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{{#Wiki_filter:Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
{{#Wiki_filter:}}
 
==Title:==
Advisory Committee on Reactor Safeguards Radiation Protection and Nuclear Materials Docket Number:    (n/a)
Location:        teleconference Date:            Wednesday, March 16, 2022 Work Order No.:  NRC-1885                          Pages 1-135 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
 
1 1
2 3
4                              DISCLAIMER 5
6 7  UNITED STATES NUCLEAR REGULATORY COMMISSIONS 8        ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9
10 11          The contents of this transcript of the 12 proceeding of the United States Nuclear Regulatory 13 Commission Advisory Committee on Reactor Safeguards, 14 as reported herein, is a record of the discussions 15 recorded at the meeting.
16 17          This transcript has not been reviewed, 18 corrected, and edited, and it may contain 19 inaccuracies.
20 21 22 23 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433          WASHINGTON, D.C. 20005-3701  www.nealrgross.com
 
1 1                    UNITED STATES OF AMERICA 2                  NUCLEAR REGULATORY COMMISSION 3                                + + + + +
4            ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5                                    (ACRS) 6                                + + + + +
7          RADIATION PROTECTION AND NUCLEAR MATERIALS 8                              SUBCOMMITTEE 9                                + + + + +
10                                WEDNESDAY 11                            MARCH 16, 2022 12                                + + + + +
13                  The Subcommittee met via Teleconference, 14 at 9:30 a.m. EDT, David A. Petti, Chair, presiding.
15 16 COMMITTEE MEMBERS:
17            DAVID A. PETTI, Chair 18            RONALD G. BALLINGER, Member 19            VICKI M. BIER, Member 20            CHARLES H. BROWN, JR. Member 21            VESNA B. DIMITRIJEVIC, Member 22            GREGORY H. HALNON, Member 23            JOSE MARCH-LEUBA, Chairman 24            JOY L. REMPE, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309    www.nealrgross.com
 
2 1 ACRS CONSULTANTS:
2            DENNIS BLEY 3            STEPHEN SCHULTZ 4
5 DESIGNATED FEDERAL OFFICIAL:
6            WEIDONG WANG 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309    www.nealrgross.com
 
