels, and the 13 accidental
- release, those estimates meet the 14 regulatory requirements.
15 Staff concludes that the Applicant meets 16 the early site permit regulatory requirements 17 associated with the hydrologic engineering.
18 Slide 12, please. For geological site 19 characterization, Section 2.5.1, vibratory ground 20 motion, Section 2.5.2, surface deformation, Section 21 2.5.3, stability of subsurface materials and 22 foundations, Section 2.5.4, stability of slopes, 23 Section 2.5, based on evaluation of the information 24 provided by the Applicant, and supplemented by 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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70 knowledge gained through staff direct examination 1
during site audits, the staff found Applicant 2
adequately characterized the site in these topic areas 3
in accordance with the applicable guidance.
4 Slide 13, please. Section 3.5.1.6, 5
aircraft hazards, staff agrees with Applicant's 6
conclusion that an aircraft crash probability is about 7
an order of magnitude of ten to the negative seven per 8
year or less and meets the provided NRC guidelines.
9 Staff finds that the Applicant's approach is 10 reasonable, and the probability value is acceptable.
11 Slide 14, please. So Chapter 11, 12 radioactive waste management, Section 11.2.3 and 13 11.3.3, based on the staff's review of TVA's early 14 site permit application, and subject to the staff's 15 identifying several action items, the staff concludes 16 that the normal plant permit, effluent source terms, 17 and offsite dose meet the applicable regulatory 18 requirements and are without undue risk to the public 19 health and safety.
20 Slide 15, please. Chapter 15, accident 21
- analysis, staff evaluated the application and 22 concluded that the Applicant's analysis meets the dose 23 criteria specified in the PPE, includes a bounding 24 accident release for the determination.
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71 Slide 16, please. Section 17.5, quality 1
assurance program description, staff evaluated the 2
application and concluded that the Applicant's quality 3
assurance program description for the Clinch River 4
nuclear site ESP application meets the requirements of 5
10 C.F.R. Part 50, Appendix B, and 10 C.F.R.
6 50.17(a)(1).
7 Slide 17, please. Now that we have 8
discussed all of the topic areas and their findings, 9
the staff will now describe the evaluation emergency 10 planning and related exemption requests. Recognize 11 that TVA early site permit application was submitted 12 in May --- in 2016.
13 This was before the staff started work on 14 the small module reactor and other new technologies' 15 rulemaking. According, the application and the review 16 of the application by the staff is based on the 17 current regulations and guidance.
18 TVA's early site permit application 19 includes a methodology that, if approved in the early 20 site permit, would be used in future combined license 21 application and represents the specific merger reactor 22 design and early site permit to determine the 23 appropriate site-specific plume exposure pathway 24 emergency planning zone size for the Clinch River 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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72 nuclear site.
1 The submitted early site permit 2
application requests exemption from certain emergency 3
planning zone requirements if certain conditions are 4
met. If these sorts of exemptions are approved as 5
part of the early site permit, they will be 6
accompanied by permit conditions specifying the 7
circumstances under which these plans can be used in 8
the combined license application.
9 If the exemptions are approved in the ESP, 10 this Applicant can adopt these exemptions if it shows 11 that its COLA PEP EPZ source term releases to the 12 atmosphere are bounded by the non-design specific 13 plant parameter source term information developed for 14 the early site permit.
15 A future CO application featuring an SMR 16 design, that fits within the plant parameter envelope 17 established in the ESP, could apply the plume 18 methodology to the design selected to determine the 19 appropriate PEP EPZ size for the site and also 20 demonstrate whether the conditions for either of the 21 two sets of exemptions have been met.
22 The safety evaluation report for Chapter 23 13, Section 13.3 for the TVA Clinch River nuclear site 24
--- early site plan application addresses the plans, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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73 design features, facilities, functions, and equipment 1
necessary for the meteorological emergency planning 2
that must be considered in an early site permit 3
application that includes proposed major features of 4
the emergency plans.
5 Now I'll turn the presentation over to 6
Bruce and Michelle.
7 MR. MUSICO: Thank you. My name is Bruce 8
Musico.
I'm a
senior emergency preparedness 9
specialist. I and Michelle Hart reviewed the 10 emergency planning information that TVA submitted in 11 its ESP application.
12 The next two slides are a somewhat reduced 13 version of the slides we presented before the 14 subcommittee on August 22nd. And I refer you to the 15 transcript from that day, because it provides more 16 detailed explanation as well as answers to many of 17 your questions from the subcommittee.
18 For emergency
- planning, the ESP 19 application requested review of three key areas, and 20 you're going to see an overlap with TVA's presentation 21 as well, three key areas which consist of, first, the 22 plume exposure pathway, the emergency planning zone 23 sizing methodology, which Michelle Hart will discuss 24 in detail shortly.
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74 Secondly, the two major features, onsite 1
emergency plans which were contained in Part 5 of the 2
application, these include Part 5(a) which reflects a 3
site boundary plume exposure pathway emergency 4
planning zone, and Part 5(b) which reflects the two-5 mile EPZ, and it also includes the evacuation time 6
estimate, or ETE.
7 The third review area was the 25 exemption 8
requests that they provided. These include the two 9
exemption requests which are applicable to both the 10 site boundary and the two-mile plume exposure pathway 11 emergency planning zone. And the remaining 23 12 exemption requests address portions of 10 C.F.R. 5047 13 (b), and Appendix E for offsite emergency planning 14 related to the site boundary EPZ only.
15 Next slide, please. With regard to the 25 16
--- make sure I have the right slide --- with regard 17 to the 25 exemption requests, the two exemption 18 requests from 10 C.F.R. 50.33(g) and 50.47© would 19 remove the ten-mile plume exposure pathway EPZ 20 requirement. That same requirement is in both of 21 those regulations.
22 The remaining 23 exemption requests, which 23 are from 10 C.F.R. 50.47 and Appendix E to Part 50, 24 would remove emergency planning requirements 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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75 associated with offsite emergency planning. These 1
requirements are associated with state and local 2
emergency plans, public alert and notification, 3
evacuation time estimate, and offsite exercises.
4 Next slide, please. This slide provides 5
the basis for the staff's acceptance of the requested 6
exemptions. The ESP application provides a basis for 7
the establishment in the COLA of either a site 8
boundary or two-mile plume exposure pathway emergency 9
planning zone, and this is important, which maintains 10 the same level of protection, that is dose savings in 11 the event of a radiological emergency in the environs 12 of the Clinch River site, as that which exists in the 13 basis for a ten-mile plume exposure pathway EPZ, 14 similar to what we used for the large light water 15 reactors.
16 Next slide. This slide addresses the 17 combined license application, or COLA. Upon issuance 18 of the ESP the Applicant, TVA, acquires approval that 19 is finality with conditions of the three key review 20 areas that I just spoke of, first of all, the plume 21 exposure pathway EPZ sizing methodology, the two major 22 features emergency plan, the site boundary or the two-23 mile PEP EPZ, and the 25 requested exemption requests.
24 A COLA that incorporates, by reference, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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76 the early site permit must identify the chosen SMR 1
technology for the Clinch River Nuclear site and 2
demonstrate that the EPZ sizing methodology supports 3
either the site boundary or the two-mile plume 4
exposure pathway emergency planning zone. The COLA 5
must also provide a complete and integrated emergency 6
plan.
7 For the two-mile plume exposure pathway 8
EPZ, the COLA must provide both onsite and offsite 9
emergency plans. For the site boundary plume exposure 10 pathway EPZ, the COLA must provide an onsite emergency 11 plan. And the COLA must also address all 16 of the 12 COL action items and the four permit conditions.
13 Those are 16 action items and four permit conditions 14 associated with emergency planning.
15 Next slide, please. This slide addresses 16 the EPZ size determination in the COLA. The 17 determination of the EPZ size by the COL Applicant is 18 required by two parts, two things, the COL action 19 item, 13.3-1, and this particular action item reflects 20 the language that was in the application Part 2 in 21 Section 1333-14.
22 The COLA must identify the chosen SMR 23 technology and the major features emergency plan, 24 that'll be in the two-miles of the site boundary. It 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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77 must provide detailed information that shows the 1
ability of the small modular reactor to meet the 2
chosen EPZ. And that would be utilized in the 3
methodology. And the selected SMR technology must be 4
the EPA early phase protective action guides.
5 Michelle Hart will address Permit 6
Condition 1.
7 MS. HART: Hello again, my name is 8
Michelle Hart. I'm a senior reactor engineer in the 9
Office of New Reactors, the Radiation Protection and 10 Accident Consequences Branch.
11 So for Permit Condition 1, this is related 12 to, with the exemptions approved for the ESP, the COL 13 Applicant can adopt the exemptions if it shows that 14 the plume exposure pathway EPZ source term releases to 15 the atmosphere are bounded by those in the non-design 16 specific plant parameter source term information 17 developed for the ESP. That's that table that's 18 attached to Permit Condition 1, that's 13.3-1.
19 And as stated on the slide, the permit 20 condition is that the Applicant would provide detailed 21 information to demonstrate that the accident release 22 source term information for the plume exposure pathway 23 EPZ size determination analysis, using the selected 24 SMR design, is bounded by the non-design specific 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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78 plant parameter source term information used in the 1
analysis supporting the exemption requests.
2 And that analysis would be done in 3
accordance with COL Action Item 13.3-1 using the 4
methodology in the SSAR, Chapter 13.3.
5 To go to your question, Dr. Kirchner, 6
about what would happen if one of --- let's just say 7
one of the isotopes is not less than the rest of the 8
--- or the isotope in that table. My understanding 9
is, because of the ministerial nature of the permit 10 condition, if they cannot show that they are within 11 that condition specifically, they may ask for an 12 exemption, but they do not --- or a variance, but they 13 do not automatically get to use the exemption requests 14 that were approved in the ESP based on the condition 15 with that design
- envelope, that source term 16 information.
17 However, they may still be able to prove, 18 through the use of the methodology, that although the 19 source term is slightly different, or it may slightly 20 exceed, that they still can prove that they have a 21 site boundary or a two-mile emergency planning zone 22 size according to the methodology.
23 MEMBER KIRCHNER: You've hit my question.
24 Because it struck me, reviewing all the material, that 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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79 it's almost --- you have to have agreement on the 1
source term, of course. But the real thing you're 2
regulating against is not the composition of the 3
source term, it's the dose to the public.
4 MS. HART: That's correct. And there's 5
some ---
6 MEMBER KIRCHNER: I had just worried that 7
you might have an over-defined boundary value problem 8
where ---
9 MS. HART: Right. In the subcommittee 10 meeting, we did have a more full discussion of how 11 they developed that source term. And I can discuss 12 that again a little bit later if you would like. But 13 they did add in a lot of uncertainty or a lot of 14 margin to try to address that concern.
15 Next slide, please. So as TVA had told 16 you earlier today that they ---
17 MEMBER REMPE: Michelle, can you ---
18 MS. HART: I'm sorry.
19 MEMBER REMPE: -- go back. I think I 20 brought this up at the subcommittee meeting, but I 21 can't remember how it was addressed. What if one of 22 these designs happens to have a burp immediately after 23 an event? And then something comes out starting on 24 three and a half days, and it keeps going along. So 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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80 what will you do if you see that kind of analysis? Or 1
do they just get to stop after four days, and they 2
don't have to keep it going?
3 MS. HART: Right.
4 MEMBER REMPE: And I've forgotten what 5
your response was.
6 MS. HART: Right. Well, how I answered 7
that at the subcommittee phase, and this is what I 8
still
- believe, is that that's part of the 9
implementation. And when we review their actual 10 implementation, we will be looking at all the 11 information that they have. And so if there is an 12 issue there, it can be addressed.
