ML19086A344
| ML19086A344 | |
| Person / Time | |
|---|---|
| Issue date: | 03/27/2019 |
| From: | Quynh Nguyen Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Nguyen Q | |
| References | |
| Download: ML19086A344 (253) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 March 27, 2019 MEMORANDUM TO: ACRS Members FROM: Quynh Nguyen, Senior Staff Engineer /RA/
Technical Support Branch, ACRS
SUBJECT:
CERTIFIED MINUTES OF THE ACRS REGULATORY POLICIES AND PRACTICES SUBCOMMITTEE MEETING ON NOVEMBER 14, 2018 The minutes of the subject meeting were certified on February 27, 2017, as the official record of the proceedings of that meeting. Copies of the certification letter and minutes are attached.
Attachments: As stated cc w/ att. A. Veil L. Burkhart
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 MEMORANDUM TO: Quynh Nguyen, Senior Staff Engineer Technical Support Branch, ACRS FROM: Walter Kirchner, Chairman Regulatory Policies and Practices Subcommittee
SUBJECT:
CERTIFICATION OF THE MINUTES OF THE REGULATORY POLICIES AND PRACTICES SUBCOMMITTEE MEETING ON NOVEMBER 14, 2018 I hereby certify, to the best of my knowledge and belief, that the minutes of the subject meeting are an accurate record of the proceedings for that meeting.
/RA/ February 27, 2019 Walter Kirchner, Chairman Dated Regulatory Policies and Practices Subcommittee
Certified on: February 27, 2019 Certified by: Walter Kirchner ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MINUTES OF THE REGULATORY POLICIES & PRACTICES SUBCOMMITTEE MEETING November 14, 2018 The ACRS Regulatory Policies & Practices Subcommittee held a meeting on November 14, 2018 in 3WFN 1C3 & 1C5, 111601 Landsdown Street, North Bethesda, Maryland. The meeting convened at 8:30 am and adjourned at 11:40 am.
The entire meeting was open to the public.
No written comments or requests for time to make oral statements were received from members of the public related to this meeting.
ATTENDEES ACRS Members/Consultants/Staff WALTER L. KIRCHNER, Chairman DENNIS C. BLEY*, Member RONALD G. BALLINGER, Member MICHAEL L. CORRADINI, Member MATTHEW W. SUNSERI, Member QUYNH NGUYEN, Designated Federal Official CHARLES H. BROWN, JR., Member PETER RICCARDELLA, Member
- via telephone Other
Participants:
ANDY CAMPBELL, NRO YUAN CHENG, NRO RICHARD CLEMENT, NRO MICHELLE CONNER, TVA ALLEN FETTER, NRO JOSEPH GIACINTO, NRO HILLOL GUHA, TVA STU HENRY, TVA JOHN HOLCOMB, TVA KEVIN QUINLAN, NRO NICHOLAS SAVWOIR, NRO RAYMOND SCHIELE, TVA MALLECIA SUTTON, NRO ALEX YOUNG, TVA
KEVIN CASEY, TVA KELVIN MONTAGUE, TVA SHARAD JHA, BECHTEL KIM HUMMER, BECHTEL AARON ENGLAND, TVA REBECCA CARR, BECHTEL MELISSA BATES, DOE
SUMMARY
The purpose of this meeting is the review of the following sections: Meteorology (2.3);
Hydrologic Engineering (2.4); Radioactive Waste Management (11); and Quality Assurance (17) of Tennessee Valley Authoritys (TVA) Clinch River Early Site Permit (ESP) application.
The meeting transcripts are attached and contain an accurate description of each matter discussed during the meeting. The presentation slides and handouts used during the meeting are attached to these transcripts.
SIGNIFICANT ISSUES Reference Pages Issue in Transcript Mr. Young (TVA) discussed Section 11. Mr. Clement (NRC) also contributed to this topic (18), including the methodology for confirmatory 12 analysis (21-26).
Mr. Young discussed Section 2.3, including the use of historical data.
Chairman Kirchner asked for an explanation for the use of vector and 28-58 scalar wind data (34). Mr. Quinlan (NRC) talked about this section including atmospheric dispersion.
Ms. Conner (TVA) and Mr. Savwoir (NRC; 63) discussed quality 58-71 assurance.
Mr. Holcomb (TVA) discussed hydrologic engineering, including dam 72-89 locations, flooding and ice effects.
Mr. Henry (TVA) and Mr. Guha (TVA) talked about groundwater movement modeling, including the aquifer performance test (96). The 89-114, 114-129 NRC staff provided their results on hydrology.
Documents provided to the Subcommittee
REFERENCES:
- 1. Safety Evaluation, Meteorology (2.3) (ML17289B148); Hydrologic Engineering (2.4)
(ML17289B151); Radioactive Waste Management (11) (ML17289A625); and Quality Assurance (17) (ML17291A547) of Tennessee Valley Authoritys (TVA) Clinch River Early Site Permit (ESP) application 2
1 Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION OFFICIAL USE ONLY OI INVESTIGATION INFORMATION
Title:
Regulatory Policies and Practices Subcommittee Docket Number: N/A Location: Rockville, Maryland Date: 11-14-18 Work Order No.: NRC-3983 Pages 1-138 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
+ + + + +
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS)
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REGULATORY POLICIES AND PRACTICES SUBCOMMITTEE
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WEDNESDAY NOVEMBER 14, 2018
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ROCKVILLE, MARYLAND
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The Subcommittee met at the Nuclear Regulatory Commission, Three White Flint North, Room 1C3 & 1C5, 11601 Landsdown Street, at 8:30 a.m., Walter L. Kirchner, Chairman, presiding.
COMMITTEE MEMBERS:
WALTER KIRCHNER, Chairman RONALD G. BALLINGER, Member DENNIS C. BLEY, Member*
CHARLES H. BROWN, JR., Member MICHAEL L. CORRADINI, Member PETER C. RICCARDELLA, Member MATTHEW W. SUNSERI, Member
3 DESIGNATED FEDERAL OFFICIAL:
QUYNH NGUYEN ALSO PRESENT:
ANDY CAMPBELL, DLSE YUAN CHENG, NRO RICHARD CLEMENT, NRO MICHELLE CONNER, TVA ALLEN FETTER, NRO JOSEPH GIACINTO, NRO HILLOL GUHA, TVA STU HENRY, TVA JOHN HOLCOMB, TVA KEVIN QUINLAN, NRO NICHOLAS SAVWOIR, NRO RAYMOND SCHIELE, TVA MALLECIA SUTTON, NRO ALEX YOUNG, TVA
- Present via telephone
4 A G E N D A Opening Remarks....................................4 Introductions and Overview.........................7 Section 11: Radioactive Waste Management Tennessee Valley Authority (TVA)............11 NRC Staff...................................15 Section 2.3: Meteorology Tennessee Valley Authority (TVA)............27 NRC Staff...................................40 Section 17: Quality Assurance Tennessee Valley Authority (TVA)............58 NRC Staff...................................62 Section 2.4: Hydrology Tennessee Valley Authority (TVA)............71 NRC Staff..................................115 Public Comment...................................130 Next Steps.......................................131 Adjournment......................................138
5 1 P R O C E E D I N G S 2 8:30 a.m.
3 CHAIRMAN KIRCHNER: The meeting will now 4 come to order. This is a meeting of the Regulatory 5 Policies and Practices Subcommittee of the Advisory 6 Committee on Reactor Safeguards.
7 I am Walt Kirchner, Chairman of this 8 Subcommittee meeting. ACRS members in the room are:
9 Mike Corradini, Pete Riccardella, Matt Sunseri, Ron 10 Ballinger, Charlie Brown, and I think we'll see if 11 others join us. Quynh Nguyen of the ACRS staff is the 12 Designated Federal Official for this meeting.
13 The Subcommittee will hear from 14 representatives of TVA and the staff regarding the 15 following sections of the Clinch River early site permit 16 application and the corresponding Safety Evaluation:
17 Meteorology, 2.3; Hydrologic Engineering, 2.4; 18 Radioactive Waste Management, 11; and Quality 19 Assurance, Chapter 17.
20 The Subcommittee will gather information, 21 analyze relevant issues and facts, and formulate 22 proposed positions and actions, as appropriate, for 23 deliberation by the full Committee.
24 The ACRS was established by statute and 25 is governed by the Federal Advisory Committee Act, FACA.
6 1 This means that the Committee can only speak through 2 its published letter reports. We hold meetings to 3 gather information to support our deliberations.
4 Interested parties who wish to provide 5 comments can contact our offices requesting time after 6 the meeting announcement is published in the Federal 7 Register.
8 That said, we also set aside some time for 9 spur of the moment comments from members of the public 10 attending or listening to our meetings. Written 11 comments are also welcome.
12 In regard to early site permits, 10 CFR 13 52.23 provides that the Commission shall refer a copy 14 of the application to the ACRS and the Committee shall 15 report on those portions which concern safety.
16 The ACRS section of the US NRC public 17 website provides our charter, bylaws, letter reports, 18 and full transcripts of all full and Subcommittee 19 meetings, including slides presented at those meetings.
20 The rules for participation in today's 21 meeting were previously announced in the Federal 22 Register. We have received no written comments or 23 requests for time to make oral statements from the 24 members of the public regarding today's meeting.
25 We have a bridge line established for
7 1 interested members of the public to listen in. To 2 preclude interruption of the meeting, the phone bridge 3 will be placed in a listen-in mode during the 4 presentations and Committee discussions.
5 We will unmute the bridge line at a 6 designated time to afford the public an opportunity 7 to make a statement or provide comments.
8 At this time, I request that meeting 9 attendees and participants silence cell phones and any 10 other electronic devices that are audible.
11 A transcript of the meeting is being kept 12 and will be made available as stated in the Federal 13 Register notice. Therefore, we request that 14 participants in this meeting use the microphones 15 located throughout the meeting room when addressing 16 the Subcommittee.
17 The participants should first identify 18 themselves and speak with sufficient clarity and volume 19 so that they may be readily heard. Make sure that the 20 green light at the base of the microphone is on before 21 speaking and off when not in use.
22 We will now proceed with the meeting and 23 I call upon Andy Campbell of the NRO Management to begin.
24 Please, Andy?
25 MR. CAMPBELL: Thank you, Mr. Chairman.
8 1 It's a pleasure to be here today. I'm Andy Campbell, 2 I'm the Deputy Director of the Division of Siting, 3 Licensing, and Environmental Analysis in the New 4 Reactors Office at NRC.
5 I want to just make a couple very quick 6 points and then, welcome everybody here. First, this 7 is the fourth and final ACRS Subcommittee meeting on 8 the Safety Evaluations with no open items for the Clinch 9 River ESP review.
10 Second, the first ESP for an SMR plant 11 design, that's what we've been reviewing and that's 12 what this is focused on. Project review has been 13 progressing consistent with the schedule, we're on or 14 ahead of schedule right now.
15 We're looking forward to a fruitful 16 dialogue today and then, with the full ACRS Committee 17 on December 5 of this year. So, with that, I'll turn 18 it back to you.
19 CHAIRMAN KIRCHNER: Thank you, Andy. Now, 20 we'll turn to -- Ray, are you going to start? Please 21 proceed.
22 MR. SCHIELE: Good morning. My name is Ray 23 Schiele, currently the Licensing Manager for the TVA 24 Clinch River early site permit application.
25 I have over 44 years in the nuclear
9 1 industry, including service in the United States Navy, 2 commercial plant operations and licensing, and most 3 recently, since 2016, Licensing Manager supporting the 4 Clinch River early site permit application.
5 Chairman Kirchner, before we get started, 6 TVA would again like to thank you and your Subcommittee 7 for the review of this application.
8 Acknowledgment and disclaimer. This 9 slide represents the acknowledgment of the relationship 10 between DOE and TVA. DOE funding is sharing in half 11 the project costs. DOE support is gratefully 12 appreciated by TVA. However, the work and view 13 expressed in the application and this presentation are 14 TVA's alone.
15 TVA's mission. TVA's mission is serving 16 the people of the Tennessee Valley. Currently, TVA 17 is partnering with 154 local power companies serving 18 more than nine million customers in parts of seven 19 states. They directly serve 56 large industries and 20 federal installations.
21 A quick review of the schedule and where 22 we are. This Gant chart is broken into three sections.
23 The top piece is the safety review. As you can see, 24 this meeting today is the fourth Subcommittee meeting, 25 with the full Committee scheduled on December 5. We
10 1 anticipate that FSER to be issued on or ahead of 2 schedule.
3 The next row is the status on the 4 Environmental Review. Again, the Environmental Review 5 is on or ahead of schedule, with the FEIS scheduled 6 to be issued on June of 2019.
7 Hearings. The -- in July of 2018, the ASLB 8 dismissed the last remaining admitted contention, 9 rejected the two new proposed contentions, and 10 terminated the contested hearing.
11 Considering the progress made in both the 12 Safety Review and Environmental Review, the Commission 13 mandatory hearing could be as early as late Fiscal Year 14 2019.
15 Quick review of a Plant Parameter Envelope.
16 The Plant Parameter Envelope, PPE, is an approach the 17 provides sufficient design detail to support the NRC 18 review of the early site permit application, while 19 allowing sufficient flexibility for technical 20 developments in new reactor technologies.
21 The actual design selected for the Clinch 22 River Site would be reviewed with a Combined License 23 Application to demonstrate that the design is bounded 24 by the PPE and differences would be reviewed for 25 acceptability in the Combined License Application.
11 1 The PPE that was developed in support of 2 the Clinch River Site early site permit application 3 is based on data from the four SMR designs under 4 evaluation by TVA. Those being: BWXT, NuScale, Holtec, 5 and Westinghouse.
6 PPE use considerations. The site 7 characteristics, which have been determined in the 8 analyses presented throughout the SSAR are those 9 necessary to establish findings required by 10 CFR 52 10 and 10 CFR 100, regarding suitability of the proposed 11 site.
12 Site-related design parameters are those 13 that are related to the design of an SMR that may be 14 constructed on the CRN Site in the future. In some 15 cases, it is necessary to assume values for certain 16 site-related design parameters in order to analyze the 17 associated site characteristics.
18 The values selected for the different 19 site-related design parameters represent the bounding 20 values and include engineering, safety, and 21 environmental conservatisms, as appropriate.
22 An outline of today's presentation.
23 Today's presentation will follow the following 24 sections. Section 11, Radioactive Waste Management, 25 will be presented by Alex Young. Section 2.3,
12 1 Meteorology, presented by Alex Young.
2 Section 17, Quality Assurance, presented 3 by Michelle Conner. And the last presentation, Section 4 2.4, Hydrology, will be presented by John Holcomb, 5 assisted by Stu Henry, and Hillol Guha.
6 Right now, I'd like to introduce Alex Young 7 to present Section 11. Alex?
8 MR. YOUNG: Thank you, Ray. My name's Alex 9 Young, Design Engineer for the SMR project for TVA.
10 I've been working on this project since September of 11 2014.
12 I'd like to start off talking about some 13 key NRC interactions associated with the Chapter 11 14 review. This piece consisted of one two-part audit.
15 The first part of that audit was conducted 16 at the Bechtel offices in Reston, Virginia in April 17 of 2017.
18 And the second part, taking place at the 19 TVA corporate offices in Knoxville, Tennessee. That 20 second part, later in April of 2017, consisted of a 21 site tour of the Clinch River Site and the surrounding 22 areas.
23 After the audit, TVA submitted a 24 supplemental letter in June of 2017, CNL-17-075, for 25 supplementary information regarding source term
13 1 development. Okay, next slide.
2 So, Chapter 11 is broken down into 3 Subsections 11.2, for liquid release, and 11.3, for 4 gaseous release. But for each of these subsections, 5 the release source terms were developed using the same 6 approach.
7 TVA utilized the Plant Parameter Envelope 8 approach using the guidance of NEI 10-01 to develop 9 the source terms. Each of the four vendors submitted 10 annual release, releases for individual reactor units, 11 and those were reviewed by TVA.
12 The site release annual activities were 13 developed by multiplying each vendor's values by their 14 respective number of units considered for the CRN Site.
15 Then, for both unit and site-basis values, 16 TVA developed composite tables utilizing the highest 17 annual activity for each isotope from any of the 18 vendors.
19 It was identified that some of the annual 20 activity in the composite table included excessive 21 conservatisms. We adjusted those isotopic activities.
22 The composite source terms were then 23 assessed for reasonableness by comparing to previously 24 approved source terms, scaled by reactor thermal power.
25 This comparison showed that the composite source term
14 1 was not unreasonable for use in the ESPA. Next slide.
2 So, for Section 11.2, the liquid rad 3 releases. To calculate the doses for those releases, 4 TVA implemented Regulatory Guidance 1.109 for the 5 exposure pathways considered and analytical methods 6 used.
7 LADTAP II was used to calculate the doses 8 with input parameters specific to the Clinch River Site.
9 TVA concluded that the effluent 10 concentrations are within the effluent concentration 11 limits of 10 CFR 20, Appendix B, Table 2, Column 2, 12 and that the doses are within the design objectives 13 of 10 CFR 50, Appendix I, and the environmental 14 standards of 40 CFR 190, and the limits of 10 CFR 15 20.1301. Next slide.
16 To calculate the doses for the gaseous 17 radioactive release, TVA implemented Regulatory 18 Guidance 1.109 and 1.111 for the exposure pathways 19 considered and analytical methods used. GASPAR II was 20 used to calculate the doses with input parameters 21 specific to the Clinch River Site.
22 TVA concluded that the effluent 23 concentrations are within the effluent concentration 24 limits of 10 CFR 20, Appendix B, Table 2, Column 1, 25 and that the doses are within the design objectives
15 1 of 10 CFR 50, Appendix India, and the environmental 2 standards of 40 C FR 190, and the limits of 10 CFR 3 20.1301. Thank you.
4 MR. SCHIELE: Chairman, this concludes the 5 presentation on Section 11. Do you want us to turn 6 it over to the staff?
7 CHAIRMAN KIRCHNER: So, could you just --
8 there are lots of numbers, lots of tables. When you 9 -- with your Plant Parameter Envelope, did you basically 10 conclude that, since these designs are LWR derivative, 11 essentially, it was a case of thermal power dominating 12 the source term, the liquid waste, and the gaseous 13 effluence?
14 MR. YOUNG: Sure. So, for that question, 15 the SMR designs and the information we were able to 16 review for the SMR designs currently are typical, 17 standard, LWR fuel that we see in our conventional 18 fleet.
19 And the rad waste management systems don't 20 provide greatly different methodologies or system 21 designs from what we see at our operational fleet.
22 So, we were able to justify that the general change 23 is going to be the fission products that come out of 24 the core, those driven primarily by core power.
25 CHAIRMAN KIRCHNER: Thank you.
16 1 MR. SCHIELE: So, that concludes our 2 presentation on Section 11. We'd like to turn over 3 to the staff now for their presentation on Section 11.
4 CHAIRMAN KIRCHNER: Quick moment while we 5 change out.
6 MR. CAMPBELL: So, presenting for the staff 7 will be Rich Clement and Mallecia Sutton. Please.
8 MS. SUTTON: Okay. Thank you. Good 9 morning. Again, my name is Mallecia Sutton. I'm one 10 of the Safety Project Managers for the Clinch River 11 early site permit application.
12 To my right, I have my cohort, Allen Fetter, 13 who is seated to the right of the table. Mr. Fetter 14 and I will be at the table for December 5, 2018, ACRS 15 full Committee meeting on all the Clinch River early 16 site permit evaluation covered by ACRS Subcommittee 17 meetings.
18 I've been with NRC since 2007, where I 19 started working as a Project Manager in the Office of 20 New Reactors. Prior to taking over as the Safety 21 Project Manager with Clinch River early site permit 22 review in January 2016, I was an Environmental Project 23 Manager for Bellefonte, Vogtle, Fermi, and Levy COL 24 reviews.
25 Today's ACRS Subcommittee meeting is the
17 1 fourth and final Subcommittee meeting for the Clinch 2 River application.
3 Today, NRC technical reviewers will be 4 presenting on the Safety Evaluations for Section 2.3, 5 Meteorology, 2.4, Hydrology, Radiological Management, 6 Section 11, and Quality Assurance, Section 17.5.
7 ACRS members will have an opportunity to 8 ask questions and provide comments between each 9 presentation for the sections discussed today.
10 In addition to staff's review of the TVA's 11 application, staff conducted four audits, one 12 inspection, one site visit, issued two RAIs comprising 13 of ten questions to the application in order to obtain 14 additional information to support NRC's findings.
15 The first technical staff you will hear 16 from today is Dr. Richard Clement. Today, he will be 17 presenting the review of the Site Safety Evaluation 18 Report, Section 11, Radiological Waste Management.
19 Dr. Richard Clement is a Senior Health 20 Physicist in the Division of Licensing, Siting, and 21 Environmental Analysis in the Office of New Reactors.
22 He has been involved in design certification, combined 23 license, and early site permit applications.
24 Rich has over 25 years of applied health 25 physics and operational experience, which includes
18 1 about 20 years of federal service.
2 At the NRC, Rich has also worked in the 3 Office of New Reactors, Nuclear Material Safety and 4 Safeguards, and Office of New Reactor and Regulation 5 as a technical reviewer. Now, I'll turn it over to 6 Rich.
7 MR. CLEMENT: Thank you, Mallecia. As she 8 mentioned, my name is Rich Clement, the Health Physics 9 Technical Reviewer for the Site Safety Analysis Report, 10 Chapter 11, Radioactive Waste Management, of the TVA 11 Clinch River early site permit application. Next 12 slide, please.
13 The staff's review involves source term 14 information on normal gaseous and liquid effluent 15 releases and the subsequent offsite doses described 16 in Section 11.2.3, Liquid Radioactive Releases, and 17 Section 11.3.3, Gaseous Radioactive Releases, of the 18 TVA Site Safety Analysis Report.
19 These sections also share review 20 interfaces with hydrology on the accidental liquid 21 source term and offsite dose from an postulated 22 accidental liquid release to the groundwater, evaluated 23 by staff in Section 2.4.13 of the Safety Evaluation, 24 and with meteorology on the atmospheric dispersion and 25 deposition factors for estimating an offsite dose from
19 1 gaseous effluent releases evaluated by the staff in 2 Section 2.3.5 of the Safety Evaluation that will be 3 presented to you later today. Next slide, please.
4 The staff participated in the 5 pre-application readiness assessment and acceptance 6 review of TVA's early site permit application.
7 The staff identified information that it 8 needed to understand development of the Plant Parameter 9 Envelope, or PPE, source terms and offsite doses from 10 normal effluent releases and the accident liquid source 11 term and offsite dose. As a result, TVA supplemented 12 its application.
13 The staff then conducted a face-to-face 14 audit with TVA to discuss and clarify the supplemental 15 information, which is described in the NRC Hydrology 16 and Health Physics Audit Report.
17 During the audit, the staff walked the 18 Clinch River Nuclear Site and visited the current 19 receptor locations for the assessment of offsite doses.
20 In addition, the staff conducted a virtual 21 audit of TVA's voluntary submittal involving 22 meteorology, which is described in the NRC Meteorology 23 and Health Physics Audit Report, also documented under 24 the ADAMS accession number shown. Next slide, please.
25 The staff reviewed TVA's PPE normal
20 1 effluent source term based on four small modular 2 reactor, or SMR, designs, which included: Generation 3 mPower, NuScale Power, Holtec, and Westinghouse.
4 The staff reviewed TVA's evaluation of 5 composite source terms in the surrogate plant used to 6 develop the normal PPE effluent source terms, performed 7 confirmatory calculations on unit and site effluent 8 release rates for each vendor, and reviewed adjustments 9 made to these effluent release rates and found them 10 reasonable.
11 The staff confirmed that the unity rule 12 applied in 10 CFR 20, Appendix B, Table 2, Columns 1 13 and 2, for the mixture of radionuclide concentrations 14 at the site boundary was met.
15 Based on the review, the staff found TVA's 16 methodology to develop the normal PPE effluent source 17 terms for use in calculating offsite doses reasonable.
18 Next slide, please.
19 CHAIRMAN KIRCHNER: May I stop you here?
20 MR. CLEMENT: Yes.
21 CHAIRMAN KIRCHNER: So, maybe this is a 22 place to ask about uncertainty in the application, 23 particularly the meteorology impacts on gaseous or 24 releases.
25 How -- let me see if I can -- how confident
21 1 are you in -- you did independent analyses of their 2 estimates, is that correct?
3 MR. CLEMENT: Confirmatory analysis. So, 4 we --
5 CHAIRMAN KIRCHNER: Confirmatory analysis.
6 MR. CLEMENT: -- reviewed the information 7 that was provided in the application, that was 8 supplemented. So, it was a listing of release rates 9 for each vendor. And if you follow the guidance in 10 NEI 10-01, you typically choose the highest release 11 rate for each vendor.