13 What the permit condition non-design 14 specific source term information is, is a 96-hour 15 integrated. And so their release is longer than that, 16 you know, we'll have to look at that when it comes in 17 if there's --
18 MEMBER REMPE: And did I ---
19 MS. HART: -- some problem there.
20 MEMBER REMPE: -- mention that somewhere 21 in whatever you --- again, I wasn't on the 22 subcommittee itself, I just happened to be at that one 23 meeting. And is that in your documentation somewhere, 24 that you aren't allowed to just stop it at 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />?
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81 You need to look for some sort of reduction or 1
truncation of releases.
2 MS. HART: It is not specifically 3
addressed. It's just the permit condition is written 4
in such a manner, and we will have to say if your 96-5 hour integrated release does not meet that, that it 6
would not meet the requirement to do the exemption.
7 MEMBER REMPE: Well, I can trust that 8
you'll --- this will be adhered to even if you go on 9
and get promoted to be a manager at a high level, that 10 the staff will know to do that without any ---
11 MS. HART: That should be true, correct.
12 MEMBER REMPE: Cool, thank you.
13 MS. HART: Okay, Slide 22. So as TVA had 14 mentioned earlier, they did have some technical 15 criteria for developing their EPZ size methodology, 16 that the plume exposure pathway EPZ should encompass 17 those areas in which projected dose from design-basis 18 accidents could exceed the EPA early phase PAGs.
19 The plume exposure pathway should also 20 encompass those areas in which consequences of less 21 severe core melt accidents could exceed the EPA early 22 phase PAG, and that the plume exposure pathway EPZ 23 should be of sufficient size to provide for 24 substantial reduction in early health effects in the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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82 event of a more severe core melt accident.
1 Next slide, please. TVA did go through 2
this earlier. I guess I probably don't need to repeat 3
it in detail. But certainly the features of the EPZ 4
size methodology are that they will select their 5
accident scenarios, and that would include design-6 basis accidents, just taking that directly from the 7
siting analysis that they do in Chapter 15.
8 And then you look at the severe accidents 9
using the COLA site and design specific probabilistic 10 risk assessment, should include all modes, internal 11 and external events, applicable fuel handling, and 12 spent fuel pool accidents, and also consider multi-13 module accident considerations.
14 And then you would categorize that in the 15 two different categories, the more probable less 16 severe core melt accidents with intact containment and 17 then less probable, more severe, core melt accidents 18 with either containment bypass or containment failure.
19 Once you categorize those accidents, you 20 would determine the source term releases to atmosphere 21 and its --- there's not a specific discussion as to 22 whether you can do bounding or should do all of them.
23 They can choose at that time. It's an implementation 24 thing we would also evaluate.
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83 So the source terms, there may be several 1
scenarios in a different category, and they may 2
determine to look at them all or they may categorize 3
them and get us the bounding. When you calculate the 4
dose consequences at distance from the plant, and then 5
you compare those doses to the dose base criteria.
6 Next slide, please. So to go in a little 7
bit more detail about the TVA dose-based plume 8
exposure pathway EPZ size criteria, the quantity that 9
we're looking at is the dose to an individual from 10 exposure to the airborne plume during its passage and 11 to groundshine using average atmospheric dispersion 12 characteristics for the site.
13 And what we mean by average atmospheric 14 dispersion characteristics for the site is not 15 referring to the same analysis that was done in SSAR 16 Chapter 2
and approved for the ESPS site 17 characteristics.
- Instead, it's referring to 18 evaluating the accident consequences using site-19 specific meteorological data to determine doses that 20 are based on 50th percentile atmospheric dispersion 21 factors.
22 And the staff expects that the Applicant 23 may use the calculation tools that are used for a 24 severe accident consequence analysis. For example, in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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84 the environmental report there's no specific tool 1
identified in TVA's methodology.
2 But for example, the tool that is mostly 3
used is the MACCS code, and so it can take a year's 4
worth of hourly meteorological data. And you can run 5
--- it can account for uncertainty in weather, 6
including over the duration of the accident release.
7 It models atmospheric transport and 8
dispersion by sampling one year of hourly weather data 9
for the site, and it can model shifts in wind 10 direction. It uses a Gaussian plume segment model, 11 and so each plume segment, the start time and duration 12 is chosen by the user. So it can be adjusted to the 13 shape of the accident release, if that makes some 14 sense.
15 We'll head in the wind direction and 16 speed, as sampled from the site-specific data, and the 17 start time of that sampling is random over the year.
18 So therefore, two plume segments released at adjacent 19 times may be traveling in different directions at 20 different speeds the way that MACCS does the modeling.
21 In practice, when we're saying that they 22 would look at the 50th percentile, or the mean doses, 23 excuse me, in practice the analysis would run several 24 weather trials with the same release source term for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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85 each weather trial but differing atmospheric 1
dispersion and transport based on the sampling of the 2
year's worth of data. And the resulting mean dose of 3
our weather trials would be taken as the output.
4 Yes?
5 MEMBER DIMITRIJEVIC: I don't have a 6
question, I just have a little correction, not for 7
just slide, but there was a slide there, the airplane 8
crashes where you have probability with the reactor.
9 Every time when you have a pattern, it's not 10 probability. Probability doesn't have a unit that's 11 frequent. And you should change that throughout, 12 because of the issue.
13 MS. HART: Thank you.
14 MEMBER REMPE: And if you agree with that 15 statement, and hopefully, you'll help the TVA folks 16 come to that conclusion too --
17 MS. HART: Yeah. And I think I understand 18 what you're saying. And it's not something that I 19 brought up with them before. So hopefully, we'll see 20 what happens.
21 Okay, so for the rest of this slide, it 22 reiterates the actual criteria that they have 23 proposed, is that for design-basis accidents and more 24 probable less severe accidents, those are the ones 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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86 with intact containment, that the dose criterion is 1
one rem, total effect of dose equivalent from a 96-2 hour exposure. And that is the lower end of the dose 3
range of the EPA PAG for early phase protective action 4
such as evacuation and sheltering.
5 And for the less probable, more severe 6
accidents, and you see that I have repeated it, but 7
they would calculate the distance at which the 8
conditional probability to exceed 200 rem whole body 9
from a 24-hour exposure exceeds ten to the minus 10 three. And they did say per reactor year. The 200 11 rem is, of course, the acute dose at which radiation 12 induced early health effects may begin to be noted.
13 And so I've heard ---
14 MEMBER KIRCHNER: Once more, we belabored 15 this earlier.
16 MS. HART: Yeah.
17 MEMBER KIRCHNER: But just so we're on the 18 same page, this is an integral effect, this ten to the 19 minus three?
20 MS. HART: I have to admit that this did 21 not get practiced in the example calculation, because 22 there was nothing that screened into that category.
23 In general, what we are seeing from some of these 24 small modular reactors, there's not very many 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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87 accidents that may be in that category, if any at all.
1 So I don't know that you would have more 2
than one source term affecting that evaluation. There 3
may be, depending on the design. I think it's mostly 4
going to be an effect of the weather.
5 And one of the things that we can do with 6
this is up to the implementation phase, it's not 7
discussed in their methodology or discussed in our 8
evaluation. But in implementation, you know, MACCS 9
runs one year at a time. But you can do more than one 10 year by running another set of MACCS analyses.
11 And so if there's some concern or 12 question, if you're not able to tell from the pre-13 processing of the weather, you know, to determine if 14 you've got a bad year, or a worse year or, you know, 15 from that perspective, if there's some need to have to 16 do more than one year's worth of MACCS runs, then that 17 is something that can be done. It would be evaluated 18 based on the information that we have at the time of 19 the implementation at the COL though.
20 MEMBER BLEY: Let me jump in here, Walt.
21 I've been trying to catch up a little. But this deal 22 about the ten to minus three, if you go back to 0396, 23 and you go back to the figure, and Walt asked me about 24 this last night, Figure 1-11, there's a curve for 200 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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88 rem. And what that curve says is, right at the site, 1
only eight percent of core melts can get you 200 rem, 2
even right at the site.
3 And by the time you get out to 20 miles, 4
and these are results from WASH-1400 that got adapted 5
for this report, when you get out to 20 miles, the 6
curve's dropping off so fast that you hit only one in 7
1,000 core melts can have an effect on you. 0396 8
talked about, for severe core melt accidents, you 9
ought to have a substantial reduction in health 10 effects.
11 MS. HART: Right.
12 MEMBER BLEY: And nowhere does it say that 13 substantial drop is ten to the minus three. But 14 that's kind of what everybody is doing. And it's 15 based on that one curve and then applying it to new 16 reactors as well. I thought that worth throwing in.
17 MS.
HART:
Are there any further 18 questions, concerns about that?
19 (No audible response.)
20 MS. HART: Okay. So next slide please.
21 So the staff's review of TVA's proposed plume exposure 22 pathway ETZ size methodology, we did compare the 23 methodology and the dose criteria to the study used as 24 the technical basis for the current regulatory 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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89 requirement for a ten mile plume exposure pathway EPZ 1
requirement, that is as we've been discussing NUREG 2
0396.
3 And the staff has determined that the 4
features of TVA's methodology are consistent with the 5
study that was done in NUREG 0396 in that it 6
considered a range of accidents. It performs an 7
accident consequence analysis and determines an area 8
outside of which early protective actions are not 9
likely to be necessary to protect the public from 10 radiological releases.
11 And so therefore, the staff concludes that 12 the Applicant's proposed methodology is reasonable and 13 consistent with the analyses that form the technical 14 basis for the current regulatory requirements of a 15 plume exposure pathway EPZ of about ten miles in 16 radius.
17 Next slide, please.
18 MEMBER BLEY: Michelle?
19 MS. HART: Yes?
20 MEMBER BLEY: For several reasons, I've 21 been going through 0396 in great detail recently.
22 This is one of them. Some 50 years ago, all the 23 quantitative judgements in it were based on Wash 1400 24 which was, at that point, three or four years old.
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90 Has anybody on the staff revisited 0396 1
and thought about it in light of what's been learned 2
in the last 50 years?
3 MS. HART: I can let somebody from the 4
Office of Nuclear Incident Response respond to that if 5
they would like.
6 MEMBER BLEY: They must have run out the 7
door.
8 MS. HART: Yeah. There're some folks 9
here. I mean, certainly, we are going through the 10 rulemaking for the emergency preparedness and for SMRs 11 and other new technologies.
12 MEMBER BLEY: Still point at 0396. The 13 logic there is great.
14 MS. HART: The logic is what we're using.
15 Now, if you're asking have we re-evaluated it in the 16 context of the currently operating reactors, I can't 17 necessarily speak to that. And I don't know that 18 that's what you're asking.
19 MEMBER BLEY: I think we're using more 20 than the logic. I think we're using some of the 21 quantitative information as well.
22 MS. HART: Well, I think certainly 23 continuing to use the EPA PAGs for the early phase as 24 the basis for how you determine EPZ size, we're still 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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91 sticking with that idea.
1 MEMBER BLEY: Yeah, but we're picking the 2
chance of what the dose is at some distance from very 3
old information.
4 MS. HART: In TVA's methodology, yes, they 5
did.
6 MEMBER BLEY: I didn't see anything in the 7
rulemaking. I mean, there would be a change in that.
8 MS. HART: The rulemaking, as I recall, 9
does not have that specific evaluation in the rule 10 language itself.
11 MEMBER BLEY: That's true.
12 MS.
HART:
About the very severe 13 accidents.
14 MR. SCOTT: I figured it out, thank you, 15 with help. This is Mike Scott of the NCR staff.
16 Talking to my colleagues here, we're not aware --- the 17 question is is there a current effort ongoing to 18 update 0396. Our answer is we're not aware of one.
19 Was that the --
20 MEMBER BLEY: Thanks. I'm not either.
21 And it just struck me, you know, it might be worth 22 somebody doing that.
23 MR. SCOTT: It's an interesting question 24 that we'll consider. Thank you.
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92 MEMBER BLEY: Not that I'd hang it on this 1
particular application. But I think it's about time 2
we thought about it.