12 But due to the limited fuel development 13 and rad waste system designs, there were some 14 adjustments made for each vendor, based on the amount 15 of conservatism in information that was provided from 16 the vendor at that time.
17 CHAIRMAN KIRCHNER: Okay. And -- but when 18 you did your confirmatory analyses, how well did they 19 compare, in a general sense, with what the applicant 20 supplied?
21 MR. CLEMENT: The confirmatory analysis 22 that I did consists of taking the effluent release rates 23 from each vendor and comparing those release rates for 24 each respective vendor to see what the highest release 25 rate was determined.
22 1 And during that process, we found a couple 2 radionuclides where the highest release rates were not 3 selected and, therefore, they were corrected by TVA.
4 So, we took the release rates pretty much at face value, 5 because of the preliminary nature of the information.
6 And the confirmatory analysis looked at 7 across for each vendor, what was the release rate that 8 was selected for a composite unit plant and also, for 9 the site composite?
10 CHAIRMAN KIRCHNER: So, at the respective 11 boundaries, you have confidence that there is 12 conservatism in these calculations that you've 13 confirmed?
14 MR. CLEMENT: If you look at the release 15 rates across for each vendor, understanding that these 16 SMR designs have not yet been approved by the NRC, if 17 you look at the face value of those values, you can 18 see that there were several orders of magnitude 19 difference in the release rates.
20 And I think that was primarily driven by 21 the maturity of the source term information that was 22 available from the vendor at that time.
23 So, there was discussion in the application 24 to justify the adjustments that were made to those 25 release rates in order to come up with composite source
23 1 terms.
2 CHAIRMAN KIRCHNER: I'm asking, I guess, 3 a leading indirect question. I just want to probe how 4 much margin there is, how much confidence we have at 5 the exclusionary boundary and such for these releases, 6 in terms of 10 CFR 20 and the other appropriate 7 requirements.
8 MR. CLEMENT: I would say, given the 9 information that was provided on the docket and the 10 information that the staff reviewed, the COL action 11 item that is proposed at the end --
12 CHAIRMAN KIRCHNER: Right.
13 MR. CLEMENT: -- will pretty much be the 14 catchall for anything like that.
15 MEMBER CORRADINI: So, they've got to come 16 back with the chosen design and show that they're within 17 the bound?
18 MR. CLEMENT: Absolutely. And that's one 19 staff-identified COL action items that will ensure that 20 the PPE source term is bounded and the doses are bounded.
21 MEMBER BALLINGER: How do these release 22 rates compare with a typical large PWR in the fleet?
23 MR. CLEMENT: Well, there was --
24 MEMBER BALLINGER: Or, it should be, in your 25 case, BWR?
24 1 MR. CLEMENT: One of the issues that was 2 identified by TVA is that there was a scaling power 3 level ratio done with Public Service Enterprise Group, 4 PSEG, the ESP was approved by the NRC, included in one 5 of the designs, the advance boiling water reactor 6 design. So, obviously, the release rates are a little 7 bit different.
8 So, considering that not one plant would 9 contain the highest effluent release rates, there was 10 considerations made in adjusting those release rates.
11 But many of the release rates were scaled by thermal 12 power.
13 MEMBER BALLINGER: So, it's just strictly 14 scaled by thermal power?
15 MEMBER RICCARDELLA: So, I have a general 16 question on source terms, when it comes to small modular 17 reactors, when we're considering multiple units, and 18 maybe it's a little too general for this consideration.
19 But when you have multiple units, is the 20 source term simply the multiple of the source term per 21 reactor times the number of reactors?
22 Or is there some consideration of the risks 23 of single-reactor versus multiple-reactor events?
24 Where do you think NRO is going to come down on that 25 question?
25 1 MR. CLEMENT: For the source terms, 2 essentially, the unit release rates were multiplied 3 by the number of units for that design. So, it was 4 just considered multiplicative.
5 MEMBER RICCARDELLA: I understand that in 6 this particular case, but is that going to be a generic 7 approach to SMRs, the licensing of SMRs?
8 MR. CAMPBELL: This is Andy Campbell.
9 There's no reason to believe otherwise, for routine 10 radioactive waste, that you can't just scale it to the 11 overall thermal power for each unit and multiply those 12 by the number of units.
13 It's -- fission is going to produce the 14 waste, as well as the neutron flux, and with that, you're 15 just essentially dealing with fission products, as well 16 as neutron activation products, in the radioactive 17 waste.
18 It's not a very -- I mean, it's very 19 complicated, but it's not fundamentally different when 20 you have 12 units of the same type.
21 MEMBER RICCARDELLA: I guess that's for 22 considerations with the nuclear waste, but when we get 23 into considerations for severe accidents, it would seem 24 that it might be --
25 MR. CAMPBELL: This is not a severe accident
26 1 scenario.
2 MEMBER RICCARDELLA: I understand.
3 MR. CAMPBELL: That would be a whole --
4 MEMBER RICCARDELLA: Okay.
5 MR. CAMPBELL: -- different discussion.
6 MEMBER RICCARDELLA: I'll raise that 7 question in a different forum, then.
8 MR. CAMPBELL: Okay.
9 MR. CLEMENT: All right. Next slide, 10 please. For the dose evaluation, the staff verified 11 TVA's input parameters and assumptions in the exposure 12 pathway dose analysis, which included the normal PP 13 effluent source terms:
14 Internal exposure from ingestion of 15 contaminated milk, meat, and vegetables and inhalation 16 of airborne activity. And external exposure from 17 recreation activities, ground contamination, and 18 submersion in an airborne plume.
19 The staff confirmed that the exposure 20 pathway dose calculations to the maximally exposed 21 individual who is a member of the public to receive 22 the maximum possible dose meets the design objectives 23 in 10 CFR 50, Appendix I, the Environmental Protection 24 Agency's radiation standards in 40 CFR 190, and the 25 public dose limit in 10 CFR 20.
27 1 Because the reactor design that may be 2 constructed at the Clinch River Nuclear site is not 3 known at the early site permit stage, the staff 4 identified combined license, or a COL, action item 5 11.1-1 for the COL or construction permit applicant 6 to evaluate and justify any changes in the PPE source 7 term used for normal effluent releases and verify that 8 the calculated dose evaluated in the early site permit 9 is bounded. Next slide, please.
10 Based on the staff's review of TVA's early 11 site permit application, subject to the 12 staff-identified COL action item, the staff concludes 13 that the normal PPE effluent source terms and offsite 14 doses meet the applicable regulatory requirements and 15 that there is no undue risk to public health and safety.
16 Thank you. At this point, I will take any 17 question or comments you may have.
18 CHAIRMAN KIRCHNER: Okay. Members, any 19 questions at this point? Okay. Let's then proceed 20 on. Although, we have a break scheduled, I propose 21 we go next to TVA and Section 2.3 on Meteorology.
22 MR. SCHIELE: Mr. Chairman, TVA will 23 continue with Section 2.3 and we have some folks on 24 the phone to assist the conversation if necessary.
25 Presenting 2.3 will be Alex Young. Alex?
28 1 MR. YOUNG: Good morning. Thanks, Ray.
2 All right. Can we just confirm that those people are 3 available on the phone? I'm looking for Ken Westrick 4 and Marvin Morris. You guys on the line? Hearing 5 none, okay. We'll continue on with the presentation.
6 First, I'd like to note some key NRC 7 interactions related to SSAR Section 2.3, 8 Meteorological Information. There were two audits 9 that were conducted as a part of this.
10 The first being in May of 2017, included 11 with the environmental audit in the corporate offices 12 in Knoxville, Tennessee. And this included a site tour 13 and a tour of the former location of the met tower that 14 was on the site.
15 Also, in May of 2018, there was an audit 16 conducted via the TVA Electronic Reading Room that 17 supported an April 2018 supplemental letter to the 18 staff.
19 This supplemental letter compares the 20 results utilizing vector versus scalar average wind 21 directions, which we'll talk about in a little more 22 detail later in the presentation.
23 So, SSAR Section 2.3 is broken down into 24 five subsections, the first of which is Subsection 2.3.1 25 on Regional Climatology.
29 1 This section establishes the Clinch River 2 Site characteristics that are provided in Table 2.0-1 3 of the Site Safety Analysis Report. The information 4 presented in these first three slides presents those 5 site characteristics provided in Table 2.0-1.
6 TVA utilized a variety of data sources, 7 including the National Oceanic and Atmospheric 8 Administration, the National Climatological Data 9 Center Storm Events Database and Local Climatological 10 Data Summaries, the National Weather Service records, 11 and observations from TVA Sequoyah and Watts Bar Nuclear 12 Plants.
13 The winter precipitation events presented 14 here were determined utilizing a variety of sources, 15 as suggested in Interim Staff Guidance Number 7, 16 including American Society of Civil Engineers Standard 17 Number 7-05, National Weather Service data, and 18 Hydrometeorological Report Number 53.
19 The maximum rainfall rate provided is based 20 on Hydrometeorological Report Number 52.
21 The basic wind speed is provided based on 22 the American Society of Civil Engineers Standard Number 23 7-05, with historical maximum based on local 24 climatological data.
25 And hurricane wind speeds were determined
30 1 utilizing the speed contours in Regulatory Guidance 2 1.221 and NUREG-7005. Next slide.
3 Presented here are the tornado-related 4 site characteristics presented in Table 2.0-1. These 5 were determined using Reg Guide 1.76. Next slide.
6 Here, we've presented the ambient air 7 temperatures presented in SSAR Table 2.0-1 that were 8 determined using local data from the National Oceanic 9 and Atmospheric Administration and utilizing ASHRAE 10 equations and calculations. Next slide. All right.
11 SSAR Subsection 2.3.2, on Local 12 Meteorology, compared recent and historical local and 13 regional data.
14 It was identified that topography around 15 the site strongly influences the local climate and 16 established the predominant valley-ridge access shown 17 in the figure.
18 The predominant up-valley/down-valley 19 flow depicted is readily apparently at all three 20 meteorological towers shown in the figure.
21 Comparisons of temperature, 22 precipitation, and moisture data confirmed that the 23 Clinch River Site conditions are consistent with 24 regional conditions. Next slide. Okay.
25 SSAR Subsection 2.3.3 described the onsite
31 1 meteorological monitoring program utilized to collect 2 onsite data for use in the Clinch River early site permit 3 application.
4 The onsite meteorological measurement 5 program was conducted utilizing three different 6 meteorological towers, and their locations, as shown 7 in the previous slide.
8 MEMBER CORRADINI: Can I ask a general 9 question?
10 MR. YOUNG: Sure.
11 MEMBER CORRADINI: This is too detailed for 12 me, so I'm going to take you back to something broader.
13 So, in these data, this is recent data or do you look 14 at it historically?
15 Where I'm going with that is, for Clinch 16 River, in the prior application for the fast reactor, 17 they probably had to do a similar thing. Did you look 18 at the delta change in the meteorological data from 19 the 1970s to now?
20 MR. YOUNG: Yes, we did. Well, as we 21 continue on the presentation, I'll describe some of 22 the data we used and I'll make sure to note the 23 comparisons --
24 MEMBER CORRADINI: Thank you.
25 MR. YOUNG: -- that we did.
32 1 MEMBER CORRADINI: Okay, thank you very 2 much.
3 MR. YOUNG: So, the onsite meteorological 4 measurement program was conducted using three 5 meteorological towers and their locations, as shown 6 on the previous figure.
7 This figure shows the latest tower, the 8 primary meteorological tower that was onsite at one 9 point in time. The primary meteorological tower was 10 a 110-meter tower originally constructed for the Clinch 11 River Breeder Reactor Project.
12 This tower was then reactivated from 2011 13 to 2013, at the ten-meter and 60-meter elevations, to 14 collect pre-application data for the Clinch River early 15 site permit application.
16 The supplemental tower was a ten-meter 17 tower utilized during the Clinch River Breeder Reactor 18 Project. And the temporary tower was a 61-meter tower 19 utilized to collect the pre-application data for the 20 Clinch River Breeder Reactor Project.
21 You asked specifically about the 22 comparisons of some of the historical data versus modern 23 data. And as we've described, there's multiple towers 24 and they were used at different times.
25 On the previous slide, on 2.3.2, we
33 1 mentioned that we see very similar influences for all 2 three met towers, which were at different times, for 3 similar wind conditions. All right.
4 Continuing on, 2.3.3, Onsite 5 Meteorological Measurement Program. Data collected 6 for the early site permit application satisfied the 7 guidance provided in Regulatory Guide 1.23.
8 However, the ANSI Standard 3.11-2005 is 9 a reference of Regulatory Guide 1.23, and it states 10 that the transport wind direction for straight-line 11 Gaussian models should be based on the scalar mean wind 12 direction.
13 TVA has evaluated the use of both vector 14 and scalar wind direction for the Clinch River Site.
15 There were several differences between the approaches, 16 with some sectors identifying larger atmospheric 17 dispersion values and others identifying smaller 18 values.
19 TVA considered both the Chapter 15 and 20 Chapter 11 dose consequences utilizing both vector and 21 scalar wind direction atmospheric dispersion values 22 and concluded that the vector wind direction was more 23 conservative and was utilized as the basis for the 24 following Subsections SSAR 2.3.4, 2.3.5, and their 25 associated Chapter 15 and Chapter 11 analyses.
34 1 CHAIRMAN KIRCHNER: So, Alex, for the 2 record, for the public, could you explain why vector 3 was bounding for Chapter 15 and 11, versus the scalar 4 approach?
5 MR. YOUNG: Absolutely. So, it's slightly 6 different for Chapter 15 versus Chapter 11. The 7 Chapter 15 analysis conducted for the ESPA is based 8 on the single limiting sector and single limiting 9 values.
10 So, when we compared vector versus scalar 11 results for the Chapter 15 analysis, we noticed that 12 both of them are driven by the same sector and that 13 the vector wind direction was a more conservative value 14 for that same wind direction sector.
15 For the Chapter 11 piece, which includes 16 a multitude of sectors, multitude of X/Q values and 17 D/Q values, we ran a sensitivity case of dose analyses 18 utilizing -- one case utilizing the vector, one case 19 utilizing the scalar results, and the vector results 20 showed to have more limiting dose consequences.
21 CHAIRMAN KIRCHNER: And physically, can you 22 explain for the record why that is so?
23 MR. YOUNG: Physically, it comes down to 24 vector averaging and the mathematics. It's noted that 25 we don't necessary see this for all cases, this was
35 1 a case specific to the data we analyzed and for the 2 Clinch River Site. So, for other sites, that may not 3 be the case.
4 CHAIRMAN KIRCHNER: But again, I'm probing 5 a little further, physically, why is it so that the 6 vector approach gives you a more bounding conservative 7 versus the scalar? Is it just the plume dispersion?
8 MR. YOUNG: Yes. So, it's based on the 9 X/Qs, D/Qs. So, those results that we get --
10 CHAIRMAN KIRCHNER: You're talking Greek, 11 could you --
12 MR. YOUNG: Okay.
13 CHAIRMAN KIRCHNER: -- get out of the 14 physics space --
15 MR. YOUNG: Sure.
16 CHAIRMAN KIRCHNER: -- and say what's 17 happening?
18 MR. YOUNG: So, X/Qs were -- we think of 19 it as a smoke cloud, you're releasing contamination.
20 It propagates through the air and lands and disperses.
21 So, we found that utilizing the vector results, there 22 was less of that dispersion. It was more concentrated, 23 therefore, there was more absorption in dose. All 24 right.
25 Moving on to the next slide, Section 2.3.4.
36 1 So, SSAR Subsection 2.3.4 addresses the development 2 of the short-term diffusion estimates utilized for the 3 accident evaluations in Chapter 15 of the SSAR.
4 These atmospheric dispersion calculations 5 were performed utilizing the PAVAN code and met the 6 requirements of Regulatory Guidance 1.145 and 1.23.
7 These calculations utilized the meteorological data 8 from June 1, 2011 through May 31, 2013.
9 TVA also made conservative assumptions 10 considering the use of the Plant Parameter Envelope 11 and gave no credit for building wake effects and assumed 12 a ground level release.
13 As depicted in the figure, atmospheric 14 dispersion values for the exclusionary boundary were 15 calculated at an 1,100-foot distance from the effluent 16 release boundary, for any proposed reactor location 17 onsite.
18 The low population zone atmospheric 19 dispersion values were calculated at a one-mile 20 distance from the site center point. Next slide.
21 SSAR Subsection 2.3.5 addresses the 22 development of long-term diffusion estimates utilized 23 for the normal release evaluations in Chapter 11 of 24 the SSAR.
25 These atmospheric dispersion calculations
37 1 were performed using the XOQDOQ code. These 2 calculations utilized the same meteorological data from 3 that June 1, 2011 through May 31, 2013 period, and again, 4 gave no credit for building wake effects and assumed 5 ground level releases.
6 Values were calculated for each of the 16 7 wind direction sectors at different distances out to 8 50 miles and for the nearest residents, vegetable 9 garden, and beef animal in each sector.
10 This figure depicts the sensitive 11 receptors identified within the surrounding area.
12 Next slide, please. Okay.
13 SSAR 2.3.5, Complex Terrain. As mentioned 14 previously, the topography at the site has a strong 15 influence on the local climate.
16 To evaluate the complex terrain 17 surrounding the site, TVA made a comparison of the 18 results with a variable trajectory model called 19 CALPUFF. This model utilized similar data and 20 assumptions as the other calculations.
21 We used the same meteorological data from 22 the June 1, 2011 through May 31, 2013 period and we 23 assumed ground level releases and gave no credit for 24 building wake effects.
25 The conclusion of this evaluation was that
38 1 the XOQDOQ model, previously described in the previous 2 slide, was bounding for the assessment of long-term 3 diffusion estimates.
4 CHAIRMAN KIRCHNER: When you make the 5 assumption of ground level release, then in effect, 6 that's like a release when you are having an inversion?
7 MR. YOUNG: So, that's the -- opposed to 8 having a stack that would release it higher in the 9 atmosphere versus low. We reduce it lower, which 10 limits the amount of dispersion that there is the 11 potential to happen as it approaches that boundary.
12 CHAIRMAN KIRCHNER: I'm just thinking, I've 13 driven on I-40, south of the site, under conditions 14 were essentially it was like an inversion, it was heavy 15 fog, cloud cover, very low sitting in those valleys.
16 Okay. So, this -- by making a ground level release 17 assumption, you are probably in effect --
18 MR. YOUNG: You would have more of that 19 effect --
20 CHAIRMAN KIRCHNER: -- simulating that --
21 MR. YOUNG: -- opposed to a greater 22 dispersion.
23 CHAIRMAN KIRCHNER: -- condition for the 24 release?
25 MR. YOUNG: Yes. You would have more of
39 1 that type of effect, opposed to a greater dispersion 2 at higher elevations in the atmosphere.
3 CHAIRMAN KIRCHNER: So, let me ask Walt's 4 question differently, because, again, this is an area 5 that I know you follow the guides, but I'm curious, 6 is the X/Q -- let me not do that.
7 Is the way in which you treat the 8 meteorology here regionally-dependent, so that if I 9 were to look at this in Illinois or Wisconsin or 10 Minnesota, it would be a different set of X/Qs? Or 11 are you looking for a bounding X/Q regardless of site, 12 in terms of the guide?
13 That's what I was kind of curious about, 14 kind of going with his question about hills and valleys 15 here, catching it differently, and you having a 16 different terrain.
17 MR. YOUNG: Sure. So, because of the 18 topography around our site, this is very specific to 19 the Clinch River Site --
20 MEMBER CORRADINI: Okay.
21 MR. YOUNG: -- based on the topography and 22 how winds flow through the area.
23 MEMBER CORRADINI: Okay.
24 MR. YOUNG: And that concludes the 25 presentation on 2.3.
40 1 CHAIRMAN KIRCHNER: Okay, thank you. Let 2 us turn to the NRC staff at this point.
3 MR. CAMPBELL: This is Andy Campbell, again.
4 Presenting for the NRC is Kevin Quinlan, for the 5 meteorology, and Mallecia Sutton.
6 CHAIRMAN KIRCHNER: Okay.
7 MR. CAMPBELL: And I will add, stepping into 8 an area that I don't know much about, all X/Q, D/Qs 9 are site-specific. There really are no generic ones, 10 you really have to look at each and every site to make 11 that determination.
12 MEMBER CORRADINI: Andy, since you brought 13 that up, how local do you get, in terms of distance?
14 You go out ten -- you look at some sort of averaging 15 over, like, a ten-mile radius?
16 MR. CAMPBELL: Now, you're talking in 17 Kevin's talk, so I'm going to --
18 MEMBER CORRADINI: Well, that's all right.
19 He can wait, when the time comes, but I was kind of 20 curious. That's fine.
21 MR. CAMPBELL: It'll look at a variety of 22 differences, and he's nodding his head yes, so I 23 answered that correctly.
24 (Laughter.)
25 MR. CAMPBELL: And that's the extent of my
41 1 knowledge.
2 CHAIRMAN KIRCHNER: Okay. Proceed.
3 MS. SUTTON: Kevin Quinlan graduated from 4 Millersville University of Pennsylvania in 2006 with 5 a bachelor's of science in meteorology. He then went 6 on to earn his masters of science degree from the 7 University of Alabama in Huntsville, atmospheric 8 science.
9 Mr. Quinlan has been working in the Office 10 of New Reactors since July 2008. He is or has been 11 the lead NRC Meteorology Reviewer on 12 new reactor 12 applications and design reviewed by the NRC. Now, I'll 13 turn the presentation over to Kevin.
14 MR. QUINLAN: Good morning. My name is 15 Kevin Quinlan and I'm a meteorologist in the Office 16 of New Reactors, Division of Licensing, Siting, and 17 Environmental Analysis.
18 Section 2.3, Meteorology, discusses the 19 site-specific information related to regional 20 climatology, local meteorology, the onsite 21 meteorological measurements program, short-term 22 atmospheric dispersion estimates for accidental 23 releases, and long-term atmospheric dispersion 24 estimates for routine releases.
25 I'd like to note that this section included
42 1 technical input from other staff meteorologists, 2 notably Mike Mazaika, Jason White, and the 3 Meteorological Team Leader, Brad Harvey. Next slide, 4 please.
5 Section 2.3.1, Regional Climatology, 6 provides information related to the regional 7 climatology that could potentially influence the design 8 and operating basis of safety and non-safety-related 9 structures, systems, and components.
10 Section 2.3.1 is where most of the 11 meteorological site characteristics are developed and 12 reviewed.
13 Staff performed a review and analysis of 14 the following site characteristics: the tornado and 15 hurricane wind speeds and associated missiles; the 16 100-year return period wind speed for three-second 17 gusts; the maximum winter precipitation; ambient air 18 temperature and humidity.
19 And staff concluded that the 20 identification and consideration of the climatic site 21 characteristics are acceptable at the Clinch River 22 Site. Next slide, please.
23 Section 2.3.2 discusses the local 24 meteorology in the area surrounding the site. This 25 section provides summaries of local meteorological
43 1 conditions, an assessment of the potential influences 2 of the plant on the local meteorological conditions, 3 and a topographical description of the site and its 4 surroundings.
5 Staff reviewed the Clinch River analysis 6 of the onsite wind speed and direction summaries, 7 atmospheric stability, and ambient air temperature and 8 humidity.
9 Staff also confirmed meteorological 10 information related to precipitation, fog, and 11 potential changes in air quality near the site. Staff 12 reviewed and verified that the local meteorological 13 data provided by TVA are representative of the site 14 area as impacted by the local topography.
15 Section 2.3.3 discusses the onsite 16 meteorological measurements program, in support of the 17 early site permit application. NRC staff visited the 18 site and reviewed the onsite meteorological 19 measurements program during an environmental site audit 20 conducted in May of 2017.
21 The audit topics were related to the 22 meteorological monitoring. They included location and 23 exposure of previously sited meteorological 24 instrumentation and the tower, instrument maintenance, 25 and the data quality assurance program.
44 1 NRC staff completed a quality assurance 2 review of the onsite meteorological database submitted 3 by TVA as part of the early site permit application 4 and staff confirmed that the TVA meteorological tower 5 conformed to Regulatory Guide 1.23 criteria for siting 6 of the tower in relation to the proposed Clinch River 7 Site.
8 One concern that the staff had with the 9 onsite meteorological measurements program, and this 10 was just previously discussed in TVA's presentation 11 was related to TVA's use of the vector average wind 12 direction and scalar average wind speed data as input 13 to the atmospheric dispersion models.
14 TVA chose an alternative method to the best 15 practice cited in Regulatory Guide 1.23 and ANSI 16 Standard 3.11-2005, Determining Meteorological 17 Information at Nuclear Facilities, which states that 18 the transport wind direction for straight-line Gaussian 19 models should be based on the scalar mean or unit vector 20 wind direction.
21 TVA voluntarily provided a submittal that 22 evaluated the effects of using vector average wind 23 directions rather than the suggested scalar average 24 wind directions for the atmospheric dispersion 25 estimates.
45 1 The analysis showed that the dose modeling 2 results were bounding, based on the average of the 3 vector average wind directions, as provided in the SSAR.
4 However, TVA acknowledged the atmospheric 5 dispersion and deposition factors for routine 6 radiological releases were greater in some directions 7 and lower in others, when compared to using the scalar 8 average wind directions. Okay.
9 MEMBER SUNSERI: I have a question about 10 that.
11 MR. QUINLAN: Sure.
12 MEMBER SUNSERI: So, is there a suggestion 13 there that the scalar method that's referenced in the 14 Reg Guide is non-conservative, or not as conservative, 15 as using the vector?
16 MR. QUINLAN: It likely varies 17 site-by-site. However, the ANSI 3.11 standard, as 18 referenced in the Regulatory Guide, suggests the use 19 of the scalar average wind direction, just as a best 20 practice.
21 However, in this case, some areas were --
22 some directions were a little more conservative or a 23 little higher and some were lower.
24 MEMBER SUNSERI: So, based on the TVA 25 experience, would you anticipate updated the Regulatory
46 1 Guidance?
2 MR. QUINLAN: When we get to updating the 3 guidance, it may be an area to take an additional, a 4 closer look at, and maybe compare some other sites as 5 well.
6 MEMBER SUNSERI: Okay, thank you.
7 MR. QUINLAN: Based on the aforementioned 8 analysis, TVA concluded that for normal and accident 9 gaseous release dose assessments, the existing dose 10 analysis in the SSAR is conservative and remains the 11 basis for the ESP application.
12 NRC staff conducted an audit of the 13 submittal and agreed with the applicant's conclusion 14 that the SSAR dose analysis is bounding.
15 The staff concluded that the onsite 16 meteorological monitoring system provides adequate 17 data to represent the onsite meteorological conditions 18 at the Clinch River Site during the time frame in which 19 it was collected. Next slide, please.
20 The staff identified and has proposed three 21 COL action items related to the onsite meteorological 22 measurements program.
23 COL Action 2.3-2 states that an applicant 24 referencing this early site permit should demonstrate 25 the onsite meteorological measurement program
47 1 continues to meet the guidance provided in Regulatory 2 Guide 1.23. This was necessary, since the system that 3 recorded the meteorological data for the early site 4 permit application has since been removed.
5 COL Actions items 2.3-3 and 2.3-4 are 6 related to the collection and use of vector and scalar 7 average wind data averaging for COL or a CP referencing 8 this early site permit.
9 MEMBER CORRADINI: So, can you -- 2.3-3, 10 so the way I read that is, they've got to go back and 11 check to make sure which one is bounding? That's how 12 I read that. Am I misreading it?
13 MR. QUINLAN: I believe the intent of this 14 one was, because we're granting a finality on the X/Q 15 values and the onsite data that was collected for use 16 in the early site permit, but the tower and the system 17 that recorded the meteorology data has since been 18 removed, when they come in for a COL or CP and they 19 build a new tower, that it remains the same as what 20 the early site permit assumed. And if not, then a 21 comparison can be --
22 MEMBER CORRADINI: I think you're answering 23 2.3-2, I was asking about 2.3-3. I think I understand 24 the first one.
25 2.3-3 leads me to believe that they're
48 1 going to have to come back, whoever -- if they decide 2 to go forward and if they pick one of the four, that 3 design is going to have to compare scalar to vector 4 and pick the bounding of the two. Am I misunderstanding 5 that?
6 MR. QUINLAN: It says that it should verify 7 whether the operational phase of the onsite 8 meteorological measurement program will include wind 9 data averaging on the basis of scalar or vector 10 averages.
11 So, I think they need to say at that time 12 which program they're going to be using, or which 13 averaging type they'll be using going forward.
14 MEMBER CORRADINI: And either one would then 15 be -- I'm still back to Matt's question about either 16 one would be acceptable. But in this case, because 17 of this locale and this weather, it turns out vector 18 averaging was more bounding?
19 MR. QUINLAN: In this case, yes.
20 MEMBER CORRADINI: Okay.
21 MEMBER SUNSERI: So, you would think that 22 that would be more specific, since the regulatory best 23 practice is to use the scalar, that the COL item should 24 reference using vector.
25 MR. QUINLAN: Well, it would be up to them
49 1 to -- if they change the averaging type, then they could 2 take a departure from the early site permit.
3 So, it's really up to TVA at that point 4 to decide which they would want to use. If it's 5 inconsistent with the ESP, then they could always take 6 a departure. However, it is up to them for how they 7 set up their system.
8 MEMBER CORRADINI: So, this is kind of in 9 the weeds, so let me say it back to you so I get it.
10 I think I get it now. Your point is, they can do 11 either.
12 If they choose to do vector, they're in 13 compliance and consistent with the ESP. If they choose 14 to do scalar, they've got to essentially say why and 15 ask for an exemption.
16 MR. QUINLAN: I believe that's correct.
17 MEMBER CORRADINI: Okay, I got it.
18 MR. QUINLAN: Okay. Section 2.3.4 relates 19 to the short-term atmospheric dispersion estimates used 20 to determine the amount of airborne radioactive 21 materials expected to reach a specific location during 22 an accident situation.
23 These atmospheric dispersion factors, or 24 X/Q values, estimate the relevant concentrations at 25 the exclusion area boundary, the EAB, and at the outer
50 1 boundary of the low population zone, or LPZ, for 2 postulated design-basis accidental radioactive 3 airborne releases.
4 As part of the review, staff performed an 5 independent verification of the applicant's accident 6 diffusion estimates.
7 Staff created a joint frequency 8 distribution from wind speed, wind direction, and 9 atmospheric stability data collected as part of the 10 onsite meteorological data, and used for input to the 11 PAVAN atmospheric dispersion computer model.
12 Staff then executed the model and generated 13 offsite X/Q values for all sectors along the uniform 14 analytical EAB and LPZ boundaries. Next slide, please.
15 As described in SSAR Section 2.3.4.2, the 16 nuclear island effluent release boundary, or the small 17 green and blue circles on the figure on the screen, 18 are used to conservatively enclose all possible release 19 points for the selected reactor technologies.
20 The distance from the outer edge of the 21 power block area to the exclusion area boundary is 335 22 meters, or 1,100 feet, as shown in the figure on the 23 slide.
24 To account for the potential of multiple 25 units on the site, nuclear islands are positioned at
51 1 multiple locations within the power block, with 2 associated effluent release boundaries and exclusion 3 area boundaries as shown in the figure.
4 A circular analytical EAB is established 5 1,100 feet from the effluent release boundary, as 6 denoted by the yellow circles.
7 All of the potential nuclear island sites 8 are bounded by the red ellipse that encompasses all 9 of the analytical effluent release boundaries and is 10 completely contained within the Clinch River Site.
11 Since the distance from the outer edge of 12 the power block to the effluent release boundary is 13 less than the actual distance from the nuclear island 14 to the EAB, and will result in higher or more 15 conservative X/Q values, the NRC staff considers the 16 assumptions in the dispersion analysis to be 17 reasonable.
18 Through this confirmatory analysis, the 19 staff found the applicant's EAB and LPZ site 20 characteristic X/Q values to be acceptable.
21 CHAIRMAN KIRCHNER: Let me explore with you, 22 yes, they were acceptable, so I'm not looking to change 23 what the requirements are.
24 I wanted to explore more, how close were 25 their X/Q values to yours, after you did your
52 1 confirmatory analysis? And what I'm looking at is 2 uncertainty sensitivity, as might impact the analysis 3 of Chapter 15 analyses.
4 MR. QUINLAN: Sure. I'm opening up the SER 5 to see if we provided an exact number for how close 6 they were. But we did use the same two-year onsite 7 meteorological dataset as TVA. And we created our own 8 joint frequency distribution, used the same distances 9 --
10 CHAIRMAN KIRCHNER: Right.
11 MR. QUINLAN: -- for each direction. So, 12 they were very close. I don't have an exact number 13 for you, but usually, if it's any more than a couple 14 of percent, maybe two to four percent difference, then 15 we start to explore a reason why we have a larger 16 difference.
17 In this case, I remember the results being 18 very close, either right on, the exact same, or just 19 within one or two percent.
20 CHAIRMAN KIRCHNER: Okay, thank you.
21 MR. QUINLAN: You're welcome.
22 MR. CAMPBELL: This is Andy Campbell.
23 Just, if you want to pursue that, we can point you to 24 the specific area of the SER where the numbers are 25 compared.
53 1 CHAIRMAN KIRCHNER: To the point, Andy, I 2 was trying to integrate that. So, yes, I know where 3 the numbers are in the SER, I'm trying to really have 4 a feeling of margin and confidence when it comes to 5 issues like the emergency planning topic. So, that's 6 why I'm pushing on this.
7 I would hope that a slight change in the 8 weather wouldn't put them over any of the requirements 9 that have to be met here with a much smaller emergency 10 planning zone.
11 That's the one in particularly I'm looking 12 at, because, in effect here, we're ahead of the 13 rulemaking, with what the applicant is proposing, so 14 I'm pushing to understand and have confidence that the 15 analyses that had been done and the confirmatory 16 analyses done across the board by the staff show that 17 we have reasonable confidence on this official issue.
18 MR. CAMPBELL: And that there's sufficient 19 margin --
20 CHAIRMAN KIRCHNER: Yes.
21 MR. CAMPBELL: -- between these analyses 22 and what the site boundary could be. And I think Kevin 23 can speak to the conservatisms that are inherent in 24 these types of analyses, in terms of that margin.
25 CHAIRMAN KIRCHNER: Thank you.
54 1 MR. QUINLAN: If there are no further 2 questions on this slide, I can -- okay to move on?
3 CHAIRMAN KIRCHNER: Yes.
4 MR. QUINLAN: Okay. Section 2.3.5 relates 5 to the long-term dispersion estimates that are used 6 to determine the amount of airborne radioactive 7 materials expected to reach a specific location during 8 normal operations.
9 These dispersion estimates address the 10 requirement concerning atmospheric dispersion and dry 11 deposition estimates for routine releases of radiologic 12 effluents to the atmosphere.
13 For the review, the staff performed an 14 independent verification of the applicant's routine 15 release diffusion estimates.
16 As with Section 2.3.4, discussed 17 previously, staff created a joint frequency 18 distribution from the onsite meteorological data for 19 use as part of the input to the XOQDOQ atmospheric 20 dispersion computer model.
21 Staff then executed the XOQDOQ computer 22 model and generated X/Q and D/Q values for receptors 23 of interest. Based on the XOQDOQ results, the staff 24 concluded that representative atmospheric dispersion 25 and deposition conditions have been calculated for the
55 1 receptors of interest.
2 In conclusion, all regulatory requirements 3 for Section 2.3, Meteorology, have been satisfied.
4 In this section, we have no open items and we do have 5 three confirmatory items, which are expected to be 6 closed at the next revision of the SSAR. And I'll take 7 any questions that you may have.
8 CHAIRMAN KIRCHNER: Okay, thank you.
9 Members?
10 MEMBER BROWN: I've got one, I'm not a 11 meteorology person either, but back on Slide 3 -- I'm 12 going to get it right sooner or later if I say it often 13 enough -- you noted you did your review of the original 14 climatology.
15 MR. QUINLAN: Yes.
16 MEMBER BROWN: And in an earlier discussion, 17 we talked and it was mentioned that Clinch River Breeder 18 Reactor also had a similar type of analysis that was 19 done. And that was, what?, 30?, how many years ago?
20 MR. QUINLAN: Yes, I believe mid-1970s.
21 MEMBER BROWN: Mid-1970s?
22 MR. QUINLAN: Forty years ago.
23 MEMBER BROWN: Thirty-five, 40 years ago?
24 Okay. Was there any comparison or look back and see 25 what the results were there? Were these more severe
56 1 than they were then?
2 MR. QUINLAN: There was a comparison of the 3 wind speed and wind directions for the dataset that 4 they collected for an early site permit from 2011 to 5 2013, compared to the 1970s data, there were two 6 separate datasets collected in the 1970s.
7 There was a comparison in the SSAR, that 8 compared the -- it was a wind rose as well as, I believe, 9 wind speeds and wind directions. So, there was a 10 comparison done.
11 The staff, we compared the data that they 12 provided for the early site permit, we did our own 13 internal analysis and quality check of the data, and 14 compared it against what they provided in the SSAR, 15 to make sure that we were arriving at the same results.
16 We did not independently do a verification of the 1970s 17 data, but --
18 MEMBER BROWN: Well, I wasn't looking for 19 that --
20 MR. QUINLAN: Sure.
21 MEMBER BROWN: -- it just was the end result.
22 I mean, you confirmed that their characteristic values 23 now were appropriately derived from the Reg Guides.
24 They were also probably appropriately 25 derived from whatever the Regulatory Guides were at
57 1 that time.
2 MR. QUINLAN: Yes.
3 MEMBER BROWN: And I'm just wondering, were 4 the conditions more severe now, predicted to be more 5 severe now than they were then? In other words, was 6 there a change in the severity of the wind speeds, 7 100-year return, et cetera?
8 MR. QUINLAN: You always expect at least 9 a small variation from year-to-year.
10 MEMBER BROWN: I don't --
11 MR. QUINLAN: But the --
12 MEMBER BROWN: -- disagree with that.
13 MR. QUINLAN: -- comparisons were very 14 close, between the more recent dataset and the 1970s 15 dataset.
16 MEMBER BROWN: Okay. That's -- thank you.
17 MR. QUINLAN: Yes, you're welcome.
18 CHAIRMAN KIRCHNER: Okay. Thank you.
19 MR. QUINLAN: Thank you.
20 CHAIRMAN KIRCHNER: It seems that, 21 according to the agenda, we are at lunch.
22 (Laughter.)
23 CHAIRMAN KIRCHNER: So, I'm going to try 24 and reorganize here a bit. What I would propose is 25 to take a 15 minute break at this juncture.
58 1 But I want to check with both the applicant 2 and the staff, whether we have the necessary people 3 on-hand if we take up Quality Assurance and Hydrology 4 after the break.
5 MR. SCHIELE: TVA can support it.
6 CHAIRMAN KIRCHNER: Okay.
7 MR. CAMPBELL: And the staff can --
8 MS. SUTTON: The staff, yes.
9 MR. CAMPBELL: -- support that as well.
10 CHAIRMAN KIRCHNER: Excellent, okay.
11 Then, we will recess for 15 minutes. Let's use the 12 clock up there and return at five minutes of 10:00.
13 (Whereupon, the above-entitled matter went 14 off the record at 9:38 a.m. and resumed at 9:54 a.m.)
15 CHAIRMAN KIRCHNER: Let's reconvene. Let 16 me, for the record, mention that Dennis Bley, a member, 17 is on the phone line. And with that, we're going to 18 turn to Quality Assurance. Ray, would you proceed?
19 MR. SCHIELE: Thank you, Mr. Chairman. I'd 20 like to introduce Michelle Conner, who will be 21 presenting SSAR Section 17, Quality Assurance.
22 MS. CONNER: Thank you, Ray. My name is 23 Michelle Conner, I'm the TVA SMR Senior Project Manager 24 for Operations, Training, and Programs, with 19 years 25 of experience in nuclear regulatory affairs and
59 1 operations. I held an NRC license as a Reactor Operator 2 and a Senior Reactor Operator for 12 of those years.
3 This presentation is for the ESPA Site 4 Safety Analysis Report Section 17.5, Quality Assurance 5 Program Description.
6 We'll go through the chronology, the Clinch 7 River ESPA activities, the program description, quality 8 assurance implementation, and then, a conclusion.
9 So, first, the chronology. The ESPA Rev 10 1 was submitted to the NRC in December of 2017. The 11 NRC issued an RAI on QA on March 9, 2018. TVA provided 12 our RAI response on April 9 and the NRC Quality Assurance 13 Inspection was on April 16-20.
14 TVA issued the NQAP Rev 36 subsequent to 15 that inspection on May 8, 2018. NRC issued the QA 16 Inspection Report on June 1, 2018. And we'll talk about 17 each of those activities in more detail.
18 So, first, the TVA Nuclear Quality 19 Assurance Plan Description. The TVA NQAP is the top 20 level document that defines the Quality Assurance 21 policy and assigns major functional responsibilities.
22 Section 17.5 of the application provides 23 a summary of the TVA Clinch River QA Plan attributes.
24 It is a separately controlled document and is included 25 in Part 8 of the ESPA.
60 1 The activities performed during the ESPA 2 development for Clinch River using the TVA Fleet Nuclear 3 Quality Assurance Plan. The NQAP is an NRC approved 4 10 CFR 50, Appendix B, Quality Assurance Plan that is 5 used by the three operating sites for TVA.
6 The TVA NQAP was based on an early set of 7 standards endorsed by the NRC. The early standards 8 were the foundation of the subsequent development of 9 the NQA-1 standards, which are endorsed by Reg Guide 10 1.28 Rev 4, the Quality Assurance Program Requirements 11 for Design and Construction.
12 The NRC issued an RAI to TVA to clarify 13 conformance to SRP 17.5 Rev 1, and to provide 14 clarification of that conformance to proposed 15 alternatives to some of the 17.5 acceptance criteria 16 and commitments.
17 So, TVA developed a conformance matrix that 18 provided those requirements with a TVA QA Plan. Where 19 conformance was not provided, commitments were added 20 to the TVA QA Plan and where the existing TVA QA Plan 21 had an acceptable alternative, that alternative was 22 submitted.
23 In most cases, the previous commitments 24 to N-45 standards provided the appropriate controls 25 for activities related to the ESP application.
61 1 Following the inspection, TVA did revise the Fleet NQAP 2 to show conformance with 17.5.
3 The revision clarified or included 4 requirements for certain site-specific activities 5 occurring at various stages of facility life. Work 6 activities include, but are not limited to: management, 7 planning, site investigation, design, and procurement.
8 Next slide.
9 As I mentioned, the NRC came and did an 10 inspection between April 16 and April 20. Areas 11 inspected included 10 CFR 21, corrective actions, QA 12 records, internal audits, organization, design 13 control, procurement, document control, and control 14 of purchased materials, equipment, and services.
15 The conclusion in the NRC Inspection Report 16 was of no violations or non-conformances being 17 identified.
18 So, based on that information, TVA 19 concludes that the TVA Quality Assurance Plan meets 20 the requirements of 10 CFR 50, Appendix B, and 10 CFR 21 52.17. That concludes my presentation.
22 CHAIRMAN KIRCHNER: Thank you, Michelle.
23 Any questions from members? We're missing Dick 24 Skillman, he usually has a very pointed question to 25 ask, this is with license renewals, about commitment
62 1 of the organization to its QA Program.
2 So, Ray, I'm going to ask you about that.
3 So, how does the management stand behind this 4 application? I mean, pretty much, right now, we're 5 talking about paper. But where are you in terms of 6 an actual implemented program?
7 MR. SCHIELE: So, right now, we are using, 8 taking credit for, the TVA program, which is fully 9 implemented and used at all three sites. So, we are 10 part of that program right now.
11 It is the plan to eventually transition 12 to a full standalone NQA-1 program for the project, 13 should it decide to move forward. But right now, we 14 are part of the fleet, fully implemented, NQA Program.
15 CHAIRMAN KIRCHNER: Thank you.
16 MR. SCHIELE: Yes.
17 CHAIRMAN KIRCHNER: Anyone else? Okay.
18 With that, then I believe we would turn here to the 19 staff. Thank you, Michelle. Okay. Allen, are we 20 set?
21 MR. FETTER: Okay. Good morning. Allen 22 Fetter. As Mallecia said, I'm the other Safety Project 23 Manager on this review.
24 Mr. Nicholas Savwoir is from the Office 25 of New Reactors, in the Division of Construction,
63 1 Inspection, and Operational Programs, under the Quality 2 Vendor Inspection Branch I.
3 He has four years of quality assurance 4 experience at the NRC and has an electrical engineering 5 degree from North Carolina's A&T State University.
6 Prior to the NRC, he performed ship 7 alterations and troubleshooting on analog and digital 8 instrumentation and control systems for submarines and 9 aircraft carriers under NAVSEA's Nuclear Propulsion 10 and Planning Department at Norfolk Naval Shipyard.
11 Today, his first presentation before the 12 ACRS, he will be presenting the review of the Site Safety 13 Evaluation Report, Section 17.5, Quality Assurance 14 Program Description. Okay. Go ahead, Nick.
15 MR. SAVWOIR: Good morning, ACRS. Again, 16 my name is Nicholas Savwoir, I'm part of the Division 17 of Construction, Inspection, and Operational Programs 18 under the Quality Vendor Inspection Branch I. And good 19 afternoon, good morning. Next slide.
20 The Chapter 17.5 regulations which pertain 21 to the early site permit consist of the 18 quality 22 assurance criteria of 10 CFR 50, Appendix B, and also, 23 10 CFR 52.17(a)(1)(xi) and (a)(1)(xii).
24 (a)(1)(xi) specifically requires the ESP 25 applicants to provide a description of the Quality
64 1 Assurance Plan applied to the site-related activities.
2 And (a)(1)(xii) requires the ESP 3 applicants to include an evaluation against the NRC's 4 most current quality assurance guidance six months 5 prior to the docketed date. Next slide.
6 I guess I'll start a little bit with the 7 background history, and, basically, some of the 8 information, to summarize the application, which led 9 to my review.
10 So, as required by 10 CFR 52.17, an 11 applicant is to provide a description of the Quality 12 Assurance Plan applied to site-relate activities. And 13 as a result, TVA, they submitted their operating NQAP, 14 which was Revision 32.
15 TVA's NQAP, it commits to the ANSI 16 N45.2-1971, as endorsed by the NRC's Reg Guide 1.28 17 Rev 3. However, at the time, six months prior to the 18 docketed date, NQA-1-2008 was in effect and endorsed 19 by NRC's Regulatory Guide 1.28 Rev 4.
20 And because we evaluate submittals using 21 the current regulatory framework, we conducted multiple 22 public meetings and clarification calls to resolve any 23 differences with the operating fleet's NQAP and a 24 submittal, in accordance with the regulations, which 25 is 10 CFR 52.17 stated, as earlier.
65 1 From the staff's review, we issued one RAI 2 with eight questions, and as a result of the staff's 3 review, TVA did revise the submittal, the NQAP Revision 4 32 to Revision 26, to address the staff's questions.
5 Next slide.
6 So, as a part of my review, I reviewed all 7 the 18 criteria of Appendix B, 10 CFR 50, and also, 8 I performed my own gap analysis for my review against 9 the Reg Guide 1.28. And also indicated by my SE, you 10 can see that in my gap analysis.
11 So, for this presentation, I would just 12 like to summarize this into -- summarize my review and 13 the RAIs into three overall key areas.
14 The first area is for the Quality Assurance 15 Program Description, which is in accordance with 16 Criterion I for Organization, and also Criterion II 17 for Quality Assurance Program.
18 The second key review area is for the 19 Quality Assurance Gap Analysis, in accordance with 20 Criterion XVII, which is QA Records, Criterion VII for 21 Control of Purchased Material, Equipment, and Services, 22 and Criterion XV for Nonconforming Materials, Parts, 23 or Components.
24 And last but not least, the third key review 25 area is for the QA Implementation and Inspection. That
66 1 was conducted April 16-20 of this year at TVA 2 Headquarters in Chattanooga. Next slide.
3 So, the first key review area is for the 4 Quality Assurance Program Description, specifically, 5 at the Clinch River Nuclear Site.
6 And as a result of my interactions with 7 TVA, the NRC staff identified the need for additional 8 information for the small modular reactor organization 9 for the Clint River Nuclear Site, and also, the 10 independent assessments that would be conducted at the 11 Clinch River Nuclear Site, in addition to the reference 12 or the commitment of 10 CFR 52, because inside their 13 NQAP that was submitted, there was no indication of 14 that at all.
15 So, as a result of the staff's review, TVA, 16 they revised the NQAP to Revision 36, which basically 17 added the Appendix K, which addressed the roles and 18 responsibilities, and also the authorities.
19 Also, they added Appendix L, which is an 20 organization chart specific for the small modular 21 reactor or organization which, in their Appendix I, 22 didn't address at all. And also, they added 10 CFR 23 52 to the NQAP Revision 36 that I'll talk about later.
24 Next slide.
25 So, my second key review area --
67 1 MEMBER RICCARDELLA: Why is it Revision 36?
2 MR. SAVWOIR: So, there were several 3 iterations of the revisions. From my knowledge and 4 experience, they revise it, I believe, every Christmas.
5 And so, basically, after this two-year period, there 6 were internal revisions and things of that nature.
7 MEMBER RICCARDELLA: Thank you.
8 MR. SAVWOIR: Yes. So, my second key review 9 area is for the gap analysis, and also, the Criterion 10 XVII for the Quality Assurance Records.