3 MS. HART: Okay. So then to Slide -- what 4
is this, 26 -- so for the exemption requests to 5
determine if --- to put a boundary around what we 6
considered when we were looking at it in the ESP, 7
since there is not a specific design included in this, 8
TVA developed a non-design specific accident release 9
source term that would meet the plume exposure pathway 10 EPZ size criteria which are intended to be used as 11 plant parameters for the purposes of the exemption 12 request.
13 This source term is in Table 13.3-1. It 14 is an isotopic total release activity over 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 15 which results in a Total Effective Dose Equivalent of 16 about 0.9 rem at the site boundary. It's the same 17 idea as the plant parameter envelope in general that's 18 done for the ESP, specifically for the design basis 19 accident source term. And it's intended to envelope 20 an unknown design. And it's referenced in Permit 21 Condition 1 for the adoption of the EP exemptions.
22 This non-design specific source term used 23 information from two different designs from three 24 accidents, two DBAs, and one severe accident. The two 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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93 SMRs were at the lower end of the range of the PPE and 1
at the upper end of the range of the PPE as far as 2
reactor thermal power.
3 And when they did this, they took the 4
maximum activity that could be released in any time 5
period from any of those three accidents from the two 6
reactors. They added a 25 percent margin, and when 7
they tried to back calculate from the 1 rem criterion, 8
there was also some additional adjustment to some of 9
the isotopic values. And then they calculated the 10 final source term to result in some margin to the dose 11 criterion, so about 0.9 rem at the site boundary.
12 And so it's the plant condition, plant 13 parameters for the condition to use it for either the 14 site boundary or the two-mile emergency planning zone.
15 There's not a separate table for those two different 16 distances.
17 And that concludes my portion of the 18 presentation. I will turn it back over to Mallecia.
19 MS. SUTTON: The staff presented its 20 review on findings on emergency planning for TVA 21 Clinch River early site permit application. The staff 22 concludes that the PEP EPZ size methodology is 23 acceptable for determining the appropriate size of the 24 PEP EPZ for the Clinch River nuclear site. Because 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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94 it's consistent with analysis that formed the clinical 1
basis for the current ten-mile PEP EPZ.
2 The two major features in emergency plans 3
are acceptable, because they meet the applicable 4
standards of 10 C.F.R. 5047 and requirements of 5
Appendix E to 10 C.F.R. Part 50. If the exemptions 6
are approved for the ESP, the Applicant can adopt 7
exemptions if it shows that its COLA PEP EPZ source 8
term release to the atmosphere is bounded by the non-9 design specific plant parameters source term 10 information developed for the ESP.
11 The exemption requests are acceptable, 12 because they are authorized by law, will not present 13 an undue risk to the public health and safety, are 14 consistent with the common defense and security, and 15 special circumstances are present.
16 In previous subcommittee meetings, we have 17 presented the staff's review and findings relative to 18 this application for an early site permit at the 19 Clinch River nuclear site. Today we presented an 20 overview including more details in the emergency 21 planning and related exemption request. The safety 22 evaluation is complete with no open items.
23 The next step in the process is the 24 mandatory hearing in front of the Commission in 2019.
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95 The staff looks forward to an ACRS letter on the staff 1
review. And this completes the staff presentation.
2 MEMBER KIRCHNER: Thank you, Mallecia.
3 MEMBER BLEY: I want to make just a ---
4 MEMBER KIRCHNER: Yes?
5 MEMBER BLEY: -- very minor comment which 6
could --
7 MEMBER KIRCHNER: Go ahead, Dennis.
8 MEMBER BLEY: -- be editorial. In the 9
licensee's report, Chapter 13, they go through the 10 steps and the methodology. And they do that well, and 11 they say find these scenarios, then group the 12 scenarios by the kind of things that failed and what 13 the consequences are. The next step should say for 14 the groups, scenario groups, find the frequency. And 15 it doesn't. It just says for the scenario. Just a 16 comment for you.
17 MEMBER KIRCHNER: Other members, any 18 questions of the staff while they're here in front of 19 us? Then if not, we'll turn to the public.
20 (No audible response.)
21 MEMBER KIRCHNER: Okay, thank you again.
22 Are there any members of the public in the room who 23 wish to make a statement or a concern? Please step 24 forward, identify yourself at the mic, and make your 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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96 comment.
1 Seeing no one coming forward, is there 2
anyone, member of the public, on our bridge line who 3
wishes to make a comment? If so, state your name, 4
please, and make your comment.
5 (No audible response.)
6 MEMBER KIRCHNER: Hearing none, at this 7
point, Mr. Chairman, I'll turn it over to you.
8 CHAIRMAN CORRADINI: Thank you. So I'll 9
thank members of TVA and the staff. And I think we're 10 done with this subject. So we're going to take a 11 short break, so we change out and talk about Seabrook 12 next. So we'll be coming back at 3:15.
13 MEMBER BLEY: We're ahead of time. We 14 can't start that until the scheduled time.
15 (Off the record comments.)
16 MEMBER BLEY: 2:30 or 2:45. I don't have 17 my glasses.
18 (Off the record comments.)
19 MEMBER BLEY: That's 45 minutes, not 50.
20 (Off the record comments.)
21 (Laughter.)
22 CHAIRMAN CORRADINI: So once again, we'll 23 see you in 15 minutes. Thank you all.
24 (Whereupon, the above-entitled matter went 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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97 off the record at 2:59 p.m. and resumed at 3:13 p.m.)
1 CHAIRMAN CORRADINI: Okay, why don't we 2
come back into session. Our next topic is going to be 3
Seabrook, Unit 1, license renewal application. And 4
I'll turn it over to Member Skillman.
5 MEMBER SKILLMAN: Yes sir, thank you, 6
Mike. Ladies and gentlemen, this meeting this 7
afternoon brings us to a very important time in 8
Seabrook's life. We have been involved in license 9
renewal of Seabrook since our meeting in 2012. It has 10 been over six years. And in intervening time, from an 11 original application, and then updates to the 12 application and the safety evaluation, through years 13 of work on Alkali-Silica Reaction, we come to today.
14 And so through the presentation and letter 15 writing we will address both the license renewal 16 application and Alkali-Silica Reaction. And with that 17 opening comment, I will turn it over Joe Donoghue, 18 please.
19 MR. DONOGHUE: Okay, good afternoon.
20 Thank you, Chairman Corradini, and Mr. Skillman, and 21 members of the ACRS full committee. I'm Joe Donoghue, 22 I'm the deputy director of the Division of Materials 23 and License Renewal in NRR.
24 We thank you for the opportunity given us 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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98 to present the results of the staff's review of the 1
license for an application for Seabrook Station, Unit 2
- 1. This review began many years ago, and as Mr.
3 Skillman alluded to, one of the main technical issues 4
that prolonged the review was the Alkali-Silica 5
Reaction affecting concrete structures and then the 6
licensee's development of methods. And I'll review 7
those methods for managing the phenomenon.
8 On October 31st, the License Renewal 9
Subcommittee of the ACRS heard detailed presentations 10 from both the Applicant and the staff on ASR and the 11 basis for closing out that one open item of the 12 license renewal. On November 15th, the subcommittee 13 heard from the Applicant and the staff on the closeout 14 of the remaining open items in the SER.
15 Our presentation will be led by our 16 project manager, Butch Burton, and other members of 17 the staff and the management that are here, Dr. Allen 18 Hiser, our senior technical advisor, Eric Oesterle, 19 chief of the project's branch in our division, and 20 there's other managers and other technical staff that 21 contributed to the review that are present and that 22 will support answering any questions you have.
23 We also have, I think, maybe on the phone, 24 Region I staff who will provide inspection support and 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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99 provided presentations to you during the subcommittee 1
meetings.
2 Again, we look forward to answering any 3
questions you have and having a full discussion. And 4
I'll turn the presentation over at this point to the 5
NextEra team and their regional vice president from 6
the northern region, Mr. Eric McCartney.
7 MR. McCARTNEY: Thank you, Mr. Donoghue.
8 Good afternoon. My name is Eric McCartney. I'm the 9
regional vice president for NextEra Energy with 10 responsibility for the Seabrook Station, Point Beach 11 Station in Wisconsin, and the Duane Arnold Station in 12 Iowa.
13 Today we're here to talk about the 14 Seabrook Station. We appreciate the opportunity to 15 come and provide our presentation of our license 16 renewal application and all the work we've done over 17 the last six years, as Mr. Skillman mentioned. And we 18 look forward to a good discussion and answering any 19 questions that the Committee may have about our 20 program and our process.
21 We are committed to the safe, and 22 reliable, and sustained operation of our nuclear 23 fleet. And as we do that --- if you'll turn the 24 slide, please --- there we go. This is our nuclear 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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100 excellence model. And this provides the framework for 1
how our fleet has operated since 2008. It's based on 2
a set of core values and principles, and those have 3
not changed since its inception, and they will not 4
change.
5 And we use this as a road map of how we 6
operate our fleet going forward. So I won't go 7
through this as we've discussed this a number of times 8
already. But this continues to be at the heart of how 9
we manage our stations and our leadership model to 10 drive safe, reliable, and sustainable operations of 11 our fleet.
12 Today I have with me Mr. Mike Collins.
13 He's our engineering director. Next to him is Mr. Ed 14 Carley. Ed Carley's our license renewal supervisor.
15 And next to Ed is Ken Browne. Ken Browne is our 16 licensing manager. And then seated over here to my 17 right is Rudy Gil. And Rudy Gil is our engineering 18 program manager. And they will provide the technical 19 responses to your questions today.
20 And with that, I'll turn the presentation 21 over to Mr. Collins.
22 MR. BROWNE: Thank you, Eric. Good 23 afternoon, Mr. Chairman. I'm Ken Browne, licensing 24 manager for NextEra Seabrook. I've been at Seabrook 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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101 for approximately 28
- years, beginning in the 1
Operations Department as a non-licensed operator, then 2
as a licensed senior reactor operator working and 3
controlling various positions, including shift manager 4
and eventually director of operations.
5 I've also held the position of training 6
manager of accredited programs and most recently as 7
the licensing regulatory compliance manager and also 8
the management sponsor for the Alkali-Silica Reaction 9
project at Seabrook.
10 As we discussed at our ACRS Subcommittee 11 meeting last month, this station has continued to 12 engage in accumulating the best practices from the 13 industry in developing our existing engineering 14 programs as well as enhancing our aging management 15 plans to ensure Seabrook is maintained to the highest 16 safety and material standards.
17 Since we've been operating, NextEra 18 Seabrook has always made it our highest priority to 19 operate our facility with nuclear and public safety as 20 the overriding focus in all that we do. Each of us 21 that work there and live near the area recognize the 22 location of our facility places a
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102 surrounding Seabrook.
1 We also recognize the valuable resource 2
that Seabrook represents and continues to provide for 3
many years as a major proportion of safe, reliable, 4
and clean energy in the New England area. We look 5
forward to the Committee's questions. And I'm going 6
to turn the panel over to Mike Collins, our 7
engineering director, to guide us through the 8
presentation, including some background on the 9
station. Mike?
10 MR. COLLINS: Good afternoon. Again, my 11 name is Mike Collins, Director of Engineering at 12 Seabrook Station, 37 years in the industry, 17 years 13 with Stone and Webster Engineering, with new build and 14 continuing services, the last 20 years with NextEra 15 Energy, Seabrook Station, five of which as engineering 16 director.
17 (No audible response.)