11 So, as a result of my interactions with 12 TVA, the NRC staff, we identified the need for 13 additional information for the gap analysis, which was 14 the difference between Revision 3 and Revision 4 of 15 Reg Guide 1.28.
16 And also, the Clinch River Nuclear Quality 17 Assurance Records and also, the Clinch River Nuclear 18 Electronic Records Controls.
19 So, as a result of the staff's review and 20 the RAIs, we -- the RAI I generated with the eight 21 questions, TVA, they revised the NQAP to Revision 36.
22 TVA, they submitted a gap analysis 23 evaluation during the inspection that was conducted 24 this April and they also added Appendix M to address 25 the Clinch River Nuclear Commitments and Clarifications
2 They also committed Reg Guide 1.28 Rev 4.
3 And they also identified the documents that are 4 considered QA Records per Criterion XVII of the 5 regulations. They also added the Electronic Records 6 per RIS 2000-18 and the NIRMA guidance. Next slide.
7 MEMBER RICCARDELLA: Excuse me?
8 MR. SAVWOIR: Yes.
9 MEMBER RICCARDELLA: Could you just give 10 me a description as to what a gap analysis is? It's 11 a new term for me.
12 MR. SAVWOIR: Yes. So, I guess, in essence, 13 what a gap analysis is, it's basically an evaluation.
14 An evaluation as the regulations require, 15 per 10 CFR 52.17, in which -- as TVA indicated earlier, 16 they did a full matrix, which is a chart that went 17 through all the criterion of the Quality Assurance 18 Criterion of Appendix B and they did an evaluation and 19 opened corrective actions, if there was any 20 discrepancies between the two, or addressed them in 21 the revision.
22 MEMBER BALLINGER: Why did you pick Revision 23 3 and 4? Because there is a Revision 5.
24 MR. SAVWOIR: So, the regulations require 25 that it's six months prior to the docketed date.
69 1 MEMBER BALLINGER: Okay.
2 MR. SAVWOIR: And at the time, Revision 4 3 was the -- so, yes, same slide. Oh, next slide. Yes.
4 Okay.
5 So, to continue with the second key review 6 area for the gap analysis, which addresses the Criterion 7 VII, which is the Control of Purchased Materials, Parts, 8 and Equipment, and Services, and Criterion XV, which 9 is Nonconforming Materials.
10 So, as a result of the staff's interactions 11 with TVA, the staff, we identified the need for 12 additional information, because there was an incorrect 13 exemption for the use of accreditation in lieu of 14 commercial grade surveys for procurement of laboratory 15 calibration and test services.
16 And also, TVA, they did not address the 17 notification of affected organizations for 18 nonconforming material and parts and components within 19 this NQAP they submitted.
20 So, as result of the staff's review and 21 the RAI generated, TVA, they revised the NQAP. They 22 revised the ILAC conditions per the NEI 14-05 guidance, 23 which is the guidelines for the use of accreditation 24 in lieu of commercial grade surveys for procurement 25 of laboratory calibration and test services.
70 1 And also, they added an Appendix M and the 2 commitments to address the notification of affected 3 organizations. Next slide.
4 So, my last, but not least, my third key 5 review area was for the Quality Assurance 6 implementation, that I was a part of, and also, Greg 7 Galletti, who's sitting over there on the side, that 8 was conducted April 16-20 of this year, 2018. And we 9 used the Inspection Procedure 350117, which is the QA 10 Implementation Inspection.
11 And, basically, this inspection assessed 12 the aspects of TVA's process, their procedures, and 13 their implementation of the Quality Assurance 14 activities used for the Clinch River Nuclear early site 15 permit application, which also included the 16 organization, the Quality Assurance Program, the QA 17 Records, the design control, corrective actions, 18 audits, oversight of contractor activities, and also, 19 10 CFR 21.
20 And based upon, at this inspection, we 21 actually -- this was the initial review, where we were 22 able to look at the Revised 36. So, basically, the 23 draft, what it would look like and what it would contain, 24 as far as addressing the RAIs.
25 At the time, there were no findings of
71 1 significance were identified and the qualification and 2 the Quality Assurance Inspection Report is publicly 3 available at the accession number here on the slide.
4 So, in conclusion, on the basis of the 5 staff's review of Chapter 17.5 of the Clinch River 6 Nuclear Site early site permit application and the NQAP 7 Revision 36, the staff concludes the applicant's QAP 8 Description for the Clinch River Nuclear Site early 9 site permit meets the regulatory requirements of 10 10 CFR 50, Appendix B, and also, 10 CFR 52.17. Any 11 questions?
12 CHAIRMAN KIRCHNER: Thank you. Members, 13 any further questions at this point? Okay. Thank you, 14 Nicholas. We let you get off easily this time.
15 (Laughter.)
16 CHAIRMAN KIRCHNER: But, welcome.
17 MR. SAVWOIR: Thank you.
18 CHAIRMAN KIRCHNER: Okay. Let's move on 19 to Hydrology. I know we're showing a break, but I think 20 we can push on and probably get this done before lunch.
21 So, are we ready? Okay, Ray, you're ready? Please 22 proceed.
23 MR. SCHIELE: Yes, Mr. Chairman, we'd like 24 to continue our presentation with Section 2.4, 25 Hydrology. I'd like to introduce John Holcomb, who'll
72 1 be presenting. John?
2 MR. HOLCOMB: Thank you, Ray. Good 3 morning. My name is John Holcomb, I'm a civil engineer.
4 I've been with TVA for nine years, on various 5 construction operations and licensing projects. I'm 6 currently service as the TVA SMR Engineering Manager.
7 The presentation for ESPA Site Safety 8 Analysis Report Section 2.4, Hydrologic Engineering, 9 has been divided into three areas. There will be a 10 brief description of the NRC interactions related to 11 Section 2.4.
12 We'll present an overview of the Tennessee 13 River System and the Clinch River Watershed, prior to 14 the technical presentations. We will also have an 15 overview of each of the 14 sections of 2.4. Next slide.
16 In April of 2017, the NRC conducted an audit 17 to review the site hydrologic engineering information 18 presented in Site Safety Analysis Report Section 2.4 19 of the ESPA.
20 The audit consisted of an office visit, 21 with a general presentation of the Clinch River Site.
22 The staff provided 40 audit information needs to TVA 23 prior to the audit. And TVA's responses were presented 24 and discussed during the audit.
25 Following the audit, TVA docketed their
73 1 responses to the NRC. These responses have been 2 incorporated into Revision 1 of the early site permit 3 application.
4 The audit also consisted of a site tour, 5 including site hydrologic engineering features in terms 6 of four TVA dams upstream of the Clinch River Site.
7 We'll discuss more of these dams later in the 8 presentation. Next slide.
9 Before we get into technical details of 10 the presentation, I would like to give an overview of 11 the Clinch River Site as it relates to the hydrologic 12 characteristics of the site.
13 Details of the Clinch River Site hydrologic 14 description are provided in SSAR Section 2.4.1, 15 Hydrologic Description of the ESPA. Next slide.
16 On this slide, you'll get a perspective 17 for the spatial relationship between the significant 18 dams near the Clinch River Site. The Clinch River Site 19 is shown by the red circle on the left of the map, and 20 I'll also use the pointer here.
21 One of the most important dams relative 22 to flooding of the Clinch River Site is Norris Dam.
23 As shown here in the map, Norris Dam is located 52 miles 24 above the site on the Clinch River.
25 Melton Hill Dam is located approximately
74 1 five miles upstream of the Clinch River, as shown here 2 on the map, and has a small amount of storage capacity.
3 The Watts Bar Dam backwater is a primary 4 factor in the water elevation at the Clinch River Site.
5 The Watts Bar Dam is located about 50 miles downstream 6 of the site, and that's shown here on the map.
7 Because of the importance of the Watts Bar 8 Dam backwater on the site elevation, we also show on 9 this map the key dams above Watts Bar, on the Tennessee 10 River and its main tributaries.
11 The most important of these are the 12 Cherokee Dam, the Douglas Dam, the Fontana Dam, and 13 the Fort Loudoun/Tellico Dam Complex. Next slide.
14 Go ahead.
15 MEMBER CORRADINI: Is this the same 16 information, just shown differently? The one you just 17 flipped to?
18 MR. HOLCOMB: That one right there?
19 MEMBER CORRADINI: Yes.
20 MR. HOLCOMB: Yes. This is so you can get 21 an idea of the hydraulic flow of the dams, this is --
22 MEMBER CORRADINI: Okay.
23 MR. HOLCOMB: Yes.
24 MEMBER CORRADINI: All right.
25 MR. HOLCOMB: The other one gives you a
75 1 spatial --
2 MEMBER CORRADINI: Because the other one, 3 I didn't catch. This one --
4 MR. HOLCOMB: Yes, this is just a pictorial 5 to easily show all the dams on one slide.
6 MEMBER CORRADINI: Okay.
7 MR. HOLCOMB: The other one is for spatial 8 description.
9 MEMBER CORRADINI: Sure.
10 MR. HOLCOMB: All right.
11 MEMBER CORRADINI: So, the site is the red 12 dot and water flows up the screen?
13 MR. HOLCOMB: So, the -- yes.
14 MEMBER CORRADINI: Or water flows down the 15 screen?
16 MR. HOLCOMB: Water flows down the screen.
17 MEMBER CORRADINI: Down the screen?
18 MR. HOLCOMB: Yes.
19 MEMBER CORRADINI: Okay.
20 MR. HOLCOMB: All right. The TVA water 21 control system is large and diverse, as you can see 22 in this diagram.
23 Unlike many utilities that have dams 24 affecting flooding at their site which are under control 25 by external entities, such as the Army Corps of
76 1 Engineers, the Tennessee River System is controlled 2 by TVA.
3 The exceptions are small dams controlled 4 by the Corps of Engineers and other power generation 5 entities.
6 MEMBER CORRADINI: So, what -- all of the 7 ones we see here are controlled by TVA?
8 MR. HOLCOMB: Except for two or three 9 smaller ones on here, but --
10 MEMBER CORRADINI: Can you just kind of --
11 MR. HOLCOMB: -- TVA's River --
12 MEMBER CORRADINI: -- highlight where those 13 are? I'm sorry.
14 MR. HOLCOMB: Stu, do you mind pointing 15 those out?
16 MR. HENRY: Yes. The ones that are not 17 controlled by TVA are up here on the Little Tennessee:
18 Chilhowee, Cheoah, Santeetlah, Thorpe. I think TVA 19 does handle the Nantahala.
20 MEMBER CORRADINI: So, it's on the upper 21 right where these are not controlled by you all?
22 MR. HENRY: Correct.
23 MEMBER CORRADINI: And due to flood control, 24 there are procedures that are normally instituted in 25 terms of what to handle, based on season and location?
77 1 MR. HOLCOMB: That is correct.
2 MEMBER CORRADINI: Okay. Lot of dams.
3 MR. HOLCOMB: The TVA River Forecasting 4 Center regulates the Tennessee River and major 5 tributary flow to maximize flood management, power 6 generation, and recreation.
7 The main reservoirs are lower in the late 8 fall, winter, and early spring, to maximize flood 9 storage. The main reservoirs are raised in late 10 spring, summer, and early fall, to increase electric 11 generation and provide for general recreation.
12 The staff toured the River Forecasting 13 Center as part of the April 2017 audit. The River 14 Forecasting Center is staffed 24/7 to monitor and 15 control the TVA River System.
16 These operation characteristics, known as 17 operating rules, as well as established flood guides, 18 are integrated into the hydrologic analysis for the 19 Clinch River Site.
20 The TVA dams within the water control 21 system are under the TVA Dam Safety Program. Changes 22 in the TVA water control system that potentially impact 23 the flooding analysis at the TVA Nuclear Sites are 24 evaluated by the TVA Nuclear Power Group.
25 MEMBER CORRADINI: So, I'm sorry to get
78 1 particular, I'm just trying to understand. So, the 2 red dot is actually where it is or is the red dot really 3 a little bit higher, where the river kind of winds around 4 the site? I'm trying to get geographically oriented.
5 MR. HOLCOMB: So, if this was actually --
6 MEMBER CORRADINI: Clinch River is to the 7 left, right? Upper left?
8 MR. HOLCOMB: Yes. So, the Clinch River 9 is here. You got Melton Hill Dam, the Clinch River 10 Site is just south on the river of the dam. And then, 11 you have the Watts Bar Backwater Reservoir, which we've 12 been discussing.
13 MEMBER CORRADINI: Okay. Well, the reason 14 I'm asking the question --
15 MR. HOLCOMB: Yes, go ahead.
16 MEMBER CORRADINI: -- is that on the actual 17 map, which is back on some slide that you don't have 18 to go back to, shows that the river winds around the 19 site.
20 And yet, the way you have it described here, 21 it's off to the side of the winding around. So, I assume 22 that's wrong and the actual map is right.
23 MR. HOLCOMB: Ray, can you go to the next 24 slide, please?
25 MEMBER BROWN: The red dot's in the wrong
79 1 place?
2 MR. HOLCOMB: Yes.
3 MEMBER BROWN: Because that's what he's 4 trying to say.
5 MR. HOLCOMB: Yes, so --
6 MEMBER BROWN: I had the same question --
7 MR. HOLCOMB: -- right here --
8 MEMBER BROWN: -- but he got ahead of me.
9 MR. HOLCOMB: Yes. So, the red dot is, that 10 is basically a cartoon drawing depicting --
11 MEMBER CORRADINI: Yes, it's fine, it's 12 fine, it's fine.
13 MR. HOLCOMB: Yes. So, if you look here, 14 you'll see the site is --
15 MEMBER CORRADINI: That's fine.
16 MR. HOLCOMB: -- north of the river.
17 MEMBER CORRADINI: I like the cartoon 18 drawing, because I can understand the geography of all 19 the various dams and what feeds what. But that's one 20 thing that confused me. All right, thank you.
21 MR. HOLCOMB: Yes, you are correct. As 22 shown in this picture, the Clinch River Site is on the 23 north bank of the Clinch River, about five miles 24 downstream of the Melton Hill Dam.
25 The planned finished grade at the site is
80 1 821 feet, approximately 80 feet above the normal river 2 water elevation. The Watts Bar Dam Backwater Reservoir 3 level is typically the main factor in the actual water 4 level at the Clinch River Site.
5 MEMBER CORRADINI: Can you repeat that last 6 statement, please?
7 MR. HOLCOMB: Yes. The Watts Bar Dam 8 Backwater Reservoir level is typically the main factor 9 in the actual water level at the Clinch River Site.
10 MEMBER CORRADINI: So, the downstream dam 11 and what it holds up determines the base level, due 12 to any sort of event?
13 MR. HOLCOMB: That is correct.
14 MEMBER CORRADINI: Okay. Okay, thank you.
15 MR. HOLCOMB: All right. The Watts Bar 16 Operating Guide is set at 735 feet in the winter and 17 740 feet in the summer.
18 Since the building of the dams on the Clinch 19 River and Tennessee Rivers, the maximum floods occurred 20 in 1973 and 2003, and were estimated to have reached 21 elevations of 749 at the Clinch River Site.
22 The site has a significant margin of over 23 70 feet between historical flooding levels and the 24 planned plant grade. Next slide. Go ahead.
25 MEMBER CORRADINI: I am sure there's a
81 1 Regulatory Guide that tells you what to worry about, 2 so those aside. If you go back historically, you said 3 it was 1970 and something and 2003. If you go back 4 even further, there's nothing that was higher than those 5 in recorded --
6 MR. HOLCOMB: So, when they installed the 7 dams, it drastically changed the river systems. So, 8 that's why --
9 MEMBER CORRADINI: Oh, and so, the Watts 10 Bar Dam is of what vintage?
11 MR. HOLCOMB: Stu, can you --
12 MEMBER CORRADINI: So, what you're saying 13 is, prior to that, it was lower?
14 MR. HENRY: Watts Bar is later than that.
15 We can get that information for you.
16 MEMBER CORRADINI: Well, I'm just trying 17 to understand historically. But your point, I just 18 want to make sure I don't confuse the issue, your point 19 is, when the dam comes up, what it holds back determines 20 the base from which you have to worry about the flood 21 level? And that is back decades ago, in terms of the 22 Watts Bar Dam?
23 MR. HOLCOMB: Yes. So, what this slide is 24 saying is that, since the dams have been installed, 25 this is the highest flood level. Now, there may be
82 1 different flooding levels historically, but that was 2 before the dams were installed.
3 MEMBER CORRADINI: Okay.
4 CHAIRMAN KIRCHNER: The system was begun 5 in the mid-1930s.
6 MEMBER CORRADINI: That's what I 7 remembered, yes.
8 MR. HOLCOMB: Okay.
9 CHAIRMAN KIRCHNER: Now, when you give these 10 nominal elevation numbers, you are considering, what?,
11 an A and B site on the actual map?
12 MR. HOLCOMB: That is correct.
13 CHAIRMAN KIRCHNER: And that hasn't been 14 resolved yet. Is there any significant differential 15 elevation between A and B?
16 MR. HOLCOMB: No, the planned site elevation 17 is 821 for either site. Next slide.
18 Section 2.4, Hydrologic Engineering, 19 describes hydrological characteristics of the Clinch 20 River Site. This section addresses hydrologic 21 characteristics and natural phenomena that have the 22 potential to affect the design-basis for the surrogate 23 plant.
24 This section is divided into 14 25 subsections, for each hydrological characteristic, as
83 1 shown here. We will briefly describe how TVA addressed 2 the majority of these and give more detail to describe 3 the 2.4.3.4 and 2.4.3.12 characteristics. Next slide.
4 CHAIRMAN KIRCHNER: Before you go into great 5 detail here, could you just refresh for the record and 6 for the members, just refresh at least my memory on, 7 with your Plant Parameter Envelope, what your heat sink 8 is and what your requirements are, if any, from the 9 river system that you're on?
10 MR. HOLCOMB: For the PPE, we looked at all 11 four of the reactor vendor technologies and none of 12 them utilized the river system as the ultimate heat 13 sink. So, it is all passive technologies, so they're 14 not dependent on the river system for a heat sink.
15 CHAIRMAN KIRCHNER: And for heat rejection, 16 it's cooling towers?
17 MR. HOLCOMB: For the PPE, that's what was 18 assumed for the analysis --
19 CHAIRMAN KIRCHNER: Right.
20 MR. HOLCOMB: -- for the ESP, it was cooling 21 towers.
22 CHAIRMAN KIRCHNER: Thank you.
23 MR. HOLCOMB: Next slide, Ray. With the 24 exception of three characteristics that we'll discuss 25 in more detail, we'll present the remainder of the
84 1 characteristics in three groups.
2 The first group is hydrologic 3 characteristics demonstrated to have no safety-related 4 impacts. These include Subsection 2.4.2, Floods. For 5 this characteristic, the preliminary plant grade of 6 821 feet is well above the maximum flood level.
7 For Subsection 2.4.7, Ice Effects. Due 8 to climate conditions and the elevated design, the plant 9 grade in combination with the SMR plant design, it is 10 concluded that the ice effects will not cause flooding 11 or water availability concerns.
12 MEMBER CORRADINI: So, the rive has had ice 13 on it in the past, it's just, again, the elevation 14 precludes concern? That's what I wanted to understand.
15 MR. HOLCOMB: That and also, the design of 16 the SMRs in consideration.
17 MEMBER CORRADINI: What does the ice do?
18 Since we have a minute or two. Does it back the water 19 up or does it cause it to divert into tributaries?
20 I'm kind of --
21 MR. HOLCOMB: So, the ice could have varying 22 effects, depending on what you're analyzing. It could 23 be blocking of the cooling water source, if you were 24 depending on it for a heat sink, or it could be changing 25 in the flood level due to blockage of the river system.
85 1 MEMBER CORRADINI: Have you had that 2 combination of ice effects and a flood event 3 historically there?
4 MR. HOLCOMB: Stu, can you speak to that?
5 MR. HENRY: Not that I'm aware of. There's 6 very little icing on the river. We just, we don't get 7 enough cold weather in that area of the country, in 8 order for the ice to form and build up sufficiently.
9 MEMBER CORRADINI: I see, okay. I'm from 10 a different climate. Thank you.
11 MR. HOLCOMB: Next slide. The third 12 characteristic in this category is Subsection 2.4.9, 13 Channel Diversions.
14 A review of the hydrologic, hydraulic, 15 climatic, topographic, and geologic evidence and 16 anthropogenic impacts on the Clinch River arm of the 17 Watts Bar Reservoir indicates that the channel 18 diversions are not expected in the Clinch River during 19 the operating life of the plant. Next slide.
20 The fourth characteristic in this grouping 21 is Subsection 2.4.10, Flooding Protection 22 Requirements.
23 The design-basis flood level is well below 24 the grade elevation of the site and minimal backwater 25 effects are anticipated due to the local intense
86 1 precipitation event.
2 The local intense precipitation event 3 would be evaluated further at COLA. There are no 4 expected flood protection requirements. Next slide.
5 The last characteristic in this group is 6 Subsection 2.4.13, Accidental Releases of 7 Radionuclides in Ground and Surface Waters.
8 Subsection 2.4.13 describes the evaluation of an 9 accidental release of the liquid radio effluents into 10 the ground and surface waters.
11 This evaluation assumes the contents of 12 a radwaste tank stored onsite are released into the 13 groundwater. The contents of the tank were determined 14 utilizing a PPE approach.
15 The source term is conservatively based 16 on unfiltered RCS fluid, with a failed fuel fraction 17 of one percent. To assess the source term for 18 reasonableness, the values were compared to those that 19 were previously approved by the NRC.
20 This assessment concluded that the PPE 21 values were reasonable and once released into the 22 groundwater, it is transported to the Clinch River, 23 that is 1,400 feet away.
24 That is based on the shortest travel 25 distance from any assumed release point on the Clinch
87 1 River Site to the Clinch River. The resulting total 2 dose from all exposure pathways to the river receiving 3 the maximum dose meets the 10 CFR 20.1301 limit. Next 4 slide.
5 MEMBER CORRADINI: It doesn't meet it, what 6 was the estimate in comparison to the limit? I guess 7 I --
8 MR. HOLCOMB: Alex, do you have a number 9 you can provide?
10 MR. YOUNG: Alex Young, Design Engineer for 11 the SMR Project. Before I attempt to read the number 12 off the top of my head, let me just confirm with our 13 calculations.
14 MEMBER CORRADINI: We're not in a rush, take 15 your time.
16 (Laughter.
17 MR. YOUNG: Yes, 93 rem TEDE, compared to 18 the --
19 MEMBER CORRADINI: Okay.
20 MR. YOUNG: -- 100 --
21 CHAIRMAN KIRCHNER: Millirem?
22 MR. YOUNG: Millirem, yes --
23 MEMBER CORRADINI: I figured --
24 MR. YOUNG: -- 93 millirem TEDE, excuse me.
25 MEMBER CORRADINI: I figured you meant that,
88 1 thank you.
2 (Laughter.)
3 MEMBER CORRADINI: All right, thank you.
4 CHAIRMAN KIRCHNER: Just to put that in 5 perspective, you assumed one percent failed fuel. The 6 branch technical position suggests a lower number than 7 that?
8 MR. HOLCOMB: That is correct.
9 CHAIRMAN KIRCHNER: That's a big 10 difference, that's --
11 MR. HOLCOMB: But that adds --
12 CHAIRMAN KIRCHNER: -- an order of magnitude 13 difference.
14 MR. HOLCOMB: Yes. That adds some 15 conservatism.
16 CHAIRMAN KIRCHNER: All right, thank you.
17 MR. HOLCOMB: The next group of hydrologic 18 characteristics are those considered to be unlikely 19 hazards at the site. This group includes Subsections 20 2.4.5 and 2.4.6.
21 Subsection 2.4.5, Probable Maximum Surge 22 and Seiche Flooding. Because the site is not located 23 on an open or large body of water, surge or seiche 24 flooding will not produce the maximum water levels at 25 the site.
89 1 For Subsection 2.4.6, Probable Maximum 2 Tsunami Hazards. The site is not subject to any tsunami 3 events originating from the ocean, due to the distance 4 from the nearest seacoast. Next slide.
5 The third and last group of hydrologic 6 characteristics are those demonstrated not to apply 7 due to the design of the SMR reactors under 8 consideration.
9 Because the Clinch River is not used as 10 a safety-related water supply for the small modular 11 reactor designs being considered, Subsection 2.4.8, 12 Cooling Water Canals or Reservoirs, and Subsection 13 2.4.11, Low Water Considerations, do not apply.
14 And as shown on the next slide, Subsection 15 2.4.14, Technical Specifications and Emergency 16 Operation Requirements, also does not apply.
17 As we begin the remainder of the 2.4 18 presentations, I would like to introduce the Subject 19 Matter Experts TVA employed to assist us in preparing 20 these subsections.