18 MR. COLLINS: So our agenda for this 19 afternoon, again, our introductions, I'll provide an 20 overview of site and station description. Ed Carley 21 will then review our license renewal application and 22 our Aging Management Programs, review the safety 23 evaluation report and closure of the previous open 24 items. There'll be then closing remarks. And in 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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103 summary, we'll end with NextEra Seabrook has met the 1
requirements of 10 C.F.R. 54 for issuance of a renewed 2
license for Seabrook Station, Unit 1.
3 Just so we won't bore the group, I've 4
changed up the slide from previous. This is a picture 5
of the station and some of the main structures, our 6
intake, excuse me, discharge and intake structure, a 7
circ water and service water pumphouse, certainly our 8
containment enclosure building where the reactor 9
building is housed within, our Unit 1 turbine 10 building, fuel storage building, waste process 11 building. And this area of the plant is our primary 12 auxiliary building, our control building which houses 13 our two emergency diesel generators.
14 As you know, the Atlantic Ocean is the 15 normal heatsink for cooling at 100 percent power. We 16 also have a standby cooling tower which is a seismic 17 Cat 1 mechanical draft cooling tower which provides 18 additional safe shutdown capability for the station.
19 Next slide, please. Plant status, 20 recently completed our latest refueling outage, 21 fueling outage OR-19, which we completed 10/29/18.
22 Our next refueling outage at the end of Cycle 20 is 23 spring 2020 in the April timeframe.
24 Our capacity factor for 15 of 19 cycles 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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104 has been greater than 94 percent with a lifetime 1
capacity factor of 87 percent. As you can see with 2
the listing of our cycle capacity factors, we've had 3
an excellent operating history over the last cycle.
4 Capacity factor performance is representative of our 5
solid equipment reliability and our material condition 6
for the station.
7 Next slide, please. In order to maintain 8
high capacity factors, Seabrook continues to improve 9
equipment reliability and material conditions of the 10 station. Running down just through some items, for 11 equipment reliability improvements, our main generator 12 stator rewind, in the process replacing our vital 13 batteries and our vital inverters, our generator step-14 up transformers replaced --- there's three of those 15 that we fully replaced two outages ago.
16 As part of our Aging Management Program, 17 our mechanical stress improvement process completed 18 for all reactor vessel nozzles. Also Aging 19 Management, we continue with our process of replacing 20 all our above-ground service water piping with the 21 high chroma AL6XN material. We've upgraded our incore 22 detectors and have been aggressive with replacing our 23 process control circuit cards and our solid state 24 system circuit cards.
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105 Two outages ago, we sent out our rod 1
control motor and generator sets for refurbishment.
2 And lastly, for all four reactor coolant pumps, we now 3
have shutdown reactor coolant pump seals.
4 We are committed, NextEra Energy Seabrook, 5
to maintain high levels of safety, reliability, and 6
performance of our plant equipment.
7 DR. SCHULTZ: Mike, excuse me. You 8
mentioned two of the items on the list that you 9
attributed to the Aging Management Programs. And then 10 you stopped listing what the remainder were for. You 11 mentioned reliability and Aging Management halfway 12 down the list. Is that the full characterization of 13 why you made these changes?
14 MR. COLLINS: Yes, it is. With the ones 15 that I didn't mention, Aging Management, those are 16 driven by system engineer advocacy, trends of 17 equipment such as the GSU. We're watching very 18 closely the offgassing of the old generator step-up 19 transformers. Those go in for our long term plant 20 reliability plans. And we put them through the 21 process, do the engineering, do the maintenance, and 22 do the replacements either online or during the 23 outages.
24 DR. SCHULTZ: So that's how you separate 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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106 them from what you would term an Aging Management 1
Program improvement?
2 MR. COLLINS: That's correct.
3 DR. SCHULTZ: Okay, thank you.
4 MR. COLLINS: Thank you. At this time, 5
I'll turn the program over to Ed Carley to start our 6
discussions on our license renewal application.
7 MR. CARLEY: Good afternoon. I'm Ed 8
Carley. I am a 35-year veteran of Seabrook Station 9
and been in various organizations, quality assurance, 10 licensing, engineering projects. In 2008 I joined 11 the team developing the license renewal application as 12 the time limit aging analysis lead and the 13 environmental lead.
And shortly
- after, the 14 application was submitted in 2010. I also took on the 15 role as licensing lead.
16 For the last four years, I've been the 17 project manager for the license renewal application 18 and resolution of ASR and the current licensing basis 19 for concrete affected structures.
20 The original license renewal application 21 was prepared onsite by Seabrook Station personnel.
22 The team was supplemented by staff and contractors 23 with various experience in license renewal and those 24 that were former plant employees that were familiar 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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107 with the history of the plant. Our corporate fleet 1
also provided experienced personnel in license renewal 2
and provided oversight to the project.
3 Our application was prepared in accordance 4
with the standard review plan that's listed up here, 5
followed the standard format for an application. We 6
filed the guidance of NEI 95-10. And we developed our 7
Aging Managing Programs in accordance with NUREG-1801, 8
commonly referred to as GALL. Our initial application 9
was submitted as GALL Rev 1.
10 Since that time of submittal, we have 11 performed over 65 updates, some of those were 12 proactive, some were related to REIs, and also 13 produced eight annual updates to keep the application 14 current.
15 We've addressed all ISGs that have been 16 issued since the initial application was submitted.
17 And we have performed a consistent review to GALL Rev 18 1 and GALL Rev 2, and provided updates to our program 19 where we felt necessary to come in compliance with 20 GALL Rev 2 for those programs.
21 Next slide, please. This is a table of 22 our relationship in the final SER to the GALL.
23 Fifteen of our programs we consider new. We consider 24 29 that were existing. We do have six plant-specific 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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108 programs to which we have discussed in very much 1
detail being the ASR and Building Deformation 2
Programs.
3 Next slide, please. In relation to the 4
Safety Evaluation Report that was issued by the staff 5
on September 28th, it documented no open items and no 6
confirmatory items. There were seven open items in 7
the previous SER in 2012 as discussed earlier which 8
are listed here. The first six we did discuss on 9
November 15th. Of those, the programs for treated 10 borated water, operating experience, and part of the 11 Steam Generator Tube Integrity Program were resolved 12 by adoption and incorporation of the ISG guidance that 13 was applicable to those programs.
14 The other portion of the steam generator 15 tube integrity and the pressure temperature limit open 16 item were addressed by licensing actions in Part 50 17 for license amendments that changed our operating 18 license to resolve the open item. Pressure-19 temperature limits have been approved out to 55 20 effective full power years which will take us through 21 the period of extended operation.
22 Of the remaining open items that we 23 haven't discussed, Bolting Integrity Program was 24 related to a seal cap enclosure that was placed on a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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109 safety injector valve that was leaking during one of 1
our operating cycles. We have, following cycle 2
outage, we removed the seal cap enclosure, replaced 3
the valve to eliminate the leakage of that valve.
4 In relationship to the IWE program, this 5
was in relationship to the water that had accumulated 6
in our annulus area. And there was a concern that we 7
may have had degradation against our liner. To 8
resolve that issue, we have established a weekly PM 9
that verifies that area is in a dewatered state.
10 We have performed UT measurements around 11 the liner at the area of the moisture barrier, 12 confirmed there is no degradation of the liner in 13 those areas. And we also have a commitment to perform 14 that UT every five years, excuse me, every five 15 cycles.
16 And the last item, which is the Structures 17 Monitoring Program, we discussed quite extensively on 18 October 30th -- can I have the next slide there -- and 19 this is related to our Structures Monitoring Program.
20 Structures Monitoring Program was developed in 21 accordance with the GALL. However, because of ASR and 22 building deformation, it is now augmented by 23 supplemental plant-specific Aging Management Programs, 24 one for ASR and one for building deformation.
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110 As we discussed use of this flow chart 1
earlier, the structural capacity that came out of the 2
large scale test program at the University of Texas, 3
dose limits have been incorporated, that keep us bound 4
by that testing program, have been incorporated in the 5
Structures Monitoring Program.
6 And also, the structural demand portion 7
where we have performed -- in our performing analysis 8
of our seismic category Cat 1 structures, those 9
parameters to maintain us within the bounds of those 10 evaluations area also incorporated into the Structures 11 Monitoring Program. Frequencies, limits, and trending 12 are performed in accordance with the Structures 13 Monitoring Program to verify that we will not exceed 14 the limits prior to reaching the next inspection 15 interval.
16 MEMBER REMPE: Excuse me.
17 MEMBER MARCH-LEUBA: Go ahead.
18 MEMBER REMPE: Just to make sure that we 19 have the facts correct, because we've seen some 20 different states, I believe, and so just confirm for 21 me, that you first detected visual indications of ASR 22 in year of 2009. Is that correct to your 23 understanding.
24 MR. CARLEY: That is correct, in the Bravo 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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111 electrical tunnel.
1 MEMBER REMPE: That's okay. I just wanted 2
to make sure. Thank you.
3 MR. CARLEY: You're welcome.
4 MEMBER MARCH-LEUBA: With respect this 5
slide, you said NextEra has implemented the two ASR 6
programs, ASR and building deformation. Is that 7
correct?
8 MR. CARLEY: That is correct.
9 MEMBER MARCH-LEUBA: And the moment the 10 licensee's amendment request gets signed, you'll be 11 caring for it. Right now, you're doing it on your 12 own. At the moment this licensee's amendment request 13 related to the ASR methodology, correct?
14 MR. CARLEY: Yes.
15 MEMBER MARCH-LEUBA: At this point, you're 16 doing it on your own. At the moment that LRA gets 17 signed, you will be able to take care for it.
18 MR. CARLEY: We'll be able to close our 19 PODs that are related to building deformation.
20 MEMBER MARCH-LEUBA: Okay. The real 21 question is on Unit 41, after you get the LRA, 22 anything will change, or everything will be solved 23 before that?
24 MR. CARLEY: Everything -- we do have a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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112 commitment. And I apologize, I do not remember the 1
commitment number. We have two structures that are 2
non-seismic Category 1
structures, intake and 3
discharge structures, that we have committed to 4
analyze before 2020 and will implement the program for 5
those structures when that analysis is done. So 6
that'll be --
7 MEMBER MARCH-LEUBA: So the only --
8 MR. CARLEY: -- a couple of years prior.
9 MEMBER MARCH-LEUBA: The only change that 10 will happen on Unit 41 will be those addition of two 11 additional non-Category 1 structures?
12 MR. CARLEY: Those will actually be 13 incorporated in 2020.
14 MEMBER MARCH-LEUBA: Before the LRA gets 15 issued.
16 MR. CARLEY: No, before the period of 17 extended operation.
18 MEMBER MARCH-LEUBA: Correct. That's the 19 appropriate terminology. Okay, thank you.
20 MEMBER SKILLMAN: Ed, anything else?
21 MR. CARLEY: With that, I'll turn it over 22 for concluding remarks.
23 MEMBER SKILLMAN: Okay.
24 MR. COLLINS: With regards to our 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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113 concluding
- remarks, as presented, Seabrook is 1
committed to the continuous improvement and long term 2
operation of NextEra Seabrook Station. Seabrook will 3
manage the effective agency in accordance with 10 4
C.F.R. 5421(a)(1). Seabrook has conducted time 5
limited aging analysis that require evaluation under 6
7 In summary, in closing, NextEra Energy 8
Seabrook has demonstrated compliance with the 9
requirement of 10 C.F.R. 54 for issuance of a renewed 10 license for Seabrook Station, Unit 1.
11 This concludes our presentation at this 12 time. I'll turn it over to Ken Browne.
13 MR. BROWNE: Now, as Mr. Collins noted, 14 Mr. Chairman, that concludes NextEra's presentation 15 for license renewal.
16 MEMBER SKILLMAN: Okay, Seabrook team, 17 anything else? No? Call-ins, any questions for the 18 Seabrook team before we change out to the NRC team?
19 (No audible response.)
20 MEMBER SKILLMAN: Seabrook team, thank 21 you. Please stay in the room. And we call out the 22 NRC team.
23 (Pause.)
24 MEMBER SKILLMAN: Thank you, Kendra.
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114 Butch? Take it away, please.
1 MR. BURTON: All right. Good afternoon.
2 Chairman Corradini, Chairman Skillman, and members of 3
the ACRS. My name is Butch Burton, and I am the 4
license renewal project manager for the Seabrook 5
Station, Unit 1 Safety Review.
6 We're here today to discuss the staff's 7
review of the Seabrook License Renewal Application 8
which we -- otherwise known as the LRA, as documented 9
in the safety evaluation report that was issued on 10 September 28, 2018.
11 Joining me here at the table today are Dr.
12 Allen Hiser, senior technical advisor in NRR's 13 Division of Materials and License Renewal, and Mr.
14 Eric Oesterle, branch chief of the projects branch in 15 the division.
16 Also seated in the audience and available 17 on the phone are members of the NRC technical staff 18 who participated in the review of the license renewal 19 application and conducted onsite audits and 20 inspections.
21 The presentation is short and sweet. I'll 22 begin the presentation with a general overview of the 23 staff's review. And since there are no open or 24 confirmatory items in the SCR, we'll then proceed to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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115 the staff's conclusions.
1 On May 25th, 2010, NextEra Energy Seabrook 2
submitted an application for renewal of the Seabrook 3
operating license for an additional 20 years or until 4
March 15th, 2050. For the review of the Seabrook 5
license renewal application, the following audits and 6
inspections were conducted onsite.
7 First, in September 2010, the staff 8
conducted an audit to review NextEra's administrative 9
controls governing the scoping and screening 10 methodology and the technical basis for the scoping 11 and screening results. The staff documented the 12 scoping and screening methodology audit results in a 13 report dated February 4th, 2011.
14 Second, during two weeks in October 2010, 15 the staff audited NextEra's Aging Management Programs, 16 which we call AMPs, and relayed a documentation to 17 verify NextEra's claim that the programs were 18 consistent with those described in the NRC's Generic 19 Aging Lessons Learned or GALL report and, considering 20 any enhancements or exceptions to the AMPs, whether 21 the programs were adequate to manage aging during the 22 period of extended operation.
23 The staff considered plant conditions and 24 operating experience during the audits and documented 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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116 the results in a report dated March 21st, 2011.
1 Third, during three weeks in March and 2
April 2011, Region I inspectors conducted a 71002 3
inspection in support of the review of the Seabrook 4
LRA and documented the results in a report dated May 5
23rd, 2011.
6 Fourth, during the last week of April 7
2018, Region I inspectors conducted a second 71002 8
inspection on Aging Management programs for concrete 9
structures affected by alkali silica reaction, known 10 as ASR. Region I documented the results of this 11 focused inspection in a report dated August 10th, 12 2018. And this issue was discussed with the ACRS 13 Subcommittee on Plant License Renewal at its October 14 31st meeting.
15 In June 2012, the staff issued a safety 16 evaluation report for the Seabrook LRA with seven open 17 items which are listed on this table. In September of 18 2018, the staff issued a second safety evaluation 19 report which resolved these seven open items.
20 Following issuance of the SER with open 21 items, the staff and NextEra met with the ACRS 22 Subcommittee on Plant License Renewal in July 2012 to 23 discuss the staff's findings. Of the seven open items 24 documented in the SER, the open item associated with 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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117 the structure's monitoring program, and how it manages 1
aging associated with ASR, dominated the discussions 2
between the ACRS Subcommittee, NextEra and the staff.
3 The resolution and closure of the seven 4
open items was documented in the staff's SER issued in 5
September of 2018. During the staff's in depth 6
technical review of the LRA over the last eight years, 7
including two audits and two inspections, a total of 8
291 RAIs were issued, 58 of which were follow-up RAIs.
9 Following issuance of the SER in September 10 2018, the ACRS Subcommittee on Plant License Renewal 11 held meetings with the NRC staff and NextEra, as I 12 mentioned, on October 31st and on November 15th, 2018.
13 The October 31st meeting was focused on 14 ASR at Seabrook including resolution of the open item 15 associated with the structure's monitoring program, 16 and how the aging effects on structures and components 17 affected by ASR would be managed during the period of 18 extended operation. The November 15th subcommittee 19 meeting focused on the closeout of the remaining open 20 items.
21 SER Section 2 describes the scoping of 22 systems, structures, and components, known as SSCs, 23 and screening of structures and components to identify 24 those subject to an aging management review, known as 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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118 an AMR. The staff reviewed NextEra's scoping and 1
screening methodology, procedures, quality controls 2
applicable to the development of the LRA, and training 3
of its project personnel.
4 The staff also reviewed the various 5
summaries of safety related SSCs, non-safety related 6
SSCs affecting safety functions, and SSCs relied upon 7
to perform functions applicable to Seabrook in 8
compliance with the Commission's regulations for fire 9
protection, environmental qualification, station 10 blackout, and anticipated transients without scram.
11 Based on its review, results from the 12 scoping and screening audit and additional information 13 provided by NextEra, the staff concludes that 14 NextEra's scoping and screening methodology and its 15 implementation were consistent with the standard 16 review plan for license renewal, known as the SRP, and 17 the requirements of 10 C.F.R. 54.4(a).
18 SER Chapter 3 and its subsections cover 19 the staff's review of NextEra's programs for managing 20 aging in accordance with 10 C.F.R. 5421(a)(3).
21 Sections 3.1 through 3.6 include the AMR items in each 22 of the general system areas within the scope of 23 license renewal. For a given AMR item, the staff 24 reviewed the item to determine whether it is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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119 consistent with the GALL report.
1 For the AMR items not consistent with the 2
GALL report, the staff reviewed NextEra's evaluation 3
to determine whether NextEra has demonstrated 4
reasonable assurance that the effects of aging will be 5
adequately managed so that the intended functions will 6
be maintained consistent with the current licensing 7
basis for the period of extended operation, as 8
required by 10 C.F.R. 5421(a)(3).
9 The license renewal application was 10 submitted in 2010 and described a total of 42 Aging 11 Management Programs, 13 of which were new and 29 of 12 which were existing. As a result of the staff's 13 review, two additional plant-specific Aging Management 14 Programs, the ASR Monitoring Program and the Building 15 Deformation Monitoring Program, were developed to 16 address the management of structures affected by ASR, 17 for a total of 44 Aging Management Programs.
18 All AMPs, with the exception of the plant-19 specific AMPs, were evaluated by the staff for 20 consistency with Revision 2 of the GALL report. For 21 the plant-specific AMPs, the staff evaluated them 22 against the program elements defined in Appendix A.1 23 of the SRP.
24 Section 4 of the SER identifies time 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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120 limited aging analyses, or TLAAs. Section 4.1 1
documents the staff's evaluation of NextEra's 2
identification of applicable TLAAs. The staff 3
evaluated NextEra's basis for identifying those plant-4 specific or generic analyses that need to be 5
identified as TLAAs and determined that NextEra has 6
provided an accurate list of TLAAs as required by 10 7
C.F.R. 5421(c)(1).
8 Section 4.2 through 4.7 document the 9
staff's review of the applicable TLAAs as shown.
10 Based on its review, and the information provided by 11 NextEra, the staff concludes that either the analyses 12 remain valid for the period of extended operation, or 13 the analyses have been projected to the end of the 14 period of extended operation, or the effects of aging 15 on the intended functions will be adequately managed 16 for the period of extended operations as required by 17 54(c)(1), Subparagraphs I, ii, and iii.
18 The staff's reviewed NextEra's responses 19 to the open items identified in the safety evaluation 20 report with open items that was issued in June 2012 21 and finds that all the open items have been 22 satisfactorily resolved and closed. With the closure 23 of the open items, the staff finds that NextEra has 24 met the requirements of 10 C.F.R. 5429(a) for the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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121 license renewal of Seabrook Station, Unit 1.
1 More specifically, the staff finds that 2
actions have been identified and have been or will be 3
taken at Seabrook Station, Unit 1 such that there is 4
reasonable assurance that the activities authorized by 5
the renewed license will continue to be conducted in 6
accordance with the current licensing basis and that 7
any changes made to the plant's current licensing 8
basis are in accordance with the Atomic Energy Act and 9
the Commission's regulations.
10 This concludes the staff's presentation, 11 and we'll be happy to take any remaining questions you 12 may have.
13 MEMBER SKILLMAN: Butch, thank you. Dr.
14 Hiser, Eric, thank you.
15 Colleagues, any questions for the NRC 16 team, please?
17 (No audible response.)
18 MEMBER SKILLMAN: If not, I would ask you 19 to stand by. Let's go to the public. Are there any 20 individuals in the room that would care to make a 21 comment? If so, I invite you to come to the 22 microphone and ---
23 (Telephonic interference.)
24 MEMBER SKILLMAN: -- I ask you to come to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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122 the microphone, and introduce yourself and speak 1
clearly into the microphone, please.
2 Seeing none, we go to the phone line.
3 Ladies and gentlemen on the phone line, if one or some 4
of you are out there, would you just please simply say 5
hello so that we know that you are there?
6 MR. OSSING: Hello?
7 MEMBER SKILLMAN: Thank you. All right.
8 For any individual on the phone line that would like 9
to make a comment, please introduce yourself and then 10 make your comment, please.
11 MR. OSSING: Hello, my name is Michael 12 Ossing from Marlborough, Massachusetts. I'd first 13 like to acknowledge the efforts by the NRC staff, and 14 the ACRS, as well as NextEra during this eight-year 15 process.
16 Seabrook is in compliance with the license 17 renewal and Aging Management Program position --- and 18 positioned, rather, for the station to operate safely 19 during the license renewal process. I would support 20 the ACRS providing a favorable recommendation to issue 21 Seabrook a license renewal for the period of extended 22 operation. Thank you.
23 MEMBER SKILLMAN: Thank you, sir. Is 24 there another individual out there that would like to 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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123 make a comment, please?
1 (No audible response.)
2 MEMBER SKILLMAN: Hearing none, please 3
close the phone line. And Chairman Corradini, back to 4
you.
5 CHAIRMAN CORRADINI: Well, thank you. I 6
was expecting there would be more public comments.
7 Okay, thank you very much to NextEra and the staff.
8 And we're going to go off the record, take a couple of 9
minutes to rearrange, and we will probably take up the 10 NextEra letters. And, Dick, you'll lead us through.
11 MEMBER SKILLMAN: Yes. Let me make one 12 comment. We are going to process two letters this 13 afternoon, we hope. One letter is on the license 14 renewal amendment that is plus 20 years. And the 15 second letter is devoted to Alkali-Silica Reaction.
16 And our desire is to process the ASR letter first and 17 then the license extension letter second. So that's 18 the plan going forward. And we're prepared. Thank 19 you.
20 CHAIRMAN CORRADINI: We'll take a few 21 minutes to kind of rearrange.
22 (Whereupon, the above-entitled matter went 23 off the record at 3:51 p.m.)
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TVA Clinch River SMR Project Early Site Permit Application Advisory Committee on Reactor Safeguards Full Committee Meeting December 6th, 2018
Acknowledgement and Disclaimer Acknowledgment: "This material is based upon work supported by the Department of Energy under Award Number DE-NE0008336."