21 We have Stu Henry of Barge Design 22 Solutions. He'll present Subsections 2.4.3 and 2.4.4.
23 And he'll be followed by Dr. Hillol Guha, who will 24 be joining us shortly, of Bechtel Engineering.
25 MR. HENRY: Thank you, John. My name is
90 1 Stu Henry. I'm a civil engineer and Vice President 2 at Barge Design Solutions, for over 20 years. I've 3 assisted TVA with nuclear site flooding potential 4 calculations for the last ten years. Next slide.
5 The flooding guidance that was used in the 6 calculations followed the Regulatory Guide 1.59, 7 supplemented by the best current practice.
8 We used the Weather Service 9 Hydrometeorological Reports 41, 51, 52, and 56, as well 10 as previous watershed-specific guidance from the 11 National Weather Service to TVA. We reviewed ANS 2.8 12 and used the current practice in NUREG/CR-7046 as well.
13 Next slide.
14 For dam failure guidance, again, we used 15 the Reg Guide 1.59 and reviewed ANS 2.8. The current 16 practice was from the Japanese Lessons Learned 17 Directorate, Interim Staff Guidance 2013, as well as 18 that in the CR-7046. Next slide.
19 The CRN simulations were run looking at 20 the probable maximum precipitation based the HMRs 21 applicable to the basin's size and location.
22 Inflows were calculated based on 100 23 percent runoff, there were no losses applied there, 24 and the unit hydrographs were adjusted for a nonlinear 25 basin response, as recommended by the CR-7046. The
91 1 routing software was the Corps of Engineers HEC-RAS 2 software.
3 And the downstream project, the Watts Bar 4 Dam, as we discussed, has an impact on the site due 5 to backwater. And it was assumed stable under all 6 conditions to maximize the impact at the site.
7 The dam stability was determined by the 8 TVA Dam Safety Organization and that was used and 9 assumed in the calculations. Next slide.
10 The controlling flood simulations were 11 found to be the probable maximum flood, produced the 12 highest calculated water surface at the site.
13 Seismically-induced and sunny day dam failure 14 simulations were performed, but were found not to be 15 controlling.
16 The PMF and seismic simulation results show 17 the site to be dry, with significant margin. And, 18 again, the local intense precipitation will be 19 evaluated at COLA, since there are no specific site 20 plans at this time.
21 And we'll -- as soon as he gets up here, 22 that concludes my part of the presentation and I will 23 hand off to Hillol Guha for the groundwater.
24 MEMBER CORRADINI: Perfect timing.
25 DR. GUHA: Perfect timing, exactly. I've
92 1 been running. Okay.
2 CHAIRMAN KIRCHNER: Feel free to take your 3 time setting up.
4 DR. GUHA: Okay, thank you. Good morning, 5 actually, it should be good afternoon, I thought. Good 6 morning to all of you.
7 My name is Hillol Guha and I'm a 8 hydrogeologist with 20 years of experience and I work 9 for Bechtel, supporting TVA on the Clinch River ESP 10 project.
11 I have been associated with this project 12 since early 2013 and undertook a few subsurface 13 investigations and originated groundwater flow and 14 transport modeling calculations. Next slide, please.
15 So, this slide provides the outline of the 16 groundwater investigation. As stated in Section 17 2.4.12, and which includes regional to local 18 hydrogeology, specific data collected from the Clinch 19 River Breeder Reactor Project and the CRN Site.
20 Also, we'll discuss maximum groundwater 21 levels from groundwater modeling, any groundwater used, 22 and construction de-watering. The figure to the lower 23 right shows Oak Ridge Reservation area, to the east 24 of the Clinch River Nuclear Site. Next slide, please.
25 So, this figure depicts a cross-section
93 1 for the east Tennessee aquifer system of the Valley 2 and Ridge province. The principal aquifer is composed 3 of carbonate rocks of the Knox group.
4 Groundwater movement is localized by the 5 repeating lithology created by thrust faulting. Older 6 rocks sits on top of younger rocks and dips towards 7 the southeast.
8 The Chickamauga and the Knox group are the 9 principal lithologic formations in the Clinch River 10 Nuclear area. The Chickamauga group is composed of 11 limestone, siltstone, shale, while the Knox group is 12 made up of dolomite.
13 Groundwater primarily flows along the 14 strike of the bedding plane, that is along the weathered 15 rocks and fractures. Groundwater flow significantly 16 diminishes with depth due to less fractures and more 17 competent bedrock.
18 Within a 1.5 mile radius of the CRN Site, 19 there are 32 residential wells, three commercial wells, 20 and one farm well, for a total of 36 individual wells.
21 The estimated yields range from 0.5 to 75 22 gallons per minute. None of these wells occur in the 23 CRN Site. Thus, there is no groundwater withdrawal 24 at the CRN Site. Next slide, please.
25 So, this slide shows the conceptual
94 1 hydrogeologic model of the CRN Site. The conceptual 2 hydrogeologic model is similar to the adjacent Oak Ridge 3 Reservation area to the east.
4 From top to bottom, the conceptual model 5 is divided into a stormflow zone, that is a thin region 6 at the surface where 90 percent or more water from 7 precipitation move at this zone.
8 This zone is absent at the CRN Site, due 9 to the Clinch River Breeder Reactor Project rework.
10 Below the stormflow zone is the unsaturated zone or 11 the Vadose zone. The thickness varies. It is thicker 12 in the ridges and reach up to 100 feet. And nearly 13 absent near stream channels.
14 Groundwater zone is the next zone. Also, 15 the water table zone. And is encountered at the top 16 of the bedrock. This zone could be few feet to more 17 than 100 feet and conveys ten percent of the subsurface 18 flow. Below the groundwater zone is the aquiclude, 19 where flow is nonexistent. Next slide, please.
20 So, the figure on the right of this slide 21 shows some of the boring locations from the Clinch River 22 Breeder Reactor Project, which was undertaken between 23 1972 to 1980.
24 Total of 129 borings, 37 observation wells, 25 11 piezometers, and 117 bedrock packer permeability
95 1 tests were undertaken. Groundwater levels fluctuated 2 by as much as 20 feet, due to response to precipitation 3 events.
4 Groundwater flows from topographically 5 high areas in the center of the peninsula to the low 6 relief areas that is towards the Clinch River arm of 7 the Watts Bar Reservoir. Chestnut Ridge, located north 8 of the site, acts as a groundwater divide. Next slide, 9 please.
10 So, CRN Site subsurface investigations 11 were undertaken between 2013 to 2015, which included 12 82 borings, three test pits, 44 observation wells, 41 13 packer tests in 30 wells, one pumping test, and two 14 chemical sampling in 34 observation wells. Also 15 included geophysical investigations.
16 Nested observation wells were installed 17 in two-well cluster and three-well cluster. The 18 adjacent figure to the right depicts the location of 19 the observation wells. The wells were screened at 20 different depths.
21 Groundwater flow was predominantly along 22 fractures and joints, with active flow at shallow depths 23 at the interface of the soil and weathered rocks.
24 Flow was predominantly along the strike 25 that is trending north, 52 degrees east. And the
96 1 frequency of fractures and joints decrease with depth.
2 Dominant groundwater flow was between 812 to 712 feet 3 elevation. The Clinch River acts as a sink for the 4 shallow groundwater flow zone.
5 Pumping test was conducted within the 6 square box, as shown in the figure on the adjacent slide.
7 The horizontal radius of pumping test influence was 8 limited to approximately 150 feet.
9 MEMBER RICCARDELLA: Excuse me, what 10 exactly is a pumping test?
11 DR. GUHA: So, the pumping test also known 12 as aquifer performance test. It is the test where, 13 what you do is, basically, you stress the aquifer and 14 once you stress the aquifer through pumping and you 15 have observation wells, and the signals, you observe 16 the signals through the drawdown in those wells. And 17 then, you analyze that data to come up with the 18 hydrogeologic parameters of the subsurface.
19 MEMBER CORRADINI: So, do you put water in 20 or take water out?
21 DR. GUHA: You basically, in this case, we 22 took out the water.
23 MEMBER CORRADINI: Okay. And then, you 24 watched the behavior on surrounding --
25 DR. GUHA: Surrounding observation wells.
97 1 CHAIRMAN KIRCHNER: Just to calibrate us 2 a bit, where is the basemat elevation expected to be, 3 approximately? How many feet?
4 DR. GUHA: So, you mean to say the basemat 5 here is the --
6 CHAIRMAN KIRCHNER: Of the -- you mentioned 7 power block, I'm thinking of the reactor building and 8 its foundation.
9 DR. GUHA: So, the --
10 CHAIRMAN KIRCHNER: What do you expect it 11 to be, approximately, in terms of elevation?
12 DR. GUHA: Yes, so, we did a PPE, which is 13 Plant Parameter Envelope --
14 CHAIRMAN KIRCHNER: Yes.
15 DR. GUHA: -- so, because this is a ESPA.
16 So, we did two analysis, which I will show in the later 17 slides.
18 CHAIRMAN KIRCHNER: Okay.
19 DR. GUHA: And we had, one was the shallow 20 foundation depth of the reactor building, which is 50 21 feet below the grade elevation of 821.
22 CHAIRMAN KIRCHNER: Okay.
23 DR. GUHA: I think it was approximately 770 24 feet elevation. And the deep was 140 below, below the 25 grade, which came close to, I think, 658, something
98 1 like that, shallow than that. I will come back to those 2 slides later.
3 CHAIRMAN KIRCHNER: Please.
4 DR. GUHA: After a few more slides.
5 CHAIRMAN KIRCHNER: Thank you.
6 DR. GUHA: Yes, sure. Next slide, please.
7 So, this figure to the right shows horizontal 8 groundwater flow directions of potentiometric surface.
9 Groundwater flows towards the southeast or southwest, 10 from the proposed nuclear island towards the Clinch 11 River arm of the Watts Bar Reservoir.
12 The figure to the left shows vertical 13 groundwater flows on equipotential lines, which 14 dominant downward vertical gradient at the center of 15 the peninsula and flows upwards to the Clinch River.
16 Next slide, please.
17 So, this slide shows a geological 18 cross-section along northwest and southeast. That is, 19 along the dipping direction of the rocks. The Chestnut 20 Ridge Fault is shown on the left of the figure and occurs 21 further north of the proposed -- this one right here, 22 excellent.
23 So, this is the Chestnut Ridge Fault, which 24 occurs further north of the proposed nuclear site.
25 The Knox dolomite of the Newala formation outcrops just
99 1 north of the proposed nuclear site. This is the Knox 2 dolomite.
3 The Chickamauga group lies on top of the 4 Knox group. And the Chickamauga group of rocks dips 5 southeasterly at an average dip of 33 degrees. This 6 is an average dip angle of 33 degrees.
7 The Chickamauga group consists mainly of 8 limestone and the Chickamauga group is divided into 9 the Blackfoot formations, the Eidson formation, and 10 the Fleanor member of the Lincolnshire formations.
11 But they are all part of the Chickamauga group.
12 The Rockdell, Benbolt, Bowen, Witten, and 13 Moccasin formations, they are also part of the 14 Chickamauga group.
15 The Fleanor member is comprised of 16 approximately 75 to 80 meter of maroon calcareous shale, 17 siltstone, with numerous light gray limestone bed.
18 So, this is the Fleanor member. So, all average dipping 19 at 33 degrees towards southeasterly dipping. Next 20 slide, please.
21 So, this is the slide where you have the 22 foundation depths that are discussed. So, this slide 23 shows a post-construction groundwater model for five 24 section along the strike of the bedding plane.
25 So, this is the strike of the rock, this
100 1 is the direction of the rock. And so, along this 2 profile is what you see the model section or the profile 3 section that has been implemented here.
4 MEMBER CORRADINI: So, just so I've got it.
5 So, you cut this so that you can see the rock angular 6 deviation, and the colored pictures on the left are 7 river-to-river.
8 DR. GUHA: That's correct, yes. This is 9 the -- the river bends from this side --
10 MEMBER CORRADINI: Yes.
11 DR. GUHA: -- like that, yes.
12 MEMBER RICCARDELLA: And those are sections 13 through the planned view in the middle?
14 DR. GUHA: This is the section planned 15 through this sections, yes. And so, you have one --
16 so, we did a PPE, Plant Parameter Envelope. So, one 17 is the deep foundation of that reactor. And this is 18 the shallow foundation of the reactor. So, but along, 19 this is particularly showing around this profile 20 section.
21 So, the slide shows the post-construction 22 groundwater model profile sections along the strike 23 of the bedding plane, that is trending north, 52 degrees 24 east, along which the predominant groundwater flows.
25 The center figure shows the location of two profile
101 1 section.
2 So, this is one profile section that has 3 been shown for the deep foundation and the shallow 4 foundation. And there is another one profile section 5 that we have done similar, but is not shown in the slide.
6 The second figure shows the location of 7 the two profile sections. So, one profile section is 8 shown on the left. The colors within the figure depicts 9 various layers within the groundwater model that are 10 different hydrogeologic properties.
11 The dark area depicts the foundation 12 embedment of the reactor. So, this is the reactor, 13 the rad waste, the auxiliary building, and the turbine 14 building, right here. This is the rad waste. So, this 15 is the deep foundation depth for the reactor building.
16 The dark area depicts the foundation 17 embedment of the reactor building, rad waste, the 18 turbine, and the auxiliary buildings. The deep 19 embedment depth of the reactor building is set at an 20 elevation of 681 feet, which is 140 feet below the site 21 grade.
22 The figure below shows the foundation 23 embedment depth of the shallow reactor. This is the 24 shallow reactor, with the auxiliary building, and you 25 have the rad waste building here, as well as the turbine
102 1 building there.
2 At an elevation of 770 feet, which is 3 approximately 50 feet below the site grade elevation.
4 The deep and shallow reactor foundation depths serve 5 as bounding limits as part of the Plant Parameter 6 Envelope.
7 The figure to the right depicts groundwater 8 contours in color blue, for both the figures with deep 9 and shallow foundation depths.
10 The maximum groundwater elevations under 11 and around the structure varies between 802.3 to 816.1 12 feet elevations. So, this value is less than the site 13 grade elevation of 821 feet.
14 So, the red arrow, the red arrows here, 15 depicts downward flows from the center of the nuclear 16 island. And the blue arrow depicts upward flow to the 17 Clinch River, which acts as a sink. Next slide, please.
18 CHAIRMAN KIRCHNER: So, before you go on, 19 in the case of Profile A, where you have a very deep 20 foundation, that's below, the bottom of that foundation 21 is below the river level, if I'm interpreting this 22 correctly.
23 And yet, you still show gradients flowing 24 out to the river. So, just explain, in physical terms, 25 why that is so.
103 1 DR. GUHA: So, the natural groundwater 2 gradient is basically -- so, you have -- this is the 3 center of the peninsula --
4 CHAIRMAN KIRCHNER: Right.
5 DR. GUHA: -- so, that's where the buildings 6 are. And the natural, just pre-construction --
7 CHAIRMAN KIRCHNER: Right.
8 DR. GUHA: -- is basically you have the flows 9 going towards the Clinch River, off the Breeder Reactor.
10 CHAIRMAN KIRCHNER: Okay.
11 DR. GUHA: One went this side, one went 12 towards your southeasterly and other going toward 13 southwesterly.
14 So, this -- only difference from the 15 pre-construction, now this is the post-construction, 16 only difference what you have is basically 17 incorporation of this foundation depths, the structure 18 depths.
19 So, you have the reactor building, the 20 turbine, all the structures up there. So, and then, 21 surrounding the structures are your -- the structural 22 backfill material.
23 But your -- it still remains, even within 24 that area, only in the limited area just around the 25 structures, the gradients a little bit could be altered
104 1 a little bit. But overall, you still have -- it still 2 remains the natural flow gradient direction.
3 CHAIRMAN KIRCHNER: Okay.
4 DR. GUHA: So, it's still flowing towards 5 the river.
6 CHAIRMAN KIRCHNER: Okay, thank you.
7 MEMBER CORRADINI: So, this is not to scale.
8 So, whether it's Site A or B, what is the width of 9 the hole versus, you said the depth was 150 and 50?
10 DR. GUHA: So, this is --
11 MEMBER CORRADINI: It's not to scale, that's 12 what I'm trying to get at.
13 DR. GUHA: Yes. This is --
14 MEMBER CORRADINI: So, what's the width of 15 the black thing versus the depth? The depth is 150 16 and 50 in Profile A and Profile B.
17 DR. GUHA: Right.
18 MEMBER CORRADINI: What is the width?
19 DR. GUHA: So, the width, you mean to say 20 the width from here to here?
21 MEMBER CORRADINI: The width of the black 22 stuff.
23 DR. GUHA: Oh, the width of the black stuff, 24 this would be approximately -- I got to look at it, 25 I got to look at the thing.
105 1 MEMBER CORRADINI: Approximately.
2 DR. GUHA: Yes. Approximately, I think it 3 will be close to about, about minimum will be maybe 4 200 feet.
5 MEMBER CORRADINI: So, the L/D is still it's 6 wider than it is deep?
7 DR. GUHA: The --
8 MEMBER CORRADINI: Right? What you just 9 said was, since this is not to scale, the black cylinder 10 is more like a couple of hundred feet wide and 150 deep 11 versus 50, have I approximately got it right?
12 DR. GUHA: Yes.
13 MR. HOLCOMB: It is wider than it is deep.
14 DR. GUHA: Yes.
15 MEMBER CORRADINI: Okay, fine. Because 16 this --
17 DR. GUHA: This is, yes, exaggerated.
18 MEMBER CORRADINI: Okay.
19 MEMBER RICCARDELLA: Excuse me, what are 20 the contour lines on the box on the right? The contours 21 of what?
22 DR. GUHA: So, this is the contour of the 23 groundwater levels.
24 MEMBER RICCARDELLA: Okay.
25 DR. GUHA: So, potentiometric surfaces.
106 1 So, you're seeing, on the one, the arrows, depicts the 2 direction of the groundwater flow. The red's showing, 3 you have a downward groundwater flow direction.
4 And the ones which is in your blue shows, 5 depicts flow towards the river, which is an upward 6 gradient. But the water is basically discharging to 7 the river. Okay. So, next slide.
8 So, there is no groundwater usage at the 9 CRN Site SMR designs. Potable and other water for site 10 usage will come from Oak Ridge Department of Public 11 Works. The makeup water for the closed cycle cooling 12 system will be sourced from the Clinch River arm of 13 the Watts Bar Reservoir. Next slide, please.
14 So, there will not be any permanent 15 de-watering system during operation of the plant.
16 Temporary de-watering will be required during 17 excavation, which will be based on similar techniques 18 was was done during the CRBRP excavation, such as 19 installation of horizontal gravity drains in the 20 excavation rock faces, pumping from sumps located in 21 the perimeter of the excavation and the base of the 22 excavation. And the flow rate is expected to be 23 minimal, as was observed in the CRBRP excavation.
24 MEMBER CORRADINI: So, can you back to the 25 black cylinder?
107 1 DR. GUHA: Sure.
2 MEMBER CORRADINI: So, let me ask it 3 differently. So, if I had a deep embedment, will that 4 create a sink and I'll have water accumulation there?
5 In the soil?
6 Or is the calculation or the estimate is 7 that essentially it is unperturbing the flow past the 8 cylinder and you're still feeding the river? That's 9 what I'm trying to understand with the different 10 embedments.
11 DR. GUHA: So, what we are seeing here is 12 -- so, this is the -- so, basically, only change from 13 the pre-construction, the existing condition, is 14 basically, you're incorporating this black, the 15 reactor, all these buildings, and you have this backfill 16 that's been included.
17 MEMBER CORRADINI: Well, what I guess --
18 okay. So, you've actually gotten to what I was going 19 to ask. Is the way the backfill is designed such that 20 you won't have essentially an accumulation of water 21 around the cylinder, it essentially will flow past?
22 DR. GUHA: Yes.
23 MEMBER CORRADINI: Okay. And that's a 24 typical construction approach? Since I'm not 25 familiar, but that's what is normally done?
108 1 DR. GUHA: Yes.
2 MEMBER CORRADINI: Okay. So, one last 3 question, at least for the moment, are these -- 50 feet 4 embedment seems typical, 150 feet seems atypical. Are 5 those typical embedments in certain civil structures?
6 DR. GUHA: You're talking relative to the 7 nuclear reactors or --
8 MEMBER CORRADINI: Well, let's start 9 generally and then --
10 DR. GUHA: Okay.
11 MEMBER CORRADINI: -- we can get specific.
12 So, generally, I would think, yes. But specifically 13 to nuclear structures, I'm not familiar with 150-foot 14 embedments.
15 DR. GUHA: Nor do I actually. So, this is 16 SMR, I guess that's --
17 MEMBER CORRADINI: I'm just looking for 18 experiential deviations that cause me concern.
19 CHAIRMAN KIRCHNER: But again, the gray 20 matter in the -- you're illustrating -- unfortunately, 21 we can't read this very well. The gray contour there 22 is like bedrock, essentially. Is that what I'm to infer 23 from the left-hand --
24 DR. GUHA: You --
25 CHAIRMAN KIRCHNER: -- with the shallower
109 1 foundation, you're probably going to go in and backfill 2 and then, put the mat down, so to speak, the bottom 3 of the foundation. It looks like the upper one would 4 reach into a bedrock-like structure, in terms of the 5 foundation conditions.
6 MEMBER CORRADINI: But they still would have 7 to put something -- they would still have to put back 8 -- they'd have to make a bigger hole and put backfill.
9 DR. GUHA: Yes. So, basically, anywhere 10 where you have -- so, the way -- so, I should have said 11 before, the way the geology goes here -- so you have 12 this construction backfill, so this is something coming 13 in during the construction.
14 Then, you have this -- your -- below that, 15 you have the fill, the soil materials. And below that, 16 you have this -- your weathered zone, weathered rock, 17 which is an interface of the bedrock, as well as in 18 the soil. And below that, you have this competent 19 bedrock.
20 CHAIRMAN KIRCHNER: Yes.
21 DR. GUHA: So, where the fractures are very 22 less.
23 CHAIRMAN KIRCHNER: Yes.
24 DR. GUHA: So, most of this, the foundation, 25 the depth that you're seeing, the deep foundation
110 1 basically rests within that competent bedrock, as you 2 pointed out.
3 And on the shallow foundation, basically, 4 is still in the part of the competent, still it is 5 competent, but not as competent as that --
6 CHAIRMAN KIRCHNER: Okay.
7 DR. GUHA: -- because it's shallower. The 8 fracture frequency basically increases as you go up.
9 CHAIRMAN KIRCHNER: Basically, you're going 10 to be in dolomite or limestone with that foundation 11 on the top?
12 DR. GUHA: So, you'll be basically --
13 CHAIRMAN KIRCHNER: I'm trying to marry 14 several different elevation views of the geology to 15 convince myself where your foundation is sitting in 16 each of these pictures on the left. It looks like 17 you'll be in either -- let me get the correct group.
18 You'll be in the Knox group or the --
19 DR. GUHA: The Chickamauga group.
20 CHAIRMAN KIRCHNER: -- Chickamauga group.
21 DR. GUHA: So, the Knox group actually 22 outcrops further north of the site.
23 CHAIRMAN KIRCHNER: And that means you'll 24 be sitting in limestone, a well-anchored foundation 25 on the top left picture. The bottom left, you would
111 1 probably then put some kind of material in and then, 2 float the foundation, the concrete.
3 DR. GUHA: So, the Chickamauga group, yes, 4 is mostly composed of limestone.
5 CHAIRMAN KIRCHNER: Okay.
6 DR. GUHA: But it's basically -- it's 7 composed of also siltstone. Siltstone, shale 8 materials. So, yes, exactly. And so, it's very likely 9 it'll be anchored within that siltstone --
10 CHAIRMAN KIRCHNER: Right.
11 DR. GUHA: -- group.
12 CHAIRMAN KIRCHNER: So, you're not going 13 to be in sandstone?
14 DR. GUHA: It's not going to be in the 15 sandstone.
16 CHAIRMAN KIRCHNER: Okay, good --
17 DR. GUHA: The sandstone is --
18 CHAIRMAN KIRCHNER: -- for seismic reason.
19 DR. GUHA: Yes.
20 CHAIRMAN KIRCHNER: Good. All right. I'm 21 just -- and once again, the river level is --
22 DR. GUHA: At 740 feet elevation.
23 CHAIRMAN KIRCHNER: And that upper left one 24 is like 680 feet elevation at the bottom?
25 DR. GUHA: Yes, 680 approximately, yes.
112 1 CHAIRMAN KIRCHNER: Okay. Thank you.
2 DR. GUHA: I think we are in the last slide, 3 I guess. Yes, so basically, this is the second to the 4 last slide, yes.
5 CHAIRMAN KIRCHNER: Yes. We were trying 6 to ask questions until you got here.
7 (Laughter.)