Disclaimer: "This presentation was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof."
l 2 Advisory Committee on Reactor Safeguards
Presentation Outline Clinch River Nuclear Site - Overview Dan Stout Early Site Permit Application - Overview Ray Schiele Emergency Preparedness Archie Manoharan l 3 Advisory Committee on Reactor Safeguards
Clinch River Nuclear Site - Overview Dan Stout Director, Nuclear Technology & Innovation l4 Advisory Committee on Reactor Safeguards
TVAs Mission l 5 Advisory Committee on Reactor Safeguards
TVA Clinch River Site Site Access to 500 KV and 161 KV transmission Neighbor to DOE, an interested customer Basic Infrastructure Abundant and skilled workforce Strong community support TVA owned/controlled Advisory Committee on Reactor Safeguards l 6
Early Site Permit Application (ESPA)
An Early Site Permit assesses site suitability for potential construction and operation of a nuclear power plant.
Application includes:
Site Safety Analysis Report to address impacts of the environment on the plant
Environmental Report
Emergency Plans (Part 5A and Part 5B)
Exemptions (Part 6)
ESPA based on a plant parameter envelope (PPE)
Composite of reactor and engineered parameters from four U.S. light-water SMR designs with unique design features that bound the safety and environmental impact of plant construction and operation
Developed based on NEI 10-01 guidance with margin added to specific parameters
Assumes two or more SMR units of a single design
Up to 800MWt for a single unit with a combined nuclear generating capacity not exceeding 2420 MWt (800 MWe)
Advisory Committee on Reactor Safeguards l 7
l 8 NRC Issues ESP Environmental Review Hearing(s)
Safety Review Notice of Hearing, Opportunity 4/4 4 Contentions Filed 6/12 Audits & RAIs Comment Period DEIS 4/26 FEIS 6/21 Scoping Meeting 5/15 PSER 8/4 ACRS Subcomm. Meetings FSER 8/17 SER w/ no OIs 10/20 Commission Hearing Audits & RAIs Notice of Intent 4/13 2 Contentions Admitted 10/10 2017 2018 2019 2020 ESPA Accepted 12-30-16 ESPA Rev. 1 Submitted 12-15-17 ESPA Rev. 2 Planned Submittal Dec 18 5/15 8/22 FEIS FSER Commission Ruling 5/3 ASLB Ruling 7/31 TVA Appeals 11/6 Contested Hearing Terminated Full ACRS 12/5 NRC Review of ESPA 10/17 11/14 Advisory Committee on Reactor Safeguards
ESPA Summary Advisory Committee on Reactor Safeguards l 9 NRC Commenced Review in FY 17 Contains more than 8000 Pages Supported by over 80,000 pages in referenced documents Efficient Use of Audits Few Requests for Additional Information (RAIs)
Frequent, Clear, and Candid Communication
Early Site Permit-Overview Ray Schiele Licensing Manager l10 Advisory Committee on Reactor Safeguards
Application Organization Part 1 - Administrative Information Part 2 - Site Safety Analysis Report
Chapter 1 - Introduction and General Description
Chapter 2 - Site Characteristics
Chapter 3 - Aircraft Hazards
Chapter 11 - Radioactive Waste Management
Chapter 13 - Emergency Planning
Chapter 15 - Transient and Accident Analysis
Chapter 17 - Quality Assurance Part 3 - Environmental Report Part 4 - Limited Work Authorization - Not Used Part 5 - Emergency Plan Part 6 - Exemptions and Departures Part 7 - Withheld Information Part 8 - Enclosures Advisory Committee on Reactor Safeguards l 11
ESPA Development Regulatory bases for the SSAR:
NRC Regulations10 CFR 20, 10 CFR 50, 10 CFR 52, and 10 CFR 100 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition NRC Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition)
RS-002, Processing Applications for Early Site Permits Regulatory bases for the ER:
National Environmental Policy Act, NRC Regulations10 CFR 51 and 10 CFR 52, NRC Regulatory Guide 4.2, Preparation of Environmental Reports for Nuclear Power
- Stations, NRC Regulatory Guide 4.7, General Site Suitability Criteria for Nuclear Power Stations, NUREG-1555, Federal, regional, state and local environmental statutes, as applicable, and RS-002, Processing Applications for Early Site Permits.
Advisory Committee on Reactor Safeguards l 12
ESPA NRC Interactions Pre-Environmental Report Visit PPE Development Pre-application Site Visit Alternative Sites Visit ESPA Readiness Review Hydrology and Health Physics Audit Seismic/Geotechnical Audit Environmental and Meteorology Audit QA Inspection Meteorology and Health Physics Audit March 2013 September 2014 October 2014 June 2015 August 2015 April 2017 May 2017 May 2017 April 2018 May 2018 Advisory Committee on Reactor Safeguards l 13
ASER/ACRS Committee Timeline April 2018 May 2018 July 2018 August 2018 September 2018 October 2018 October 2018 November 2018 December 2018 1st Set ASERs Issued ACRS Subcommittee Meeting SSAR Sections 2.1, 2.2, 3.5.1.6, 15.0.3 ASER - SSAR 13.3 Issued ACRS Subcommittee Meeting SSAR Section 13.3 ASER - SSAR 2.5 Issued ACRS Subcommittee Meeting SSAR Section 2.5 2nd Set ASERs Issued ACRS Subcommittee Meeting SSAR Sections 2.3, 2.4, 11.2/11.3, 17.0 ACRS Full Committee Meeting Advisory Committee on Reactor Safeguards l 14
l 15 Emergency Preparedness Archie Manoharan Licensing Engineer Advisory Committee on Reactor Safeguards
ESPA - Emergency Preparedness Approach l 16 Emergency Planning (EP) Information Layout - 3 Areas Part 2, SSAR, Section 13.3, Emergency Preparedness Plume exposure pathway (PEP) emergency planning size (EPZ) sizing methodology Part 5, Emergency Plan Two major features (Onsite) Emergency Plans Part 5A - Site Boundary EPZ Emergency Plan Part 5B Mile EPZ Emergency Plan Part 6, Exemptions and Departures 2 sets of exemption requests Exemption requests for a PEP EPZ at Site Boundary Exemption requests for a 2-mile PEP EPZ The final EPZ size for the Clinch River Site will be determined at COLA stage Advisory Committee on Reactor Safeguards
PEP EPZ Sizing Methodology l 17 Takes SMR design and safety advancements into consideration Dose-based, consequence-oriented approach to determine an appropriate EPZ size Consistent with the NUREG-0396 sizing rationale - spectrum of accidents are addressed Approach has the same dose criteria as NUREG-0396 - 1 rem total effective dose equivalent (TEDE)
Technical Criteria - PEP EPZ should:
Criterion A - encompass those areas in which projected dose from design basis accidents (DBAs) could exceed the U.S. Environmental Protection Agency (EPA) early phase protective action guide (PAG)
Criterion B - encompass those areas in which consequences of less severe core melt accidents could exceed the EPA early phase PAG Criterion C - be of sufficient size to provide for substantial reduction in early health effects in the event of more severe core melt accidents Advisory Committee on Reactor Safeguards
PEP EPZ Sizing Methodology l 18 Step 1 - Accident scenario selection
DBA from Chapter 15
Design and site specific Probabilistic Risk Assessment (PRA) for severe accident scenarios
Considers - all modes, internal & external events, applicable fuel handling, spent fuel pool, and multi-module accidents
Sequences with mean core damage frequency (CDF) greater than 1E-8 per reactor-year (rx-yr)
Criterion B: Less severe core melt scenarios - Mean CDF greater than 1E-6 per rx-yr, intact containment
Criterion C: More severe core melt scenarios - Mean CDF greater than 1E-7 per rx-yr, containment bypass or failure Step 2 - Determine source term releases from selected accidents Step 3 - Calculate dose consequences at distance Step 4 - Compare the dose at distance to EPA early phase PAG COL applicant would perform an analysis using the PEP EPZ size methodology, with site-and design-specific input, to justify the PEP EPZ size for the COLA Advisory Committee on Reactor Safeguards
PEP EPZ Sizing Methodology - Example Analysis l 19 Criteria A & B: DBA and less severe accidents
Dose consequences do not exceed the early phase EPA PAG - 1 rem total effective dose equivalent (TEDE)
Criterion C: More severe accidents
Calculate distance at which conditional probability to exceed 200 rem whole body exceeds 1E-3 per rx-yr
Verify the PEP EPZ is of sufficient size to provide for substantial reduction in early health effects Design-Specific Example Analysis - Evaluates NuScale Power Plant at Clinch River Site Criteria Site Boundary Dose TEDE (rem)
EPA Early Phase PAG Limit TEDE (rem)
A: Design Basis Accidents 0.104 1
B: Less Severe Core Melt Accidents 0.158 1
C: Reduction in Early Severe Health Effects No accident scenarios met the required screening criteria.
Advisory Committee on Reactor Safeguards
Part 5 - Emergency Plan l 20 Part 5 of the ESPA contains the major features of two distinct Emergency Plans for Clinch River Site in accordance with 10 CFR 52.17(b)(2)(i).
Part 5A Describes major features of an Emergency Plan for a PEP EPZ consisting of the area encompassed by the Site Boundary.
Part 5B Describes major features of an Emergency Plan for a PEP EPZ consisting of an area approximately two miles in radius surrounding the Clinch River Site.
Both plans address the 16 planning standards in NUREG-0654,Section II, which reflects the requirements in 10 CFR 50.47(b)(1) through 10 CFR 50.47(b)(16) and Appendix E to 10 CFR Part 50 considering the requested exemptions described in Part 6 of the ESPA Advisory Committee on Reactor Safeguards
Part 6 - Exemptions and Departures Advisory Committee on Reactor Safeguards l 21 Pursuant to 10 CFR 52.7, Specific Exemptions, which is governed by 10 CFR 50.12, Specific Exemptions, TVA requested exemptions from the following emergency preparedness requirements for the Clinch River Site:
Certain standards in 10 CFR 50.47(b) regarding onsite and offsite emergency response plans for nuclear power reactor
Certain requirements of 10 CFR 50.33(g) and 10 CFR 50.47(c)(2) to establish PEP EPZ for nuclear power plants
Certain requirements of 10 CFR Part 50, Appendix E, which establish the elements that make up the content of emergency plans Two Sets of Exemptions
Exemptions for a PEP EPZ established at the Site Boundary
Deviate from 10-mile PEP EPZ
Various elements of a formal offsite emergency plan
Evacuation time estimates
Certain elements of offsite notifications and exercises
Exemptions for an approximate 2-mile PEP EPZ
Deviate from 10-mile PEP EPZ
Emergency Preparedness Approach - Summary l 22 ESPA COLA PEP EPZ Methodology (Part 2, SSAR, Section 13.3)
Approval of the dose-based, consequence oriented methodology for determining the PEP EPZ size Approval of design specific implementation of the methodology approved in the ESPA EPZ Size (Part 6)
Approval to deviate from the current 10-mile PEP EPZ requirements based on the methodology to determine PEP EPZ size Approval of design specific PEP EPZ size based on design specific implementation of the methodology Emergency Plan (Part 5)
Approval of the major features of the Site Boundary and 2-mile emergency plans presented in Part 5 Approval of the remaining elements of either the Site Boundary or 2-mile emergency plans OR a new plan based on design specific PEP EPZ size using methodology Advisory Committee on Reactor Safeguards
Mallecia Sutton, Project Manager, NRO/DLSE/LB3 Allen Fetter, Project Manager, NRO/DLSE/LB3 Section 13.3 Emergency Planning Michelle Hart, Technical Reviewer, NRO/DLSE/RPAC Bruce Musico, Technical Reviewer, NSIR/DPR/RLB Presentation to the ACRS Full Committee Clinch River Nuclear Site - Early Site Permit Application (ESPA) Safety Review December 6, 2018
Clinch River Nuclear Site ESP Application Review Overview Tennessee Valley Authority (TVA) submitted an ESPA for the Clinch River Nuclear Site to NRC (May 26, 2016)
Application accepted for docketing and detailed technical review on December 30, 2016. Federal Register Notice on acceptance decision (January 12, 2017)
TVA requested permit approval for a 20-year term along with approval for a plume exposure pathway (PEP) emergency planning zone (EPZ) sizing methodology, 2 major features (onsite) emergency plans, and exemption requests for site boundary and 2-mile PEP EPZs Plant Parameter Envelope (PPE) based on four small modular reactor (SMR) designs 2
Staff Review Staff overview presentation to ACRS on ESP, PPE and Clinch River ESP review schedule (November 15, 2017)
NRC Staffs safety review of the application included 5 audits and 1 inspection, and issuance of 12 request for additional information (RAIs)
(comprising 50 questions)
Staff completed all Advanced Safety Evaluations (ASEs) with no Open Items and presented to ACRS Subcommittee (May 15, 2018 -
November 14, 2018)
ASEs include 42 combine license application (COL) Action Items and 8 Permit Conditions Staff cooperated with U.S. Army Corps of Engineers, consulted with Federal Emergency Management Agency, and engaged with U.S.
Department of Energy, Tennessee Department of Environment and Conservation, the U.S. Geological Survey and the Tennessee Emergency Management Agency 3
Approving an ESP Site without a Selected Reactor Technology
- ESP Plant Parameter Envelope (PPE) values can bound a variety of reactor technologies rather than one specific technology (an amalgam of values representing a surrogate nuclear plant)
- The PPE values are bounding criteria used by staff to determine the suitability of an ESP site for construction and operation of a nuclear plant
- In the combined license application (COLA), when a specific technology is identified, the PPE values are compared to those of the selected technology. If design parameters of the selected technology exceed bounding ESP PPE values, additional reviews are conducted to ensure that the site remains suitable from a safety and environmental standpoint for construction and operation of the selected nuclear plant technology ESP Plant Parameter Envelope 4
ESP Plant Parameter Envelope (contd)
TVA used the following reactor designs to develop the Plant Parameter Envelope (PPE):
- NuScale SMR, 160 MWt (50 MWe)
- Holtec SMR-160, 525 MWt (160 MWe)
- Westinghouse SMR, 800 MWt (225 MWe)
TVAs PPE is based on construction and operation of two or more SMRs at the Clinch River Nuclear Site with a maximum site nuclear generating capacity of 2420 MWt (800 MWe) 5
Safety Evaluation Sections Chapter Sections Accession Numbers 2.1 Geography and Demography ML18102B203 2.2 Nearby Industrial Transportation and Military Facilities ML18102B203 2.3 Meteorology ML17289B148 2.4 Hydrologic Engineering ML17289B151 (NP)
ML18290A685 (P) 2.5.1 Geologic Characterization ML17289B252 2.5.2 Vibratory Ground Motion ML17289B253 2.5.3 Surface Deformation ML17289B254 2.5.4 Stability of Subsurface Materials and Foundations ML17289B255 2.5.5 Stability of Slopes ML17289B255 3.5.1.6 Aircraft Hazards ML18102B150 11.2 & 11.3 Radioactive Waste Management ML17289A625 13.3 Emergency Planning ML17291A052 15.0.3 Radiological Consequences of Design Basis Accidents ML18102B149 17.5 Quality Assurance Program Description ML17291A547 6
TVA provided adequate information pertaining to;
- the site setting and boundaries
- Exclusion Area Boundary (EAB) authority and control
- current and future population projections
- low population zone (LPZ) distance, population center distance and population density Based on the information provided by the applicant and staffs independent confirmatory evaluation, the staff found the information to be acceptable as it meets the requirements of 10 CFR 100.20 Section 2.1 Geography and Demography 7
TVA adequately identified potential sources and hazards in site vicinity TVA adequately evaluated potential accidents pertaining to explosions, vapor cloud explosions, hazardous/toxic chemical vapors, and fires Based on the information provided by the applicant and staffs independent confirmatory evaluation, the staff found the information to be acceptable as the information meets the guidance provided in NUREG-0800 Section 2.2.1-2.2.2 Section 2.2 Nearby Industrial, Transportation, and Military Facilities 8
Site characteristics related to extreme weather (hurricane and tornado winds, winter precipitation, temperature and humidity extremes) are acceptable Onsite meteorological monitoring system provides adequate data to represent meteorological dispersion conditions Site characteristics related to Short-Term (Accident) and Long-Term (Routine Release) dispersion estimates (X/Q and D/Q values) are acceptable Based on the information provided by the applicant, the staff found all regulatory requirements have been satisfied with no open items Section 2.3 - Meteorology 9
Short-Term (Accident) X/Q Values
- Exclusion Area Boundary (335 meters)
- Low Population Zone (1609 meters)
Based on PAVAN Atmospheric Dispersion Model
- Gaussian model
- Various time averaging periods
- 0-2 hr @ EAB
- 0-8 hr, 8-24 hr, 1-4 days, and 4-30 days @ LPZ
- Intended to represent dispersion conditions that are exceeded no more than 5% of the time Used to demonstrate compliance with 10 CFR 52.17(a)(1)(ix) dose guidelines for design basis accidents
- 25 rem at the EAB for any 2-hour period following the onset of the release
- 25 rem at the outer boundary of the LPZ for the duration of the release Short-Term (Accident) X/Q Values 10
Section 2.4 Hydrologic Engineering TVA proposed adequate site characteristics and bounding design parameters for inclusion in the ESP Design basis flood and maximum groundwater levels, and the accidental release dose estimate meet regulatory requirements Staff concludes that applicant meets ESP regulatory requirements associated with hydrologic engineering 11
Geologic Site Characterization (Section 2.5.1) - No tectonic features with the potential for adversely affecting suitability of the site occur in the site region, site vicinity, site area, or at the site location Vibratory Ground Motion (Section 2.5.2) - Applicants ground motion response spectrum adequately represents the regional and local seismic hazards, and accurately includes the potential effects of local site-specific subsurface properties Surface Deformation (Section 2.5.3) - Negligible potential exists for tectonic surface deformation at the site. Karst is the primary potential hazard for non-tectonic surface deformation that could adversely affect the site Stability of Subsurface Materials and Foundations (Section 2.5.4) - Applicant adequately determined the engineering properties of subsurface materials at the site, and properly evaluated the stability of subsurface materials and foundations based on results of field and laboratory tests and state-of-the-art methodology Stability of Slopes (Section 2.5.5) - Applicant provided necessary information on site topography and geologic conditions, and adequately described characteristics of slopes at the site Section 2.5 Geology, Seismology and Geotechnical Engineering 12
For site suitability, aircraft accidents should not lead to radiological consequences in excess of the exposure guidelines of 10 CFR 50.34(a)(1) with a probability of occurrence greater than about 10-7 per year The applicant determined an aircraft crash probability of 7.53 x 10-7 per year from two nearby airways not associated with local airport operations The staff conservatively estimates a potential aircraft crash probability of 1.5 x 10-8 per year (bounding the applicants probability), assuming all flights within 10 miles of the site follow the two airways passing near the site Staff finds that the applicants approach is reasonable and the probability value is acceptable Section 3.5.1.6 Aircraft Hazards 13
Chapter 11 Radioactive Waste Management, Sections 11.2.3 and 11.3.3 Applicants methodology to develop the normal PPE liquid and gaseous effluent release source terms for use in calculating offsite doses is reasonable Normal PPE liquid and gaseous effluent release concentrations meet the unity rule in 10 CFR Part 20, Appendix B, Table 2, Columns 1 and 2 Offsite doses from normal PPE liquid and gaseous effluent release source terms meet the design objectives in 10 CFR Part 50, Appendix I, Sections II.A, II.B, and II.C; Environmental Protection Agencys (EPA) radiation standards in 40 CFR Part 190, as implemented under 10 CFR 20.1301(e);
and public dose limit in 10 CFR 20.1301 Reactor designs falling within the normal PPE effluent release source terms and offsite doses for the Clinch River Nuclear Site are without undue risk to public health and safety 14
Evaluation of the radiological consequences of postulated design basis accidents (DBAs) is based on the PPE accident source term for DBA isotopic releases to the environment (in lieu of specific plant design information) in conjunction with site characteristic short term (accident) atmospheric dispersion factors The same dose criteria are used for siting and postulated accident dose analysis requirements:
The evaluation must determine that:
1.
An individual located at any point on the boundary of the exclusion area for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the postulated fission product release would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE).
2.
An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE Staff concluded that the applicants analysis meets the dose criteria specified, and the PPE includes the bounding accident releases for the determination Chapter 15 Accident Analysis 15
NRC Staff identified one RAI, March 9, 2018 NRC Staff conducted Quality Assurance Implementation Inspection, April 16-20th 2018.
TVA issued Nuclear Quality Assurance Plan, Revision 36; May 8, 2018 Staff concluded that the applicants quality assurance program description for the Clinch River Nuclear site ESP application meets the requirements of 10 CFR Part 50, Appendix B and 10 CFR 52.17(a)(1)(xi) and (xii)
Section 17.5 Quality Assurance Program Description 16
13.3 Emergency Planning The ESPA requested review of 3 key areas, which consist of:
Plume exposure pathway (PEP) emergency planning zone (EPZ) sizing methodology 2 major features (onsite) emergency plans (ESPA Part 5)
- ESPA Part 5A reflects a site boundary PEP EPZ
- ESPA Part 5B reflects a 2-Mile PEP EPZ (including an ETE) 25 Exemption Requests (ESPA Part 6)
- 2 exemption requests (applicable to both the site boundary and 2-mile PEP EPZs)
- 23 exemption requests address portions of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50 for offsite emergency planning (EP) related to the site boundary PEP EPZ only 17
25 Exemption Requests (EP) 10 CFR 50.33(g) & 50.47(c)
- 2 requests for exemptions from the 10-mile PEP EPZ requirement 10 CFR 50.47 & Appendix E to 10 CFR Part 50
- 23 requests for exemption from the emergency planning requirements associated with offsite emergency planning
- State & local emergency plans
- Public alert & notification
- Evacuation Time Estimate (ETE)
- Offsite exercises 18
Basis for Acceptance The ESPA provides a basis for the establishment (in the COLA) of either a Site Boundary or 2-mi PEP EPZ, which maintains the same level of protection (i.e., dose savings in the event of a radiological emergency) in the environs of the Clinch River Nuclear Site as that which exists in the basis for a 10-mi PEP EPZ 19
Combined License Application Upon issuance of the ESP, the applicant acquires approval, with conditions, of:
- The PEP EPZ sizing methodology
- The 2 major features emergency plans (site boundary/2-mile PEP EPZ)
- The 25 requested exemptions A COLA that incorporates by reference the ESP must:
- Identify the chosen SMR technology for the Clinch River Nuclear site
- Provide a complete & integrated emergency plan
- 2-mile PEP EPZ must provide onsite & offsite emergency plans
- site boundary PEP EPZ must provide an onsite emergency plan
- Address all 16 COL Action Items and 4 Permit Conditions 20
EPZ Size Determination in COLA COL Action Item 13.3-1 (reflects ESPA Part 2 Section 13.3.3.1.4)
- Identify chosen SMR technology & major features emergency plan
- Provide detailed information that shows the ability of the SMR to meet the chosen PEP EPZ
- The selected SMR technology must meet the EPA early phase protective action guide (PAG)
Permit Condition 1
- Provide detailed information to demonstrate that the accident release source term information for the PEP EPZ size determination analysis using the selected SMR design is bounded by the non-design-specific plant parameter source term information used in the analysis supporting the exemption requests (ASER Table 13.3-1)
- Based on non-design-specific bounding 4-day accident release source term that meets EPZ size criteria 21
TVA PEP EPZ Size Methodology Technical Criteria PEP EPZ should encompass those areas in which projected dose from DBAs could exceed the EPA early phase PAG PEP EPZ should encompass those areas in which consequences of less severe core melt accidents could exceed the EPA early phase PAG PEP EPZ should be of sufficient size to provide for substantial reduction in early health effects in the event of more severe core melt accidents 22
TVA PEP EPZ Size Methodology SSAR Section 13.3.3.1 Accident scenario selection
- Use bounding DBA from COLA Final Safety Analysis Report Chapter 15
- Use COLA site-and design-specific probabilistic risk assessment to categorize severe accident scenarios
- All modes, internal and external events, applicable fuel handling and spent fuel pool accidents, multi-module considerations
- Assess all sequences with mean core damage frequency (CDF) > 10-8 per rx-yr
- More probable, less severe core melt scenarios
- Mean CDF > 10-6 per rx-yr
- Intact containment
- Less probable, more severe core melt scenarios
- Mean CDF > 10-7 per rx-yr
- Includes containment bypass or failure Determine source term releases to atmosphere Calculate dose consequences at distance from plant Determine PEP EPZ size that meets the dose-based criteria 23
TVA Dose-Based PEP EPZ Size Criteria Dose to individual from exposure to the airborne plume during its passage and to groundshine, using average atmospheric dispersion characteristics for site Staff expects the applicant may use the calculation tools used for severe accident consequence analysis in environmental report DBA and more probable, less severe accidents 1 rem TEDE from 96-hr exposure Lower end of dose range EPA PAG for early phase protective actions (e.g.,
evacuation and sheltering)
Verify that dose consequences do not exceed the EPA PAG beyond the site boundary (within owner controlled area) and 2-mile PEP EPZs Less probable, more severe accidents Calculate the distance at which the conditional probability to exceed 200 rem whole body from 24-hr exposure exceeds 10-3 per rx-yr Acute dose at which radiation-induced early health effects may begin to be noted (e.g., nausea)
Verify that the PEP EPZ supports substantial reduction in early health effects 24
Review of PEP EPZ Size Methodology Staff compared TVAs methodology and dose criteria to the study used as technical basis for current 10-mile PEP EPZ requirement (NUREG-0396)
- The features of TVAs methodology are consistent with NUREG-0396
- Considered a range of accidents
- Performed accident consequence analyses
- Determined an area outside of which early protective actions are not likely to be necessary to protect the public from radiological releases The staff concludes that the applicants proposed methodology is reasonable, and consistent with the analyses that form the technical basis for the current regulatory requirement of a PEP EPZ of about 10 miles in radius 25
EP Exemption Plant Parameters TVA developed a non-design-specific accident release source term that would meet the PEP EPZ size criteria to be used as plant parameters (ASER Table 13.3-1)
- Isotopic total release activity over 96 hrs results in TEDE of about 0.9 rem at site boundary
- Same idea as PPE DBA source term to envelope an unknown design
- Referenced in Permit Condition 1 for adoption of EP exemptions 26
Section 13.3 EP Conclusions The staff concludes that:
- The PEP EPZ sizing methodology is acceptable for determining the appropriate size of the PEP EPZ for the Clinch River Nuclear site because it is consistent with the analyses that form the technical basis for the current 10-mile PEP EPZ
- The 2 major features emergency plans are acceptable because they meet the applicable standards of 10 CFR 50.47 and requirements of Appendix E to 10 CFR Part 50
- The exemption requests are acceptable because they are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and special circumstances are present 27
Questions?