8 CHAIRMAN KIRCHNER: Okay. Keep going.
9 DR. GUHA: Okay. So, this is the concluding 10 slide. It says, is the groundwater conclusion. The 11 following can be concluded from the CRN Site 12 hydrogeology investigation.
13 The proposed CRN Site SMR designs do not 14 rely on groundwater during the operations. Permanent 15 de-watering is not required. The maximum water levels 16 are below the site grade of 821 feet. That is range 17 between 802.3 to 816.1 feet elevation.
18 CHAIRMAN KIRCHNER: Okay. Members, any 19 questions?
20 MEMBER CORRADINI: Yes, I'm still trying 21 to understand qualitatively what's going on. So, 22 you're trying to bound the proposition.
23 But what I'm trying to get at is, your 24 conclusion is, regardless of the embedment depth, and 25 the combination of essentially the backfill and the
113 1 rock structure, is there is not going to be a 2 preferential sink for water to accumulate at the bottom 3 of the black cylinder? That's what I'm worried about.
4 DR. GUHA: So, groundwater is moving 5 through, so it's like conductivity, just similarly --
6 MEMBER CORRADINI: Sure.
7 DR. GUHA: -- so, it's the same concept as 8 in electricity. So, you have various conductive 9 materials. So, the backfill is the conductive 10 material, the backfill material is higher than the 11 native hydrogeology property material. So --
12 MEMBER CORRADINI: So, it allows for -- it 13 prevents accumulation?
14 DR. GUHA: That's correct.
15 MEMBER CORRADINI: Okay. Remind me what 16 the backfill is planned to be? I'm sorry, the -- I 17 remember this, because another one of the Subcommittee 18 meetings, we had a discussion about voids and finding 19 voids and voids bigger than 15 feet, et cetera, et 20 cetera. Remind me what the backfill material is?
21 DR. GUHA: So, the backfill material is 22 going to be, it's a structured material. So, it's going 23 to be a material from the site itself. So, it will 24 be composed of, if I understood, it's composed of 25 material which is of -- could be a limestone --
114 1 MEMBER CORRADINI: Okay. So, it's the base 2 rock crushed up --
3 DR. GUHA: Yes.
4 MEMBER CORRADINI: -- and reinstituted?
5 DR. GUHA: Yes, certain size and certain 6 grade level.
7 MEMBER CORRADINI: Okay. Thank you.
8 CHAIRMAN KIRCHNER: Okay. Thank you.
9 We'll turn to the staff, please.
10 DR. GUHA: Thank you to everyone.
11 MR. CAMPBELL: So, this is Andy Campbell, 12 Deputy in DLSE. Presenting for the staff will be Yuan 13 Cheng and Joe Giacinto and Mallecia Sutton.
14 MS. SUTTON: Thank you, Andy. So, this 15 presentation you have Yuan, Joe, and Rich Clement.
16 Dr. Yuan Cheng will be presenting on the surface water.
17 Dr. Cheng holds a professional engineering 18 license in several states, including Maryland, 19 Pennsylvania, and Ohio. He has worked for NRC 20 approximately five years as a hydrologist.
21 Prior to joining NRC, he worked in the 22 private sector for approximately 35 years. From 2013 23 to 2014 at NRC, he performed the technical review for 24 probable maximum flood for the license amendment 25 request for TVA Watts Bar Nuclear Power Plant Unit 1.
115 1 From 2015 to 2016, he performed another 2 technical review for Fukushima Near Term Task Force 3 Recommendation 2.1, Flood Hazard, reevaluations for 4 TVA's Watts Bar, Sequoyah, and Browns Ferry Nuclear 5 Plants.
6 Recently, he completed a technical review 7 for hydrologic engineering for the Clinch River early 8 site permit.
9 Mr. Joseph Giacinto will also present on 10 the review related to groundwater. Joe is a certified 11 professional geologist and has been with the NRC for 12 ten years, serving as a hydrologist and a geologist.
13 He served as staff hydrologic technical 14 lead for Lee, North Anna, PSEG, and Turkey Point 15 applications, and has participated in technical review 16 for all new reactor early site permit applications 17 submitted within the last ten years, to include Watts 18 Bar 2.
19 He has approximately 30 years of combined 20 public and private industry experience in the 21 hydrologic science.
22 Also, I discussed Dr. Clement this morning, 23 so I'm going to turn the presentation over to Dr. Cheng.
24 DR. CHENG: Good morning, hello, everyone.
25 And I would like to introduce my working team. Joe
116 1 Giacinto is a hydrogeologist and Richard Clements is 2 the health physicist.
3 Together with myself, Yuan Cheng, I'm a 4 hydrologist, we are NRC's Technical Reviewers for the 5 Site Safety Analysis Report Section 2.4, Hydrologic 6 Engineering, for the Clinch River Nuclear early site 7 permit application.
8 I will start with a brief background 9 summary and then, we will work our way towards the 10 staff's key areas of review for surface water, 11 groundwater, and radionuclide transport resulting from 12 a liquid effluent source release to groundwater and 13 the resulting dose estimates.
14 As shown, the Clinch River -- next slide, 15 please. As shown, the Clinch River Nuclear Site is 16 located adjacent to the Clinch River, a tributary of 17 the Watts Bar Reservoir, along the southwestern border 18 of the Oak Ridge Reservation, with the City of Oak Ridge, 19 Tennessee. Next slide, please.
20 Within the Valley and the Ridge geographic 21 province, the Clinch River Nuclear Site occupies 22 approximately 935 acres owned by the United States and 23 operate by the Tennessee Valley Authority, or TVA.
24 Site investigations and work associated 25 with the Clinch River Breeder Reactor Project was
117 1 conducted in the mid-1970s through the early 1980s on 2 what is now the Clinch River Nuclear Site.
3 After termination of the Breeder Reactor 4 Project, the Department of Energy, the project's 5 management corporation, and the Tennessee Valley 6 Authority, in coordination with the Nuclear Regulatory 7 Commission, conducted site redress activity to prepare 8 the site for future industrial use.
9 The proposed Clinch River Nuclear Site 10 grade is 821 feet. Next slide, please.
11 The staff reviewed the applicant's Plant 12 Parameter Envelope, which was based on four small 13 reactor technologies: BWXT mPower, NuScale, Small 14 Modular Reactor 160, and the Westinghouse Small Modular 15 Reactor. Next slide, please.
16 The staff review included a 17 pre-application review, site visit, and the audit, with 18 the audit taking place in 2017.
19 During the early site permit application 20 review, the staff consulted with the Department of 21 Energy, the Tennessee Department of Environment and 22 Conservation, and the US Geological Survey.
23 The Staff's Safety Evaluation Report, or 24 SER, has been completed with no open items. I will 25 now present the staff's findings for surface water.
118 1 Next slide, please.
2 The staff's review of the computations for 3 applicant's riverine flood elevation and the 4 applicant's considerations of the probable maximum 5 precipitation, surface runoff, and the dam failures 6 included in the flooding model scenarios.
7 In addition, the staff reviewed the 8 applicant's sensitivity study and confirmed that only 9 small change in the computed flood elevation occurred 10 when the modeling parameters were varied.
11 The staff reviewed the applicant's 12 riverine hydrologic model, which utilized the US Army 13 Corps of Engineers Hydrologic Engineering Center River 14 Analysis System, or HEC RAS, model for the modeling.
15 The staff confirmed that the applicant used 16 the historical flood events to calibrate the model, 17 using reasonable parameters. The staff confirmed the 18 applicant's hydrologic models could be used to 19 reasonably estimate the probable maximum flood 20 elevation at the Clinch River Nuclear Site.
21 The staff then reviewed each of the 22 applicant's considerations in developing the flood 23 scenario as follows.
24 For the probable maximum precipitation 25 estimates, the staff confirmed that the applicant
119 1 followed the methodologies as described in the National 2 Oceanic and Atmospheric Administration's 3 Hydrometeorological Reports, or HMRs, to probably 4 compute the various storm size and they reasonably 5 select the probable maximum precipitation, or PMP.
6 Regarding surface runoff, the staff 7 confirmed that the applicant's methods for converting 8 the probable maximum precipitation to surface runoff 9 were reasonable.
10 For the dam failure scenario, the staff 11 confirmed that the applicant set applicable dams for 12 instantaneous failures. And so, the staff reviewed 13 the applicant's simulations of the resultant flood wave 14 due to the dam failures.
15 The staff found that the applicant 16 reasonably determined a probably maximum flood 17 elevation from riverine flooding utilizing 18 conservative assumptions, which includes 100 percent 19 of river dams converted into surface runoff, 20 instantaneous dam failure, and intentionally maximize 21 backwater effect on the Clinch River Nuclear Site.
22 The staff reviewed the applicant's 23 modeling results and found that the applicant's 24 probable maximum flood elevation is significantly below 25 the site grade elevation. Next slide, please.
120 1 Local intense precipitation, or LIP, 2 effects are a flood-causing mechanism associated with 3 the site drainage design and the site grading plan.
4 Because no reactor technology has been selected for 5 the early site permit applications, neither a drainage 6 system design, nor a site grading plan was included.
7 The staff deferred the evaluation of the 8 localized flooding due to local intense precipitation 9 and has posted COL action item 2.4-1 for a later 10 evaluation of local flooding, which could be included 11 in the applicant's combined license or construction 12 permits. Next slide, please.
13 The needs for a flood protection plan is 14 dependent on the evaluation of the site grading plan 15 and the site drainage designs associated with a local 16 intense precipitation event.
17 In the early site permit applications, 18 neither a reactor technology and associated site 19 drainage design, nor a grading plan has been selected.
20 Therefore, the staff deferred the evaluation of the 21 flood protection plan and has included in the COL action 22 item 2.4-2, which should be included in the applicant's 23 combined license or construction permit.
24 Now, I will hand off the presentation to 25 Joe Giacinto for a discussion of the staff's groundwater
121 1 finding. Joe, next slide, please.
2 MR. GIACINTO: Thank you, Yuan. And good 3 morning to all. Based on the Plant Parameter Envelope, 4 the staff reviewed groundwater model simulations 5 developed by the applicant for a deep and a shallow 6 excavation geometry.
7 The maximum water level for these two 8 geometries was approximately 816 feet, based on the 9 groundwater modeling results, utilizing characteristic 10 aquifer parameters.
11 Staff determined that the maximum level 12 is conservative and well above maximum levels of 13 approximately 810 feet that have been observed during 14 the period of monitoring.
15 Staff notes that the backfill hydraulic 16 properties may affect water levels and, therefore, 17 included information in COL action item 2.5-8 in the 18 staff's Safety Evaluation Report Subsection 2.5.4.4.5, 19 Excavation and Backfill, where backfill 20 characteristics are evaluated in detail.
21 COL action item 2.5-8 was included in the 22 staff's October 17, 2018 ACRS presentation discussion 23 for Site Safety Analysis Report Section 2.5.4, 24 Stability of Subsurface Materials and Foundations.
25 Next slide, please.
122 1 In reviewing the literature for the site 2 and surrounding areas, staff found that low levels of 3 radionuclides have been documented for the Clinch River 4 Nuclear Site's groundwater samples, based on 2014 and 5 2015 reports associated with the Tennessee Department 6 of Environment and Conservation and the Department of 7 Energy's ongoing environmental monitoring studies for 8 the Oak Ridge Reservation.
9 After staff discussions with the Tennessee 10 Department of Environment and Conservation and the DOE 11 concerning the sampling results, staff determined that 12 COL action item 2.4-3 was necessary to differential 13 accident releases from existing background 14 concentrations, consistent with minimizing 15 contamination in accordance with 10 CFR 20.1406, 16 Minimization of Contamination.
17 MEMBER RICCARDELLA: And what is the source 18 of these current levels of radionuclides?
19 MR. GIACINTO: It's not been conclusively 20 been determined, but they are consistent with the 21 radionuclides that are coming off the Oak Ridge Site, 22 as a result of the weapons production from the 1940s 23 on.
24 MEMBER RICCARDELLA: Okay.
25 MR. GIACINTO: Yes.
123 1 MEMBER RICCARDELLA: Thank you.
2 MEMBER CORRADINI: Remind me how far away 3 the Oak Ridge Site is?
4 MR. GIACINTO: This site is adjacent to the 5 Oak Ridge Site, so it's --
6 MEMBER CORRADINI: So, we're talking ten 7 miles?
8 MR. GIACINTO: No --
9 MEMBER CORRADINI: Not even?
10 MR. GIACINTO: -- a matter of feet.
11 MEMBER CORRADINI: Oh, it's that close?
12 MR. GIACINTO: Now, the release areas from 13 the Oak Ridge Reservation are on the order of a mile 14 away.
15 MEMBER CORRADINI: Okay. But the signature 16 of what is being detected is judged to be from that 17 site?
18 MR. GIACINTO: Yes, fission products and 19 transuranics.
20 MEMBER CORRADINI: Okay. Thank you.
21 MR. GIACINTO: Next slide, please.
22 CHAIRMAN KIRCHNER: Before you go on, I'm 23 looking at the wording of this, is there any basis for 24 a contention later on, in terms of level of 25 contamination that's coming offsite onto this planned
124 1 site? In other words, from the Oak Ridge Reservation 2 to the planned site, what if that were to increase?
3 MR. GIACINTO: Well, it's been -- like I 4 say, from the 1940s on, this was released and the levels 5 on the Cinch River Nuclear Site are very similar to 6 those in the Hood Ridge area, just to the east across 7 the river.
8 And they're all basically right about or 9 just below detection limits in drinking water 10 standards.
11 CHAIRMAN KIRCHNER: Okay. So, they're well 12 down?
13 MR. GIACINTO: Yes. And in fact, there's 14 been some actions on that by the DOE in the Hood Ridge 15 area for the residences there.
16 CHAIRMAN KIRCHNER: Thank you.
17 MR. GIACINTO: Okay. So, reviewing the 18 literature for the site -- oops, sorry about that.
19 Slide 11.
20 Consistent with Appendix A to Part 50, 21 General Design Criteria for Nuclear Power Plants, 22 General Design Criterion II, Design-Basis for 23 Protection Against Natural Phenomena, the application 24 considered the most severe natural phenomena that have 25 been historically reported for the site and surrounding
125 1 area and appropriately evaluated the design-basis flood 2 elevation, including consideration of hypothetical dam 3 failure and wind-induced wave height resulting in a 4 design-basis flood level significantly below the site 5 grade of 821 feet.
6 Additionally, the maximum estimated 7 groundwater level is approximately five feet below site 8 grade. Staff determined that site characteristics are 9 bounded by the Plant Parameter Envelope.
10 Now, Richard Clement will summarize the 11 staff's findings for the determination of the source 12 term radionuclide transport and the resulting dose 13 evaluation. Rich? Next slide, please.
14 MR. CLEMENT: Thank you, Joe, and good 15 morning. The staff reviewed TVA's basis and 16 assumptions for developing the Plant Parameter 17 Envelope, or PPE, accident liquid source term in a 18 postulated accidental release to the groundwater at 19 the Clinch River Nuclear Site.
20 The accidental liquid source term is used 21 in the radionuclide transport analysis for estimating 22 the dose to a member of the public.
23 Although the PPE is based on four small 24 modular reactor designs, the application described that 25 design information from two vendors included features
126 1 to possibly mitigate a postulated accidental release 2 and, therefore, they were excluded from further 3 evaluation.
4 For the remaining two vendors, a 5 site-specific analysis would be expected in a combined 6 license application, using source term information in 7 those designs.
8 The staff determined that the accident 9 liquid source term developed from those two designs 10 considered conservative assumptions that included a 11 higher failed fuel fraction and an entire release of 12 radioactivity in the primary coolant.
13 In addition, the staff verified TVA's 14 comparison of its accident liquid source term to that 15 approved by the NRC in the Public Service Enterprise 16 Group early site permit.
17 The staff determined from its review and 18 confirmatory analysis that TVA's methodology for 19 developing the PPE source term to bound the dose to 20 members of the public from a postulated accidental 21 liquid release to the groundwater at the Clinch River 22 Nuclear Site was reasonable. Next slide, please.
23 The staff reviewed TVA's transport values 24 and assumptions and performed confirmatory 25 calculations for a select number of radionuclides,
127 1 using the guidance in NUREG/CR-3332 and Branch 2 Technical Position 11-6.
3 Conservative assumptions in TVA's 4 radionuclide transport analysis, in addition to those 5 used in developing that accident liquid source term 6 included selection of transport parameters and values 7 to minimize travel time and maximize radionuclide 8 concentrations, a catastrophic tank release scenario 9 assuming no credit for mitigating design features, and 10 instantaneous and direct release of the failed tank 11 contents into groundwater, peak radionuclide 12 concentrations, including daughter products, and 13 assumed minimal Clinch River flow rate of 400 cubic 14 feet per second, and minimal radionuclide travel 15 distance and decay from the release point to the Clinch 16 River.
17 Based on the review, the staff found TVA's 18 methodology for estimating initial radionuclide 19 concentrations at the site boundary from a postulated 20 accidental liquid release to the groundwater 21 reasonable.
22 The staff confirmed that the unity rule 23 applied in 10 CFR 20, Appendix B, Table 2, Column 2 24 for the mixture of radionuclide concentrations at the 25 site boundary was met. Next slide, please.
128 1 The staff verified TVA's input parameters 2 and assumptions in the exposure pathway dose analysis 3 associated with the accidental liquid release to the 4 groundwater using the guidance in Regulatory Guide 5 1.109.
6 The staff reviewed TVA's modifications 7 within the LADTAP II computer code, using the dose 8 conversion factors published in the Environmental 9 Protection Agency's Federal Guidance Reports 11 and 10 12, and found them reasonable and acceptable for 11 calculating the total effective dose equivalent, or 12 TEDE.
13 The staff confirmed that the public dose 14 limit of 100 millirem TEDE specified in 10 CFR 20.1301 15 was met.
16 Because the reactor design that may be 17 constructed at the Clinch River Nuclear Site is not 18 known at the early site permit stage, the staff 19 identified combined license, or COL, action item 2.4-4 20 for the COL or a construction permit applicant to 21 evaluate and justify any changes in the PPE source term 22 used in a postulated accidental release to the 23 groundwater and verify that the calculated dose 24 evaluated in the early site permit is bounded. Next 25 slide, please.
129 1 Based on the staff's review of TVA's early 2 site permit application, subject to the 3 staff-identified COL action items, the staff concludes 4 that the site characteristics and bounding site 5 parameters meet the applicable regulatory requirements 6 and that there is no undue risk to the public health 7 and safety.
8 Thank you. At this point, we will take 9 any questions or comments you may have.
10 CHAIRMAN KIRCHNER: Thank you. Members?
11 I should turn to Dennis, if he's still on the line.
12 Dennis, are you there? Theron's going to open up the 13 --
14 MEMBER BLEY: I am here, thank you, Walt.
15 CHAIRMAN KIRCHNER: Yes, Dennis, have you 16 any questions of the applicant or the staff?
17 MEMBER BLEY: I do not. I appreciated 18 today's presentations and I think they addressed the 19 issues pretty well. Thank you.
20 CHAIRMAN KIRCHNER: Thank you. Let me 21 then, we'll turn to the public in a moment. Let me 22 thank the applicant and the staff for your 23 presentations.
24 We've become well versed in the site's 25 geology and hydrology and maybe need a little work on
130 1 the meteorology, but thank you for your thoroughness 2 in all the presentations.
3 Now, if there's any member of the public 4 here in the audience who wishes to make a comment, please 5 come forward to the microphone, state your name, and 6 make your comment.
7 Seeing no one here, is there anyone on our 8 bridge line from the public who wishes to make a comment?
9 Please state your name and make your comment. Hearing 10 no one, I think we can close the bridge line, Theron.
11 Thank you.
12 So, at this point, Andy, I would like to 13 turn in your direction, in preparation for our full 14 Committee meeting in December, I think with the time 15 allotted, I would like to ask both you and the applicant 16 to focus on the emergency planning exemptions and the 17 analyses that back that up.
18 And that, I think, would be, with an 19 appropriate amount of introductory material, would be 20 the best use of our time during the meeting coming up 21 in December.
22 MR. CAMPBELL: So, let me make sure Mallecia 23 and I understand correctly, so we can be prepared.
24 What you would like the full Committee 25 meeting staff presentation to focus on is the EPZ and
131 1 the basis for our analysis that, I think the 2 Subcommittee received a briefing in August, 13.3, 3 right, Mallecia?
4 CHAIRMAN KIRCHNER: Yes, that's correct.
5 MS. SUTTON: That's correct.
6 CHAIRMAN KIRCHNER: It was August 22.
7 MS. SUTTON: Yes.
8 CHAIRMAN KIRCHNER: And we would ask that 9 you would, both the applicant and you, focus on that, 10 given the important precedent that will be set here 11 in going to more of a performance-based approach to 12 that topic.
13 MR. CAMPBELL: So, we'll come prepared to 14 make presentations and leave it up to the applicant 15 for developing their presentation. And then, we'll 16 focus on that area.
17 MEMBER CORRADINI: We have, just so you 18 remember, we have 90 minutes set aside for the full 19 Committee presentation.
20 MR. CAMPBELL: Ninety minutes, so that's 21 usually half for presentations --
22 MEMBER CORRADINI: Half a morning.
23 MR. CAMPBELL: Okay. We can do that. I'm 24 looking for Ray, there he -- Ray's thumbs up, okay.
25 CHAIRMAN KIRCHNER: Ray, are you good for
132 1 that?
2 MR. SCHIELE: Yes.
3 CHAIRMAN KIRCHNER: Okay. And keep in mind 4 that we've asked a number of questions, at least I think 5 I have, that are in the area of uncertainty.
6 When I look at the regulations, and forgive 7 me if I don't get the number right, I'm thinking 10 8 CFR 5034, there's an admonition in a footnote that these 9 values, in the case I'm thinking of, 25 rem, are not 10 viewed as limits, but not something to really be 11 approached.
12 So, I want to explore and make sure that 13 there is sufficient margin in what is presented. So, 14 if you can address the question of uncertainty and cover 15 that as part of your presentation, so we have confidence 16 that, in your independent review or confirmatory 17 analysis, that we do indeed have margin below the 18 requisite limits.
19 MR. CAMPBELL: And we'll do that.
20 CHAIRMAN KIRCHNER: Thank you. And if 21 there are any other comments by the members? No. With 22 that, then --
23 MEMBER BROWN: I have one other --
24 CHAIRMAN KIRCHNER: Oh, yes, Charlie.
25 MEMBER BROWN: -- thing for the meeting.
133 1 They went through a number of COL items --
2 MS. SUTTON: The green light and --
3 MEMBER BROWN: Sorry, I thought I had the 4 mic that was on, I just didn't bother to talk into it.
5 They went through a number of COL items to cover a 6 couple of the critical points, in terms of the -- I 7 would think they ought to just address those, as part 8 of the evaluation, to make sure those are clear as to 9 what needs to be done, since we don't really know what 10 the reactor is going to look like.
11 That's my only suggestion, as part of a 12 full Committee presentation. There weren't a lot, 13 there were half a dozen or a dozen, whatever they were, 14 that they went through.
15 CHAIRMAN KIRCHNER: Yes. These, like, 16 confirmatory items for the --
17 MEMBER BROWN: You have to come back --
18 CHAIRMAN KIRCHNER: -- COL applicant, 19 right.
20 MEMBER BROWN: -- with whatever --
21 CHAIRMAN KIRCHNER: Can you highlight 22 those, Andy, in a table that summarizes or addresses 23 at least the key, and all of them are important of 24 course, but those that you see as key requirements --
25 MR. CAMPBELL: So, let me --
134 1 CHAIRMAN KIRCHNER: -- for the COL 2 applicant?
3 MR. CAMPBELL: Let me parrot back to you 4 what I think I'm hearing. We -- you would like us to 5 focus on the EPZ --
6 CHAIRMAN KIRCHNER: Right.
7 MR. CAMPBELL: -- and the basis for our 8 analysis and the margin and the uncertainty. And then, 9 any COL action items that are related to that?
10 CHAIRMAN KIRCHNER: Primarily, and if there 11 are any other that are worth highlighting for the entire 12 Committee. Perhaps just a -- is the tabulation of them 13 very long? I'm -- we've seen them mainly by section 14 or chapter --
15 MR. CAMPBELL: Yes, in each SER section --
16 CHAIRMAN KIRCHNER: -- so, in my own mind, 17 I don't remember how many there are, overall, 18 confirmatory items.
19 MR. CAMPBELL: Mallecia probably knows that 20 answer, but not right off the top of her head.
21 MS. SUTTON: There's approximately 18 COL 22 action items for 13.3. So, if you want me to highlight 23 all of them and explain the substance of each?