28
Technical Reviewers Dan Barss Luissette Candelario Yuan Cheng Richard Clement Joseph Giacinto Michelle Hart David Heeszel Michael Mazaika Bruce Musico Kevin Quinlan Nicholas Savwoir Gerry Stirewalt Seshagiri (Rao) Tammara Jenise Thompson Weijun Wang Jason White 29
Acronyms ASE - Advanced Safety Evaluation CFR - Code of Federal Regulations COL - Combined License COLA - Combined License Application CDF - Core Damage Frequency CP - Construction Permit CRN - Clinch River Nuclear DBA - Design Basis Accidents DBF - Design Basis Flood EAB - Exclusion Area Boundary EP - Emergency Planning EPA - Environmental Protection Agency EPZ - Emergency Planning Zone ESP - Early Site Permit ESPA - Early Site Permit Application ETE - Evacuation Time Estimate FRN - Federal Register Notice LOCA - Loss of Coolant Accident LPZ - Low Population Zone NP-Non-Public MWe - Megawatts Electric MWt - Megawatts Thermal NP-Non-Public NRC - Nuclear Regulatory Commission P-Public PAG - Protective Action Guide PEP - Plume Exposure Pathway PPE - Plant Parameter Envelope RAI - Request for Additional Information SER - Safety Evaluation Report SMR - Small Modular Reactor SSCs - Structures, Systems and Components TEDE - Total Effective Dose Equivalent TVA - Tennessee Valley Authority USGS - U.S. Geological Survey 30
Advisory Committee on Reactor Safeguards Full Committee Meeting December 5, 2018 Seabrook Station Unit 1 License Renewal Application
2 The foundation for everything we do are the Values and Core Principles of our Nuclear Excellence Model
3 Agenda Introduction Site and Station Description License Renewal Application and Aging Management Programs Safety Evaluation Report and Closure of Previous Open Items Closing Remarks NextEra Energy Seabrook has met the requirements of 10 CFR 54 for issuance of a renewed licensed for Seabrook Station Unit 1
4 Personnel in Attendance Eric McCartney Regional Vice President -
Northern Region Michael Collins Engineering Director Ken Browne Licensing Manager Edward Carley License Renewal Supervisor Rudy Gil Programs Engineering Manager
5 Site and Station Description
6 Plant Status Completed latest refuel outage (OR19) 10/29/18 Next Refuel Outage - Spring 2020 (End of Cycle 20)
Capacity Factor 15 of 19 cycles > 94%
Lifetime 87%
Lifetime excluding refueling outages 95.2%
Cycle 19:
99.86%
Cycle 18:
98.34%
Cycle 17:
99.27%
Cycle 16:
99.71%
Capacity factor performance is representative of solid equipment reliability and material condition
7 Recent Station Improvements
- Main Generator Stator Rewind
- Vital Batteries
- Vital Inverters
- Generator Step-Up Transformers
- Mechanical Stress Improvement Process completed for all Reactor Vessel Nozzles
- Service Water Piping (AL6XN)
- Incore Detectors
- Process Control Single Point Vulnerability Circuit Cards
- Solid State Protection System Circuit Cards
- Rod Control Motor/Generator Sets
- Shutdown Reactor Coolant Pump Seals NextEra Energy Seabrook is committed to maintaining high levels of safety, reliability and performance
8 License Renewal Application Scoping and Screening Aging Management Review Time Limited Aging Analysis (TLAA)
UFSAR Supplement Commitments Aging Management Programs Environmental Report
- Severe Accident Mitigation Alternatives (SAMA) Analysis
9 GALL Consistency AMPS Consistent Consistent with Enhancements Consistent with Exceptions Consistent with Exception and Enhancements Plant Specific New 15 7
1 2
1 4
Existing 29 8
12 2
5 2
Total 44
10 Safety Evaluation Report SER Issued September 28, 2018 No open items No confirmatory items Closure of Open Items from previous SER (2012)
OI 3.0.3.2.2-1 Steam Generator Tube Integrity OI 4.2.4-1 Pressure-Temperature Limit OI 3.2.2.1-1 Treated Borated Water OI 3.0.3.1.7-1 Bolting Integrity Program OI B.1.4-2 Operating Experience OI 3.0.3.1.9-1 ASME Section XI, IWE Program OI 3.0.3.2.18-1 Structures Monitoring Program
11 Approach for Addressing ASR at Seabrook Station NextEra Energy Seabrook has implemented an effective program for evaluating and managing the impacts of ASR on affected concrete structures and associated SSCs
12 Concluding Remarks Seabrook is committed to the continuous improvement and long-term operation of Seabrook Station Seabrook will manage the effects of aging in accordance with 10 CFR 54.21(a)(1)
Seabrook has evaluated time-limited aging analyses that require evaluation under 10 CFR 54.21(c)
Seabrook has met the provisions of 10 CFR 54 for issuance of a renewed license NextEra Energy Seabrook has demonstrated compliance with the requirements of 10 CFR 54 for issuance of a renewed licensed for Seabrook Station Unit 1
Advisory Committee on Reactor Safeguards Full Committee Seabrook Station, Unit 1 Safety Evaluation Report (SER)
December 6, 2018 William Butch Burton, Project Manager Office of Nuclear Reactor Regulation
Presentation Outline
- Overview of Seabrook license renewal review
- Conclusion 2
License Renewal Review:
Audits and Inspections Onsite 3
Audit / Inspection Dates Scoping & Screening Methodology Audit (ML110270026)
September 20 - 23, 2010 Aging Management Program (AMP)
Audits (ML110280424)
October 12 - 15, 2010 October 18 - 22, 2010 Region I 71002 Inspection: Scoping, Screening, and AMPs (ML111360432)
March 7 - 11, 2011 March 21 - 25, 2011 April 4 - 8, 2011 Region I 71002 Inspection: AMPs for Alkali-Silica Reaction (ASR)
April 30 - May 3, 2018
SER Overview
- SER with 7 Open Items issued June 2012
- 1. Bolting Integrity Program
- 3. Steam Generator Tube Integrity Program
- 4. Operating Experience
- 5. Treated Borated Water
- 6. Pressure-Temperature Limit
- 7. Structures Monitoring Program/ASR
- Open items closed on September 28, 2018 4
SER Overview
- SER with 7 Open Items issued June 8, 2012
- Staff met with ACRS Subcommittee on Plant License Renewal on July 10, 2012
- Final SER issued September 28, 2018
- No open items or confirmatory items
- Total of 291 RAIs issued
- 58 follow-up RAIs
- Additional meetings with ACRS Subcommittee on Plant License Renewal held October 31 and November 15, 2018 5
SER Section 2 6
- Structures and Components Subject to Aging Management Review (AMR)
- Section 2.1: Scoping and Screening Methodology
- Section 2.2: Plant-Level Scoping Results
- Sections 2.3, 2.4, 2.5: Scoping and Screening Results
SER Section 3
- Aging Management Review (AMR) Results
- Section 3.1: Aging Management of Reactor Vessel, Internals, and Reactor Coolant System
- Section 3.2: Aging Management of Engineered Safety Features
- Section 3.3: Aging Management of Auxiliary Systems
- Section 3.4: Aging Management of Steam and Power Conversion Systems
- Section 3.5: Aging Management of Containments, Structures and Component Supports
- Section 3.6: Aging Management of Electrical Commodity Group 7
SER Section 3 NextEras Disposition of AMPs 13 new programs
6 consistent
1 consistent with enhancements
3 consistent with exceptions
3 consistent with enhancements and exceptions 29 existing programs
10 consistent
10 consistent with enhancements
3 consistent with exceptions
4 consistent with enhancements and exceptions
2 plant specific Final Disposition of AMPs in SER 15 new programs
7 consistent
1 consistent with enhancement
2 consistent with exceptions
1 consistent with enhancements and exceptions
4 plant specific 29 existing programs
8 consistent
12 consistent with enhancements
2 consistent with exceptions
5 consistent with enhancements and exceptions
2 plant specific 8
Section 3.0.3 - Aging Management Programs (AMPs)
- Time-Limited Aging Analyses (TLAAs) 4.1: Identification of TLAAs 4.2: Reactor Vessel Neutron Embrittlement Analyses 4.3: Metal Fatigue Analyses 4.4: Environmental Qualification of Electric Equipment 4.5: Concrete Containment Tendon Prestress Analyses 4.6: Containment Liner Plate, Metal Containment, and Penetrations Fatigue Analyses 4.7: Other Plant-Specific TLAAs SER Section 4 9
10 On the basis of its review, the staff finds that the requirements of 10 CFR 54.29(a) have been met for the license renewal of Seabrook Station, Unit 1.
Conclusion