24 MR. NGUYEN: Chairman?
25 CHAIRMAN KIRCHNER: How much -- yes?
135 1 MR. NGUYEN: I understand what your comment 2 is, I'll work with the staff and --
3 CHAIRMAN KIRCHNER: Okay.
4 MR. NGUYEN: -- to make an effective 5 presentation.
6 CHAIRMAN KIRCHNER: All right, thank you, 7 Quynh.
8 MS. SUTTON: And one other question. Just 9 18 for the EPZ or you want me to highlight the other 10 COL action items that we think are important to the 11 project?
12 MEMBER BROWN: I would suggest a few of them 13 that were related to dose or something like that --
14 MS. SUTTON: Okay.
15 MEMBER BROWN: -- to make sure --
16 CHAIRMAN KIRCHNER: Yes.
17 MEMBER BROWN: -- groundwater 18 transportation, dispersion, a few --
19 CHAIRMAN KIRCHNER: Right.
20 MEMBER BROWN: -- those relevant to the EPZ 21 as well. So, I didn't mean all, if there's 1,500 of 22 them, I didn't mean --
23 CHAIRMAN KIRCHNER: Right.
24 MEMBER BROWN: -- I'm exaggerating 25 slightly, but those that were really critical to the
136 1 --
2 CHAIRMAN KIRCHNER: To this issue, yes.
3 MEMBER BROWN: -- main decision, yes.
4 CHAIRMAN KIRCHNER: Yes.
5 MR. FETTER: Can I ask a clarifying 6 question, because we have COL action items on the order 7 of 15 or 16 for the geosciences area. Were you 8 interested in any of those?
9 MEMBER CORRADINI: So, let me help the 10 Chairman. So, not the whole Committee has heard all 11 the Subcommittee meetings. So, there's a chance that 12 a member is going to ask you something out of the blue, 13 so to speak. So, I think you have to be prepared for 14 that.
15 But I think what Walt's really saying is, 16 because of the exemption relative to the EPZ, you need 17 to focus on that, because that really is something 18 that's different, right?
19 But I think the other things, you've got 20 to be ready for. I'm sorry, but you've got to be ready 21 for them. But I wouldn't necessarily take a good deal 22 of time doing that.
23 MS. SUTTON: Okay.
24 MEMBER CORRADINI: Does that give you a 25 little more guidance?
137 1 MS. SUTTON: Yes, that's great.
2 MR. FETTER: Yes, that's very helpful.
3 MS. SUTTON: Thank you.
4 MR. CAMPBELL: We'll work with Quynh to make 5 sure we're clearly addressing your needs for the full 6 Committee presentation and make sure that we have 7 sufficient backup information in case we get the out 8 of the blue question.
9 MEMBER CORRADINI: And the members will be 10 highly disciplined.
11 (Laughter.)
12 MR. CAMPBELL: We appreciate that.
13 CHAIRMAN KIRCHNER: Okay. With that, then, 14 we are adjourned.
15 (Whereupon, the above-entitled matter went 16 off the record at 11:40 a.m.)
17 18 19 20 21 22 23 24 25
138 Clinch River Early Site Permit Part 2, SSAR Sections 2.3, 2.4, 11.2, 11.3 and 17.0 Advisory Committee on Reactor Safeguards November 14, 2018
139 Introduction Ray Schiele, SMR Licensing l
Acknowledgement and Disclaimer 140 Acknowledgment: "This material is based upon work supported by the Department of Energy under Award Number DE-NE0008336."
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The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof."
Advisory Committee on Reactor Safeguards l 3
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Energy Environment Economic Development Partner with 154 local power companies, to serve more than 9 million customers in parts of seven states. Directly serve 56 large industries and federal installations.
Advisory Committee on Reactor Safeguards l 4
NRC Review of ESPA 142 2017 2018 2019 2020 ESPA ESPA Rev. 1 ESPA Rev. 2 FEIS FSER NRC Issues Accepted Submitted Planned Submittal ESP 12-30-16 12-15-17 Dec 18 Safety Review PSER SER w/ no OIs 8/4 10/20 FSER 8/17 Full ACRS Audits & RAIs 5/15 8/22 10/17 11/14 12/5 ACRS Subcomm. Meetings Scoping Environmental Review Meeting 5/15 Comment Period FEIS 6/21 Notice of Intent Audits & RAIs DEIS 4/13 4/26 4 Contentions 2 Contentions Commission Hearing(s)
Filed Admitted Ruling 6/12 10/10 5/3 Notice of Hearing, Commission Hearing Opportunity TVA ASLB Contested 4/4 Appeals Ruling Hearing Advisory Committee on Reactor Safeguards l 5 11/6 7/31 Terminated
What is a Plant Parameter Envelope (PPE)?
143 Composite of reactor and engineered parameters that bound the safety and environmental impact of plant construction and operation Considers 4 SMR Vendors BWXT mPower NuScale Holtec SMR-160 Westinghouse Developed based on NEI 10-01 Guidance Margin added to specific parameters as appropriate Creates Franken-plant or a Black Box Plant Advisory Committee on Reactor Safeguards l 6
PPE Use Considerations 144 Includes Appropriate Conservatism Prevents rework when vendor analysis is updated Safety conclusion becomes more apparent Document and, when possible, quantify conservatisms Allows use of multiple reactor designs, providing flexibility for future business decisions.
An integral element of 10 CFR Part 52 Works well with a future COLA Advisory Committee on Reactor Safeguards l 7
Presentation Outline 145 Part 2, Site Safety Analysis Report (SSAR) Sections:
Section 11 - Radioactive Waste Management
- Alex Young Section 2.3 - Meteorology
- Alex Young Section 17 - Quality Assurance
- Michelle Conner Section 2.4 - Hydrology
- John Holcomb, Stu Henry, Hillol Guha Advisory Committee on Reactor Safeguards l 8
146 ESPA Part 2, SSAR Section 11.2 and 11.3 Radioactive Waste Management Alex Young, SMR Engineering
Key NRC Interactions Related to ESPA SSAR Chapter 11 147 One two-part audit was conducted to review the radioactive waste management information in the ESPA Audit Part 1 - April 14-17, 2017
- Bechtel Power Corporation office in Reston, VA Audit Part 2 - April 24-27, 2017
- TVA Knoxville Office Complex, Knoxville, TN
- Tour of CRN Site and Surrounding Area Supplemental Letter - June 16, 2017
- CNL-17-075 Resubmittal of Supplemental Information Regarding Radiation Protection Accident Consequences in Support of Early Site Permit Application for Clinch River Nuclear Site Advisory Committee on Reactor Safeguards l 10
Normal Radioactive Release Source Terms 148 Plant Parameter Envelope (PPE) Source Terms Annual activities released for each vendor were reviewed Composite source term developed on individual unit and site basis Generally assumes the maximum activity by individual radionuclides Assessed for being not unreasonable by comparing to previously approved source term Advisory Committee on Reactor Safeguards l 11
SSAR Section 11.2 - Liquid Radioactive Releases 149 LADTAP II used to calculate doses Exposure pathways assumed (RG 1.109)
Within the effluent concentration limits (ECLs) of 10 CFR 20, Appendix B, Table 2, Column 2 Doses are within design objectives of 10 CFR 50, Appendix I Doses are within the environmental standards of 40 CFR 190 Doses are within the limits of 10 CFR 20.1301 Advisory Committee on Reactor Safeguards l 12
SSAR Section 11.3 - Gaseous Radioactive Releases 150 GASPAR II used to calculate doses Exposure pathways and analytical methods consistent with RG 1.109 and RG 1.111 Within the effluent concentration limits (ECLs) of 10 CFR 20, Appendix B, Table 2, Column 1 Doses are within design objectives of 10 CFR 50, Appendix I Doses are within the environmental standards of 40 CFR 190 Doses are within the limits of 10 CFR 20.1301 Advisory Committee on Reactor Safeguards l 13
151 ESPA Part 2, SSAR Section 2.3 Meteorology Alex Young, SMR Engineering
Key NRC Interactions Related to ESPA SSAR Section 2.3 152 Two audits were conducted to review the meteorology information in the ESPA Audit - May 15-19, 2017
- TVA Knoxville Office Complex, Knoxville, TN
- Tour of CRN Site Including Meteorological Tower Location Audit - May 7-11, 2018
- Conducted via TVA Electronic Reading Room
- Supporting April 9, 2018 Supplemental Letter Supplemental Letter - April 9, 2018
- Comparing results utilizing vector- versus scalar-averaged wind directions Advisory Committee on Reactor Safeguards l 15
Subsection 2.3.1 Regional Climatology 153 CRN Site Characteristics (SSAR Table 2.0-1)
Winter Precipitation
- Normal Winter Precipitation Event - 21.9 psf
- Extreme Frozen Winter Precipitation Event - 21.9 psf
- Extreme Liquid Winter Precipitation Event (48-hour Probable Maximum Winter Precipitation (PMWP)) - 23.5 in Maximum Rainfall Rate - 18.8 in/hr, 6in/5-minutes Basic Wind Speed - 96.3 mph for 3-second gust Historical Maximum Wind Speed - 87 mph for 3-second gust, 73 mph fastest mile Design-Basis Hurricane Windspeed - 130 mph for 3-second gust Advisory Committee on Reactor Safeguards l 16
Subsection 2.3.1 Regional Climatology 154 CRN Site Characteristics (SSAR Table 2.0-1)
Tornado
- Maximum Pressure Drop - 1.2 psi
- Maximum Rotational Speed - 184 mph
- Maximum Translational Speed - 46 mph
- Maximum Wind Speed - 230 mph
- Radius of Maximum Rotational Speed - 150 ft
- Rate of Pressure Drop - 0.5 psi/s Advisory Committee on Reactor Safeguards l 17
Subsection 2.3.1 Regional Climatology 155 CRN Site Characteristics (SSAR Table 2.0-1)
Ambient Air Temperatures Exceedance Criteria Max. Dry- Max. Max. Non- Min. Dry-Bulb Temp. Coincident coincident Bulb Temp.
(°F) Wet-Bulb Wet-Bulb (°F)
Temp. (°F) Temp. (°F) 2% Annual Exceedance 90 73.7 75.7 25 1% Annual Exceedance 92 74.2 76.7 21 0.4% Annual 95 74.9 77.6 16 Exceedance 0% Annual Exceedance 105 74.6 81.7 -9 100-Year Return Period 107 73.1 83.6 -9.9 Advisory Committee on Reactor Safeguards l 18
SSAR Section 2.3.2 - Local Meteorology 156 Topography around the site strongly influences the local climate Predominant up-valley/down-valley flow is readily apparent at all three meteorological towers CRN Site conditions are consistent with regional conditions Advisory Committee on Reactor Safeguards l 19
SSAR Section 2.3.3 - Onsite Meteorological 157 Measurements Program Primary Meteorological Tower [1977-1978 and 1982-1983]
- 110-meter
- CRBRP Construction
- Reactivated for CRN ESPA Pre-Application Data [2011-2013]
Supplemental Meteorological Tower [1977-1978 and 1982-1983]
meter
- CRBRP Construction Temporary Meteorological Tower [1973-1978]
meter
- Pre-application Data for CRBRP Advisory Committee on Reactor Safeguards l 20
SSAR Section 2.3.3 - Onsite Meteorological 158 Measurements Program RG 1.23 references ANSI/ANS-3.11-2005 ANSI/ANS-3.11-2005 states that the transport wind direct for straight-line Gaussian models should be based on the scalar mean (or unit vector) wind direction TVA has evaluated the use of vector and scalar wind direction for the CRN Site Various differences in results between the two approaches Vector was bounding for SSAR Chapter 15 Vector was bounding for SSAR Chapter 11 Advisory Committee on Reactor Safeguards l 21
SSAR Section 2.3.4 - Short-Term (Accident) 159 Diffusion Estimates Atmospheric dispersion calculations performed using PAVAN Met the requirements of RG 1.145 and 1.23 Meteorological data from June 1, 2011 through May 31, 2013 No credit for building wake effects Assumed ground level release Advisory Committee on Reactor Safeguards l 22
SSAR Section 2.3.5 - Long-Term (Routine) 160 Diffusion Estimates XOQDOQ-82 utilized for calculating X/Q and D/Q Meteorological data from June 1, 2011 through May 31, 2013 16 wind direction sectors out to 50 miles Nearest residence, vegetable garden, and beef animal at each wind direction sector No credit given for building wake effects Assumes a ground-level release scenario Radioactive decay and deposition were considered Advisory Committee on Reactor Safeguards l 23
SSAR Section 2.3.5 - Complex Terrain 161 Made comparison of results with a variable trajectory model CALPUFF utilized for variable trajectory model Meteorological data from June 1, 2011 through May 31, 2013 Rainfall data was taken from Oak Ridge Automated Surface Observing System (ASOS)
No credit given for building wake effects Assumes a ground-level release scenario Concluded that the XOQDOQ model was bounding Advisory Committee on Reactor Safeguards l 24
162 ESPA Part 2, SSAR Section 17 Quality Assurance Michelle Conner, SMR Operations, Training, and Programs
Agenda for Quality Assurance- Section 17.5 163 Chronology CRN ESPA activities Program Description Quality Assurance Implementation Inspection Conclusion Advisory Committee on Reactor Safeguards l 26
CRN ESPA Quality Assurance Chronology 164 ESPA Rev. 1 Submitted to NRC - December 2017 NRC issued RAI on QA - March 9, 2018 TVA provided RAI response - April 9, 2018 NRC Quality Assurance Inspection - April 16-20, 2018 TVA issued the NQAP Rev 36 - May 8, 2018 NRC issued the QA Inspection Report - June 1, 2018 Advisory Committee on Reactor Safeguards l 27
TVA Nuclear Quality Assurance Plan Description 165 TVA NQAP is the top-level document that defines the quality assurance policy and assigns major functional responsibilities.
ESPA Part 2, provides a summary of the TVA CRN QA Plan attributes. The TVA CRN QAPD is a separately controlled document and is included in Part 8 of the ESPA.
The TVA NQAP was revised to meet SRP 17.5 that was in effect six months prior to the ESPA submittal.
For ESPA, the TVA NQAP applies to site suitability activities.
Advisory Committee on Reactor Safeguards l 28
NRC QA Inspection 166 NRC staff QA implementation inspection of TVAs ESPA activities for the proposed SMR at the CRN Site, from April 16 through April 20, 2018.
Areas inspected included 10 CFR Part 21, corrective actions, QA records, QAP, internal audits, QA organization, design control, procurement document control, control of purchased material, equipment, and services, and external audits.
No violations or non-conformances were identified.
Advisory Committee on Reactor Safeguards l 29
Conclusion 167 TVA NQAP provides adequate guidance for establishing controls to comply with the applicable requirements of 10 CFR Part 52.17(a)(xi) and (xii); and 10 CFR Part 50, Appendix B.
Advisory Committee on Reactor Safeguards l 30
168 ESPA Part 2, SSAR Section 2.4 Hydrology John Holcomb, SMR Engineering Stu Henry, Barge Design Solutions Hillol Guha, Bechtel
Presentation Agenda 169 NRC Interactions Related to ESPA SSAR Section 2.4
- Overview of Tennessee River System and Clinch River Watershed ESPA Development and Subsection Presentations
- General Hydrologic Characteristics of the Site
- Specific Hydrologic Characteristics of the Site o 2.4.3 - Probable Maximum Flood on Streams and Rivers (Stu Henry) o 2.4.4 - Potential Dam Failures (Stu Henry) o 2.4.12 - Groundwater (Hillol Guha)
Advisory Committee on Reactor Safeguards l 32
Key NRC Interactions Related to ESPA SSAR Section 2.4 170 One audit was conducted to review the site hydrologic engineering information in the ESPA Audit - April 24 - 27, 2017
- Office discussion o General presentation of the Clinch River site o Presentation and discussion of responses to 40 Audit Information Needs
- Site and Dam Tour o Tour site and site hydrologic engineering features including:
>> The bend in the Clinch River and surrounding topography that controls routing of flood flows;
>> Bridges bounding the CRN Site;
>> Proposed cut/fill areas for the CRN Project and existing backfill and backfilled areas of the former Clinch River Breeder Reactor Project
>> Areas of planned cooling water intake and discharge structures o Tour the TVA Norris, Melton Hill, Douglas (and its saddle dams) and Cherokee Dams Advisory Committee on Reactor Safeguards l 33
Site Overview 171 Overview of Tennessee River System Clinch River Watershed Site Details Orients Clinch River Site Relative to Tennessee River and Other TVA Nuclear Plants Advisory Committee on Reactor Safeguards l 34
Tennessee River System 172 Advisory Committee on Reactor Safeguards l 35
TVA Water Control System 173 Advisory Committee on Reactor Safeguards l 36
Clinch River Site 174 Planned finish grade elevation is 821 ft Nominal Clinch River elevation at site varies between 735 and 740 ft (seasonally)
Advisory Committee on Reactor Safeguards l 37
ESPA - SSAR Section 2.4 Development 175 Section 2.4 - Hydrologic Engineering ESPA SSAR Section 2.4 describes the hydrological characteristics of the Clinch River Nuclear Site.
This section addresses hydrologic characteristics and natural phenomena that have the potential to affect the design basis for the surrogate plant.
The section is divided into fourteen subsections describing the following hydrological characteristics:
2.4.1 - Hydrologic Description 2.4.8 - Cooling Water Canals and Reservoirs 2.4.2 - Floods 2.4.9 - Channel Diversions 2.4.3 - Probable Maximum Flood on Streams and Rivers 2.4.10 - Flooding Protection Requirements 2.4.4 - Potential Dam Failures 2.4.11 - Low Water Considerations 2.4.5 - Probable Maximum Surge and Seiche Flooding 2.4.12 - Groundwater 2.4.6 - Probable Maximum Tsunami Hazards 2.4.13 - Accidental Release of Radioactive Liquid Effluent in Groundwater and Surface Waters 2.4.7 - Ice Effects 2.4.14 - Technical Specification and Emergency Operation Requirements Advisory Committee on Reactor Safeguards l 38
Hydrologic Characteristics Demonstrated to 176 have no Safety-Related Impact Subsection 2.4.2 - Floods
- Preliminary plant grade is well above the calculated maximum flood level.
Subsection 2.4.7 - Ice Effects
- Due to the relatively mild climatic condition at the Clinch River Nuclear Site, and the elevated design plant grade above natural drainages, in combination with the SMR plant design that does not rely on external water sources for safety-related water use, it is concluded that ice effects will not cause flooding or water availability concerns.
Advisory Committee on Reactor Safeguards l 39
Hydrologic Characteristics Demonstrated to 177 have no Safety-Related Impact Subsection 2.4.9 - Channel Diversions
- A review of hydrologic, hydraulic, climatic, topographic and geologic evidence and anthropogenic impacts on the Clinch River arm of the Watts Bar Reservoir near the Clinch River Nuclear Site indicates that channel diversions are not expected in the Clinch River during the operating life of the plant.
Advisory Committee on Reactor Safeguards l 40
Hydrologic Characteristics Demonstrated to 178 have no Safety-Related Impact Subsection 2.4.10 - Flooding Protection Requirements
- No adverse impacts to the function of safety-related and risk-significant SSCs at the CRN Site are expected during the design basis extreme flooding event and the local intense precipitation event.
Advisory Committee on Reactor Safeguards l 41
2.4 Subsections Demonstrated to have no 179 Safety-Related Impact Subsection 2.4.13 - Accidental Releases of Radionuclides in Ground and Surface Waters
- Radwaste tank rupture releases 80% (per BTP 11-6) of contents instantaneously into groundwater outside containment.
- Source is based on 1% failed fuel (BTP 11-6 suggests 0.12%).
- Groundwater transport is based on shortest travel distance from release point to Clinch River (1400 ft).
- The resulting total dose from all exposure pathways meets 10 CFR 20.1301 limit of 100 mrem TEDE.
Advisory Committee on Reactor Safeguards l 42
Hydrologic Characteristics Demonstrated to 180 be an Unlikely Hazard at Site Subsection 2.4.5 - Probable Maximum Surge and Seiche Flooding
- Because the site is not located on an open or large body of water, surge or seiche flooding will not produce the maximum water levels at the site.
Subsection 2.4.6 - Probable Maximum Tsunami Hazards
- The Clinch River Nuclear Site is located more than 300 miles from the nearest seacoast. In addition, the site finish grade elevation is at 821 feet above sea level.
Thus, the site is not subject to any tsunami events originated from the ocean.
Advisory Committee on Reactor Safeguards l 43
Hydrologic Characteristics Demonstrated to 181 not be Applicable due to Design Subsection 2.4.8 - Cooling Water Canals and Reservoirs
- The small modular reactors under consideration at the Clinch River Nuclear Site do not rely on the Clinch River arm of the Watts Bar Reservoir for a safety-related water supply, and the site does not include cooling water canals or reservoirs.
Subsection 2.4.11 - Low Water Considerations
- The Ultimate Heat Sink for the Clinch River Nuclear Site does not rely on the Clinch River arm of the Watts Bar Reservoir to perform its function.
Advisory Committee on Reactor Safeguards l 44
Hydrologic Characteristics Demonstrated to 182 not be Applicable due to Design Subsection 2.4.14 - Technical Specifications and Emergency Operation Requirements
- The current designs of the small modular reactors being evaluated for deployment at the Clinch River Nuclear Site do not require use of a safety-related source of cooling water from the Clinch River arm of the Watts Bar Reservoir, and thus related technical specifications or emergency operation requirements are not necessary.
Advisory Committee on Reactor Safeguards l 45
183 Subsection 2.4.3 -
Probable Maximum Flood on Stream and Rivers Subsection 2.4.4 -
Potential Dam Failures Advisory Committee on Reactor Safeguards l 46
Flooding Guidance 184 NRC Regulatory Guide 1.59, Design Basis Floods for Nuclear Power Plants, supplemented by best current practice Hydrometeorological Reports (HMRs) 41, 51, 52 and 56 Previous watershed specific guidance from National Weather Service (NWS)
ANSI/ANS 2.8-1992 (W2002), Determining Design Basis Flooding at Power Reactor Sites NUREG/CR-7046, Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America Advisory Committee on Reactor Safeguards l 47
Dam Failure Guidance 185 NRC Regulatory Guide 1.59 JLD-ISG-2013-01, Guidance for Assessment of Flooding Hazards Due to Dam Failure NUREG/CR-7046 ANSI/ANS 2.8-1992 (W2002)
Advisory Committee on Reactor Safeguards l 48
CRN Simulations 186 Probable Maximum Precipitation (PMP) based on HMRs applicable to basin size and location Inflows - 100% runoff and unit hydrographs adjusted for non-linear basin response USACE HEC-RAS software utilized Downstream project (Watts Bar Dam) was assumed stable to maximize CRN impacts Dam stability determined by TVA Dam Safety Organization Advisory Committee on Reactor Safeguards l 49
Controlling Flood Simulation 187 Probable Maximum Flood (PMF) was found to produce the highest calculated water surface elevation at the CRN site Seismically induced and sunny day dam failure simulations were performed but were not controlling PMF and seismic simulation results show CRN is a dry site with significant margin Local Intense Precipitation (LIP) will be evaluated at COLA Advisory Committee on Reactor Safeguards l 50
188 Subsection 2.4.12 - Groundwater Advisory Committee on Reactor Safeguards l 51
Groundwater Investigation Outline 189 Regional Hydrogeology Local Hydrogeology: Conceptual Model Site-Specific Data From the Clinch River Breeder Reactor Project (CRBRP)
Site-Specific Data From the Clinch River Nuclear (CRN) Site Groundwater Flow Directions Geological Cross Section Post-Construction Groundwater Model CRN Site: Groundwater Use Construction Dewatering Source: SSAR Figure 2.4.12-1 Advisory Committee on Reactor Safeguards l 52
Regional Hydrogeology 190 Source: SSAR Figure 2.4.12-7 Advisory Committee on Reactor Safeguards l 53
Local Hydrogeology: Conceptual Model 191 Local hydrogeology is based on information from the adjacent ORR and the CRN Site:
Source: SSAR Figure 2.4.12-12 Advisory Committee on Reactor Safeguards l 54
Site-Specific Hydrogeology: CRBRP Site 192 Interpretations Groundwater levels fluctuate as much as 20 ft - response to precipitation events Groundwater flows from topographically high areas (center of the peninsula) to topographically low areas (Clinch River arm of the Watts Bar Reservoir)
Chestnut Ridge to the north acts as a groundwater divide Source: SSAR Figure 2.5.4-1 Advisory Committee on Reactor Safeguards l 55
Site-Specific Hydrogeology: Clinch River Site 193 Interpretations Groundwater flow is predominantly along the fractures and joints - with active flow primarily at shallow depths (interface of soil and weathered bedrock)
Predominant groundwater flow occurs along the strike of the bedding plane at N520E Frequency of fractures/joints decreases significantly with depth - predominant flow is at shallow depth, i.e.,
elevation 812 to 712 ft Clinch River acts as a sink for the shallow flow zone Pumping test radius of influence limited to approximately 150 ft from the pumping well Source: SSAR Figure 2.4.12C-4 l 56 Advisory Committee on Reactor Safeguards
Groundwater Flow Directions 194 General groundwater flow direction toward the southeast or southwest in the area of the proposed nuclear island Dominant downward flow at the center of the peninsula and upward at the Clinch River Equipotential Lines in the Vertical Plane (Along Strike): June 13th, 2014 Potentiometric Surface: February 12th, 2015 Source: SSAR Figure 2.4.12-29 Source: SSAR Figure 2.4.12-26 l 57
Geological Cross Section of Clinch River Site 195 Source: SSAR Figure 2.5.1-30 Advisory Committee on Reactor Safeguards l 58
Post-Construction Groundwater Model: 196 Maximum Groundwater Levels 802.3 to 816.1 ft NAVD88 Site Grade of 821 ft NAVD88 Source: SSAR Figure 2.4.12C-23 Source: SSAR Figure 2.4.12C-27 Source: SSAR Figure 2.4.12C-13 Advisory Committee on Reactor Safeguards l 59
CRN Site: Groundwater Use 197 Proposed CRN Site SMR designs do not rely on groundwater for plant operations Potable and other water will come from the Oak Ridge Department of Public Works Makeup water for the closed-cycle cooling system will be sourced from the Clinch River arm of Watts Bar Reservoir Advisory Committee on Reactor Safeguards l 60
Construction Dewatering 198 No permanent dewatering system will be employed Temporary dewatering will be required during excavation Temporary dewatering based on similar techniques as in CRBRP excavation Flow rate will be minimal - as observed in CRBRP excavation Advisory Committee on Reactor Safeguards l 61
Groundwater Investigation Conclusion 199 The proposed CRN Site SMR designs do not rely on groundwater for operations.
Permanent dewatering is not required.
Maximum groundwater levels range between 802.3 to 816.1 ft NAVD88, below CRN Site grade of 821 ft NAVD88 .
Advisory Committee on Reactor Safeguards l 62
200 Advisory Committee on Reactor Safeguards Advisory Committee on Reactor Safeguards l 63
201 Presentation to the ACRS Clinch River Nuclear (CRN) Site Early Site Permit (ESP) Application Chapter 11 Radioactive Waste Management Sections 11.2.3 & 11.3.3 Safety Evaluation Review Presented by Richard Clement November 14, 2018
Sections 11.2.3 & 11.3.3 202
- Involves source term information and offsite doses that include:
- Liquid effluent releases (Section 11.2.3)
- Liquid exposure pathways (Section 11.2.3.1)
- Liquid effluent doses (Section 11.2.3.2)
- Gaseous effluent releases (Section 11.3.3)
- Gaseous exposure pathways (Section 11.3.3.1)
- Gaseous effluent doses (Section 11.3.3.2)
- Review interface with hydrology (Section 2.4.13) and meteorology (Section 2.3.5) 2
Key Review Areas 203
- Staff participated in the Pre-application Readiness Assessment and Acceptance Review.
- Staff conducted an audit at Bechtel Power Corporation, Tennessee Valley Authority (TVA) Knoxville Complex, Clinch River Nuclear (CRN) Site and surrounding areas (ML17341A276):
- Normal plant parameter envelope (PPE) liquid and gaseous effluent release source terms and offsite doses
- Accident PPE liquid effluent release source term and dose
- CRN site tour and current receptor locations
- Staff conducted an audit of TVAs voluntary submittal on vector- and scalar-averaged wind direction and scalar-averaged wind speed data (ML18248A113):
- Offsite gaseous effluent dose and receptor information 3
PPE Source Terms 204
- BWXT mPower (Generation mPower)
- NuScale (NuScale Power)
- SMR-160 (Holtec SMR)
- Westinghouse SMR (Westinghouse Electric Co.)
- TVA used Nuclear Energy Institute 10-01 to evaluate composite source terms in the surrogate plant and develop the normal PPE liquid and gaseous effluent release source terms.
- Staff performed confirmatory calculations of normal PPE liquid and gaseous effluent release source terms.
- Staff confirmed that the unity rule in 10 CFR Part 20, Appendix B, Table 2, Columns 1 and 2 was met.
- Staff found TVAs methodology to develop the normal PPE liquid and gaseous effluent release source terms for use in calculating offsite doses was reasonable.
4
Dose Evaluation 205
- Staff verified the input parameters and assumptions for exposure pathway dose analyses.
- Staff performed confirmatory calculations of offsite doses using Regulatory Guide 1.109 and NRCDose 2.3.20 computer code.
- Staff identified COL Action Item:
COL Action Item 11-1 An applicant for a combined license (COL) or a construction permit (CP) referencing this early site permit (ESP) should verify that the calculated doses to members of the public from normal gaseous and liquid effluent releases for a chosen reactor design at the CRN Site are bounded by the doses evaluated in this ESP application as reviewed by the NRC staff.
The applicant should evaluate discrepancies and justify any changes made to address differences in the source term for the reactor design used to calculate the doses for a COL or CP application.
5
Conclusions 206
- Staff completed its Safety Evaluation with no Open Items.
- Normal PPE liquid and gaseous effluent release concentrations meet the unity rule in 10 Code of Federal Regulations (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 radiation standards in 40 CFR Part 190, as implemented under 10 CFR 20.1301(e); and public dose limit in 10 CFR 20.1301.
- Subject to the staffs proposed condition (COL Action Item 11-1), reactor designs falling within the normal PPE effluent release source terms and offsite doses for the CRN site are without undue risk to public health and safety.
6
207 Questions?
7
Acronyms 208 CFR - Code of Federal Regulations COL - Combined License CP - Construction Permit CRN - Clinch River Nuclear ESP - Early Site Permit NRC - Nuclear Regulatory Commission NRCDose - Code system which contains three NRC endorsed computer codes used for exposure pathway dose analysis PPE - Plant Parameter Envelope SMR - Small Modular Reactor TVA - Tennessee Valley Authority 8
209 Staff Presentation to ACRS Subcommittee Clinch River Early Site Permit Application SER Chapter 2, Site Characteristics Section 2.3 - Meteorology Kevin Quinlan
210 Chapter 2, Section 2.3 - Meteorology Involves site specific information such as:
- regional climatology (2.3.1)
- local meteorology (2.3.2)
- onsite meteorological measurements program (2.3.3)
- short-term atmospheric dispersion estimates for accidental releases (2.3.4)
- long-term atmospheric dispersion estimates for routine releases (2.3.5) 2
211 2.3.1 Regional Climatology Staff performed review and analysis for the following -
- Tornado/Hurricane Wind Speeds and Associated Missiles
- Staff confirmed the applicants site characteristic values were appropriately derived from RG 1.76 and RG 1.221
- 100-year return Wind Speed (3-second gust)
- Staff confirmed the applicants site characteristic values were appropriately derived using ASCE/SEI 7-05
- Maximum Winter Precipitation
- Staff confirmed the applicants site characteristic values were appropriately derived using DC/COL-ISG-007 methodology
- Ambient Air Temperature and Humidity
- Staff independently confirmed the applicants site characteristic values using NWS data from Chattanooga, TN
- Staff concludes that the identification and consideration of the climatic site characteristics are acceptable and meet the requirements of 10 CFR 52.17(a)(1)(vi), 10 CFR 100.20(c), and 10 CFR 100.21(d) 3
212 2.3.2 Local Meteorology
- Staff reviewed and verified that the local meteorological data provided by Clinch River are representative of the site area as impacted by local topography.
- NRC Staff reviewed the Clinch River analysis of the following atmospheric phenomena recorded at the CRN site:
- Onsite wind speed and direction
- Atmospheric stability
- Ambient temperature and humidity
- NRC Staff also confirmed information recorded at offsite locations (such as National Weather Service reporting stations)
- Precipitation
- Fog
- Air quality and potential influence of the plant and related facilities on local meteorology 4
213 2.3.2 Local Meteorology (contd)
- Staff concludes that the applicants identification and consideration of the meteorological, air quality, and topographical characteristics of the site and the surrounding area meet the requirements of 10 CFR 100.20(c),
and 10 CFR 100.21(d), and are sufficient to determine the acceptability of the site.
5
214 2.3.3 On-site Meteorological Measurements Program
- Staff held an audit at the Clinch River site and surrounding area on May 15-17, 2017
- Audit topics related to meteorological monitoring included:
- Location and exposure of previously sited meteorological instrumentation and tower
- Instrument maintenance
- Data quality assurance program
- NRC staff completed a quality assurance review of the onsite meteorological database submitted by TVA as part of the ESP application.
- Staff confirmed that the TVA meteorological tower conformed to RG 1.23 criteria for siting of the tower in relation to the proposed Clinch River site 6
215 2.3.3 On-site Meteorological Measurements Program
- The SSAR used vector-averaged wind direction data as input to the straight-line Gaussian dispersion models (such as PAVAN and XOQDOQ). The applicant chose an alternative method to the best practice guidance cited in RG 1.23 and ANSI Standard 3.11-2005 which states that the transport wind direction for straight-line Gaussian models should be based on the scalar mean (or unit vector) wind direction.
- TVA voluntarily provided a submittal on April 9, 2018 (ML18100A950),
which evaluated the effects of having used vector-averaged wind directions in lieu of using scalar-averaged wind directions for the accident and routine release atmospheric dispersion estimates and the resulting doses presented in SSAR Chapters 15 and 11.
- TVAs analysis showed that the dose modeling results were bounding based on the use of vector-averaged wind directions. However, the applicant acknowledged that atmospheric dispersion and deposition factors for routine radiological releases were greater in some directions and lower in others when compared to using scalar-averaged wind directions.
7
216 2.3.3 On-site Meteorological Measurements Program
- TVA concluded that for normal and accident gaseous release dose assessments, the existing dose analyses included in the ESP application, which are based on vector-averaged wind directions and scalar-averaged wind speeds, is conservative and remains the basis of the CRN Site ESP application.
- NRC staff conducted an audit of this voluntary submittal (ML18248A113) to evaluate the potential implications of the applicants use of vector-averaged wind directions as input to the dispersion modeling analyses and wind-related data summaries.
- Staff audited CRNS atmospheric dispersion and dose analyses and agrees with the applicants conclusion.
- The staff concluded that the onsite meteorological monitoring system provides adequate data to represent onsite meteorological conditions as required by 10 CFR 100.20 and 10 CFR 100.21 8
2.3.3 On-site Meteorological 217 Measurements Program The staff proposed COL Action Items as stated below:
COL Action Item 2.3-2: An applicant for a COL or a CP referencing this ESP should verify that the onsite meteorological measurement system, including the instrument tower, expected at the site prior to operation, is as described in SSAR Section 2.3.3. Any differences in instrumentation, exposure, or siting should be identified and discussed in order to demonstrate that the meteorological measurements program continues to meet the guidance provided in RG 1.23.
COL Action Item 2.3-3: An applicant for a COL or a CP referencing this ESP should verify whether the operational phase of the onsite meteorological measurements program will include wind data averaging on the basis of scalar or vector averages.
COL Action Item 2.3-4: An applicant for a COL or a CP referencing this ESP should identify and justify the wind speed and direction averaging approach(es) (either vector or scalar) to be used in the COL or CP:
- for modeling accident-related Control Room and Technical Support Center (TSC) atmospheric dispersion; and
- to be used during the operational phase to support emergency planning.
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218 2.3.4 Short-Term (Accident)
Diffusion Estimates
- Staff performed an independent verification of the applicants accident diffusion estimates
- Staff created a Joint Frequency Distribution (JFD) from the onsite meteorological data for input to the PAVAN atmospheric dispersion computer model
- Staff executed its PAVAN computer model and generated offsite dispersion estimates (X/Q) values for all sectors along the uniform analytical Exclusion Area Boundary (EAB) (1100 feet) and the Low Population Zone (LPZ) (5279 feet) boundary
- The staff found the applicants EAB & LPZ site characteristic X/Q values acceptable
- The staff concludes that the applicant has established site characteristics and design parameters acceptable to meet the requirements of 10 CFR 52.17(a)(1)(ix), 10 CFR 100.21(c)(2), and 10 CFR 100.20(c) 10
SSAR Figure 2.3.4-1. Effluent Release 219 Boundary with Analytical EABs 11
2.3.5 Long-Term (Routine) 220 Diffusion Estimates
- Staff performed an independent verification of the applicants routine release diffusion estimates
- Staff created a JFD from the onsite meteorological data for use as part of the input into the XOQDOQ atmospheric dispersion computer model
- Staff executed the XOQDOQ computer model and generated atmospheric dispersion and deposition estimates (X/Q and D/Q) for receptors of interest
- Staff concludes that representative atmospheric dispersion and deposition conditions have been calculated for receptors of interest. The characterization of atmospheric dispersion and deposition conditions meet the requirements of 10 CFR 100.21(c)(1) and are appropriate for the evaluation to demonstrate compliance with 10 CFR Part 50, Appendix I.
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221 Conclusion
- All regulatory requirements for Section 2.3 have been satisfied
- No open items
- Three confirmatory items 13
222 Questions?
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223 Acronyms
- CFR - Code of Federal Regulations
- COL - combined license
- CP - construction permit
- DC/COL-ISG - Interim Staff Guidance for design certifications and combined licenses
- D/Q - atmospheric deposition factor
- EAB - exclusion area boundary
- ESP - early site permit
- JFD - joint frequency distribution
- LPZ - low population zone
- RG - Regulatory Guide
- SSAR - Site Safety Analysis Report
- X/Q - atmospheric dispersion factor 15
224 Presentation to the ACRS Subcommittee Safety Review of the Clinch River Nuclear Site, Early Site Permit Application Quality Assurance Program
Description:
(SSAR Section 17.5)
Presented by Nicholas Savwoir, Reactor Operations Engineer NRO/DCIP/QVIB-1 November 14, 2018 1
225 Early Site Permit (ESP) Regulations Appendix B to Part 50, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR Part 52.17, Contents of applications; technical information Subsections (a)(1)(xi) and (xii) 2
226
Background
TVA (Tennessee Valley Authority) submitted Operating Nuclear Quality Assurance Plan (NQAP), Revision 32, with their ESP application TVA NQAP, Revision 32, commits to ANSI N45.2-1971 as endorsed by RG 1.28, Revision 3 Review involved multiple public meetings and clarification calls One request for additional information (RAI) with 8 questions; TVA responded by submitting NQAP, Revision 36 3
227 Key Review Areas
[1] Quality Assurance Program Description (QAPD)
Criterion I - Organization Criterion II- Quality Assurance (QA) Program
[2] Quality Assurance (QA)
Gap analysis evaluation Criterion XVII- QA Records Criterion VII- Control of Purchased Material, Equipment and Services Criterion XV-Nonconforming Material Parts of Components
[3] QA Implementation Inspection April 16-20th 2018 at TVA (Chattanooga, TN) 4
228 Key Review Area [1]
[1] QAPD Clinch River Nuclear Site, Criterion I and Criterion II NRC Staff RAI:
Small Modular Reactor (SMR) Organization for the Clinch River Nuclear (CRN) Site Independent Assessments at the CRN site Reference or commitment to 10 CFR Part 52 As a result of the staffs review; TVA revised the NQAP to Revision 36:
Added Appendix K (roles and responsibilities) and Appendix L (organization chart) in support of the SMR organization.
Added Independent Assessments at the CRN site.
Added 10 CFR Part 52 to NQAP.
5
229 Key Review Area [2]
[2] QA Gap Analysis and Criterion XVII NRC Staff RAI:
Gap Analysis evaluation between RG 1.28 Rev 3 & 4 (10 CFR 52.17(a)(1)(xii))
CRN QA record documents CRN electronic records controls As a result of the staffs review; TVA revised the NQAP to Revision 36:
TVA provided a gap analysis evaluation during inspection (ML18143B478)
Added Appendix M (Clinch River Commitments and Clarifications for the ESP QA Program) and committed to RG 1.28 Rev 4.
Identified the documents that are considered QA records per Criterion XVII Added electronic records controls per RIS 2000-18 and NIRMA (Nuclear Information & Records Management Association), TG-11,15,16, and 21 6
230 Key Review Area [2]
[2] QA Gap Analysis, Criterion VII and Criterion XV NRC Staff RAI:
An incorrect exemption for the use of Accreditation in lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services Did not address the notification of affected organizations for nonconforming materials, parts or components As a result of the staffs review; TVA revised the NQAP to Revision 36:
Revised ILAC (International Laboratory Accreditation) conditions per NEI 14-05A Guidelines for the use of Accreditation in lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services. Revision, 0.
Added Appendix M and the commitments to address the notification of affected organizations.
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231 Key Review Area [3]
[3] QA Implementation Inspection April 16-20th, 2018 Tennessee Valley Authority (TVA) office in Chattanooga, TN Inspection Procedure (IP) 35017, Quality Assurance Implementation Inspection Initial review of TVA revised NQAP, Revision 36 No findings of significance were identified QA Inspection Report publicly-available (ML18143B478) 8
232 Conclusion QAPD for the CRN Site ESP application meets the requirements of 10 CFR Part 50, Appendix B and 10 CFR Part 52.17(a)(1)(xi) and (xii) .
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233 Presentation to the ACRS Clinch River Nuclear (CRN) Site Early Site Permit (ESP) Application Section 2.4 Hydrologic Engineering Safety Evaluation Review Presented by Yuan Cheng, Joseph Giacinto, Richard Clement November 14, 2018
CRN Site Location 234 2
CRN Site Overview 235
- Approximately 935 acres of land owned by the United States and operated by TVA
- Within Valley and Ridge Province
- Former Clinch River Breeder Reactor Project Site
- Proposed site grade of 821.0 ft 3
CRN Site PPE 236
- The applicant identified four small modular reactor (SMR) technologies for development of a plant parameter envelope (PPE):
- BWXT mPower (Generation mPower)
- NuScale (NuScale Power)
- SMR-160 (Holtec SMR)
- Westinghouse SMR (Westinghouse Electric Co.)
4
Staff Review 237
- Staffs review included a pre-application readiness assessment, acceptance review and, site visit and audit
- Staff worked in cooperation with U.S. Department of Energy (DOE), Tennessee Department of Environment and Conservation (TDEC) and the U.S. Geological Survey (USGS)
- Staff completed the safety evaluation report with no open Items 5
Probable Maximum Flood 238
- Staff reviewed the riverine flooding considering:
- Probable maximum precipitation
- Surface runoff hydrology
- Upstream dam failures with flood waves
- Sensitivity study related to modeling flood elevations
- Staff confirmed the maximum flood level computed by riverine hydraulic modeling with conservatisms including:
- 100 percent rainfall depth converted into surface runoff
- Instantaneous dam failure without breach formation time
- Maximizing backwater effect at the CRN site
- Resulting maximum flood level is significantly below site grade 6
Local Intense Precipitation 239
- Site drainage design:
- A site drainage design and site grading plan in combined license application is required to evaluate local intense precipitation (LIP) effects. Therefore, staff proposed COL Action Item 2.4-1.
- COL Action Item 2.4-1:
An applicant for a combined license (COL) or construction permit (CP) that references this early site permit should design the site grading to provide flooding protection to safety-related structures at the ESP site based on a comprehensive flood water routing analysis for a local intense precipitation (LIP) event.
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Flood Protection 240
- Flood protection evaluations:
- The flood protection should be evaluated in the COLA after a reactor technology and associated site grading plan are determined by the applicant. Therefore, staff proposed COL Action Item 2.4-2.
- COL Action Item 2.4-2:
An applicant for a Combined Operating License (COL) or Construction Permit (CP) referencing this Early Site Permit (ESP) should address whether the local flood elevation exceeds the site grade elevation and whether the local flood elevation needs to be incorporated with flood protection measures to prevent flooding of any safety-related Structures, Systems and Components (SSCs).
If so, the applicant should address necessary flooding protection for safety-related SSCs based on the flooding event and associated effects.
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Groundwater 241
- Staff reviewed two excavation geometries: a deep (681 ft. maximum) and a shallow (770 ft. maximum) elevation
- Staff confirmed maximum groundwater level of 816.1 ft. is reasonable
- Backfill properties determined for the COL, therefore staff proposed a directive for COL Action Item 2.5-8.
- COL Action Item 2.5-8:
An applicant for a COL or CP application referencing this early site permit should provide detailed design of backfill materials including identification of sources and quantity requirements, backfill material property and placement specifications, applicable industry standards, as well as related ITAAC. The in-place backfill hydraulic characteristics such as permeability and porosity should be consistent with those specified in the SSAR. If differences exists, the effect on the site conceptual model and site characterization as described in the SSAR should be evaluated. Geologic mapping of the final exposed surface after excavation is required before placement of backfill, and should be conducted under the guidelines of NRC requirements.
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Groundwater 242
- Staff noted that TDEC analyses of CRN Site groundwater samples indicate low levels of radionuclides
- Therefore, staff proposed COL Action Item 2.4-3.
- COL Action Item 2.4-3:
An applicant for a combined license (COL) or construction permit (CP) that references this early site permit will establish, as part of its plan to minimize contamination in accordance with 10 CFR 20.1406, a baseline for background radionuclide concentrations.
10
Surface and Ground Water 243 Findings
- Staff confirmed that the applicant considered most severe natural phenomena that have been historically reported for the site and surrounding area
- Staff confirmed that the design-basis flood elevation estimate, including the considerations of hypothetical dam failure and wind induced wave height, is sufficiently below site grade (821.0 ft).
- Staff confirmed that maximum groundwater level (816.1 ft) is approximately 5 ft below site grade
- Staff determined that site characteristics are bounded by plant parameter envelope design parameters 11
PPE Source Term 244
- Staff reviewed the basis and assumptions for developing the accident PPE liquid effluent release source term:
- Source term information for surrogate plant evaluated from two vendors with preliminary designs
- One percent failed fuel fraction (verses 0.12 percent in Branch Technical Position [BTP] 11-6) applied in one vendors source term
- CRN Site ESP application and Public Service Enterprise Group ESP PPE source terms compared
- Staff performed confirmatory calculations to verify the accident PPE liquid effluent release source term.
- Staff found TVAs methodology for developing the PPE source term to bound the dose to members of the public from a postulated accidental liquid effluent release to the groundwater reasonable.
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Radionuclide Transport 245
- Staff reviewed transport values and assumptions, and performed confirmatory calculations using NUREG/CR-3332 and BTP 11-6:
- Site-specific radionuclide transport values
- No credit for mitigating design features
- 80 percent of tank volume released
- Instantaneous release into groundwater
- Peak radionuclides and daughter product concentrations
- Minimum dilution flow of 400 cubic feet per second to Clinch River
- Minimal travel distance and decay
- Staff found TVAs methodology for estimating initial radionuclide concentrations from a postulated accidental liquid effluent release to the groundwater reasonable.
- Staff confirmed that the unity rule in 10 CFR Part 20, Appendix B, Table 2, Column 2 was met (considering sorption and retardation).
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Dose Evaluation 246
- Staff found TVAs methodology for estimating dose from a postulated accidental liquid effluent release to the groundwater using Regulatory Guide 1.109, Environmental Protection Agencys Federal Guidance Reports 11 and 12, and LADTAP II computer code reasonable.
- Staff confirmed that the public dose limit of 100 millirem total effective dose equivalent in 10 CFR 20.1301 was met.
- Staff identified COL Action Item:
COL Action Item 2.4-4 An applicant for a combined license (COL) or a construction permit (CP) referencing this early site permit (ESP) should verify that the calculated dose to members of the public from a postulated accidental liquid radionuclide effluent release to the groundwater from a chosen reactor design at the CRN Site is bounded by the dose evaluated in this ESP application as reviewed by the NRC staff. The applicant should evaluate discrepancies and justify any changes made to address differences in the source term for the reactor design used to calculate the dose for a COL or CP application.
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Staff Conclusions 247
- Staff proposed site characteristics and bounding design parameters for inclusion in the ESP.
- CRN ESP site characteristics meet requirements of 10 CFR Part 100, Reactor Site Criteria and 10 CFR Part 20, Standards for Protection Against Radiation.
- Subject to the staffs proposed conditions (COL Action Items 2.4-1, 2.4-2, 2.4-3, 2.4-4, and 2.5-8), technologies falling within the PPE design parameters for the CRN site characteristics are without undue risk to public health and safety.
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248 Questions?
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Acronyms 249 CFR - Code of Federal PPE - Plant Parameter Envelope Regulations SMR - Small Modular Reactor COL - Combined License SSCs - Structures, Systems and CP - Construction Permit Components CRN - Clinch River Nuclear TDEC - Tennessee Department DBF - Design Basis Flood of Environment and Conservation DOE - Department of Energy TVA - Tennessee Valley Authority ESP - Early Site Permit USGS - U.S. Geological Survey LADTAP - Liquid Annual Doses To All Persons NRC - Nuclear Regulatory Commission 17