3 1                                CONTENTS 2
3 Call to Order and Opening Remarks . . . . . . . .                          4 4
5 Introductions, Mike Franovich, NRR/DRA                  . . . . .          6 6
7 Discussion of DG-1389 Guidance                  . . . . . . . . . 10 8            Mark Blumberg, NRR/DRA            . . . . . 10, 26, 109 9            John Parillo, NRR/DRA . . . . . . . . . 24, 63 10            Elijah Dickson, NRR/DRA . . . . .              31, 69, 107 11            Steve Jones, NRR/DSS          . . . . . . . . . . . 36 12            Paul Clifford, NRR/DSS            . . . . . . . . . . 74 13            Sean Meighan, NRR/DRA . . . . . . . . . .                    112 14 15 Industry Comments on Draft DG-1389 Guidance . .                        114 16            Frankie Pimental, NEI . . . . . . . . . .                    114 17            Greg Broadbent, Senior Staff Engineer 18            in Nuclear Analysis, Entergy                                116 19 20 Opportunity for Public Comment                  . . . . . . . .        131 21            Brian Magnuson      . . . . . . . . . . . . .              131 22 23 Meeting Adjourn . . . . . . . . . . . . . . . .                        135 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
4 1                    P R O C E E D I N G S 2                                                              8:30 a.m.
3                CHAIR PETTI:          (presiding)        The meeting 4 will now come to order.
5                This  is      a  meeting          of  the    Radiation 6 Protection and Nuclear Materials Subcommittee of the 7 Advisory Committee on Reactor Safeguards.
8                I'm    Dave        Petti,          Chairman    of        the 9 Subcommittee.      Members      in    attendance      are    Charles 10 Brown, Jose March-Leuba, Joy Rempe, Ron Ballinger, 11 Vicki Bier, and Greg Halnon.
12                Vesna, are you online?
13                MEMBER DIMITRIJEVIC:                Yes, I am.
14                CHAIR PETTI:          Okay, Vesna is here.                We 15 also have our consultants, Steve Schultz, who's here 16 with us, and I saw Dennis is online remotely.
17                Weidong Wang is the Designated Federal 18 Officer for this meeting.
19                As  posted      in    the      agenda    on  the      ACRS 20 website, the topic for today is to review Draft Reg 21 Guide DG-1389; Proposed Revision 1 to Reg Guide 1.183, 22 "Alternative Radiological Source Terms for Evaluating 23 Design Basis Accidents at Nuclear Power Reactors."
24                The Subcommittee will hear presentations 25 by and hold discussions with the NRC staff and other NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
5 1 interested persons regarding this matter.
2                    Rules    for      participation          in    all      ACRS 3 meetings, including today's, were announced in The 4 Federal Register on June 12th, 2019.
5                    The ACRS section of the U.S. NRC public 6 website provides our Charter, Bylaws, agendas, Letter 7 Reports,        and  full      transcripts            of  all    full      and 8 subcommittee meetings, including slides presented.
9 The meeting notice and agenda for this meeting were 10 posted there.
11                    We have received a request to make an oral 12 statement from a public member.                    So, that will happen 13 at the end of our discussions here.
14                    The Subcommittee will gather information, 15 analyze        relevant    issues      and      facts,    and    formulate 16 proposed positions and actions, as appropriate, for 17 deliberation by the full Committee.
18                    The phone lines have been opened to all 19 members        of  the    public        to      listen    in    on      the 20 presentations          and        the      committee        discussion.
21 Additionally, we have an MS Teams link available for 22 the public.
23                    There will be an opportunity for public 24 comment, and we have set aside time at the conclusion 25 of    the    prepared  presentations            and  discussion        for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
6 1 comments from members of the public attending or 2 listening to our meetings.
3                  A transcript of the meeting is being kept.
4 And      it    is  requested      that      the      speakers  identify 5 themselves        and  speak      with    sufficient      clarity      and 6 volume, so that they can be readily heard.
7                  Additionally,        participants        should      mute 8 themselves when not speaking.                      To mute or unmute, 9 please push *6.
10                  So, to start off the staff's presentation, 11 I call on NRC management to make opening remarks.
12                  MR. FRANOVICH:          Sound check.        I hope you 13 can hear me.
14                  CHAIR PETTI:        We can.
15                  MR. FRANOVICH:          Great.
16                  Good morning, Chairman Petti and ACRS 17 Members.
18                  I am Mike Franovich, and I serve as the 19 Director of the Division of Risk Assessment in NRR.
20                  I want to thank you for the opportunity 21 today for the staff to share advances in updating our 22 Regulatory        Guidance,        Regulatory          Guide  1.183,        on 23 alternative source terms.
24                  Based on the February 17th, 2022, ACRS 25 meeting discussion on the integration of source term NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
7 1 activities, we recognize that there has been a great 2 interest in learning more about the RG update.                        Since 3 like 2020, NRR has refocused its efforts on Reg Guide 4 1.183 revision, or what we now call DG-1389, including 5 forming a steering committee consisting of Division 6 Directors in NRR and the Office of Nuclear Regulatory 7 Research, and a staff working group.
8                  Some of the working group members also 9 made presentations or interacted with the Committee 10 during the February 17th meeting on the source-term-11 related topics.
12                  During the development of DG-1389, we have 13 been fully embracing regulatory transformations and 14 the      performance-shaping        factors          to  improve        our 15 efficiency and reliability as regulators. A key theme 16 is that our licensing activities, other regulatory 17 decisions, and backfit/forward-fit actions must be 18 risk-informed.          The      Commission's          2019  direction 19 reminded the staff that we are, and have been, able to 20 use risk-informed, performance-based approaches in our 21 work. This direction also serves as an accelerant for 22 transformation to become a more modern and risk-23 informed regulator.
24                  The 2019 Staff Requirements Memorandum, 25 SRM, commonly referred to as the NuScale block valve NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
8 1 SRM, draws upon longstanding practices, in particular, 2 a 1999 Commission paper that states succinctly, "a 3 risk-informed,          performance-based                  approach    is      an 4 approach in which risk insights, engineering analysis, 5 and judgment, including the principles of defense-in-6 depth        and    incorporation            of      safety      margins        and 7 performance history, are used in decisions."
8                    In    2019,        we        received        additional 9 clarifications from the Commission regarding treatment 10 of backfits and proper treatment of forward-fits.
11 This      Commission      direction          serves        as  a  regulatory 12 stabilizer, applying the reliable principles of good 13 regulation in our licensing activities.
14                    Secondly,        I  would        like    to  point      out 15 improved realism evaluation techniques and additional 16 information          are      applied        to      our      decisionmaking 17 activities.          The NRC's application of risk-informed 18 decisionmaking          continues          to    evolve,      as  improved 19 realism,          evaluation          techniques,            and    additional 20 information are applied to our decisions.
21                    What that means to the staff, when it 22 comes to update of our guidance for alternative source 23 terms, is that there are tremendous opportunities for 24 applying engineering and risk insights.                              This vast 25 amount        of  information          includes          plant    operating NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
9 1 experience and 21st century research, as well as our 2 experiences from post-Fukushima activities.                      All of 3 this effort is focused to make more realistic and, 4 ultimately, better decisions while abiding by the 5 Commission's backfit and forward-fit expectations.
6                Today, you will hear from the collective 7 strength of a diverse team with deep experience in 8 radiological accident analysis, reactor safety and 9 systems, seismic engineering, accident-tolerant fuel, 10 high burnup fuel, and risk assessment techniques and 11 practices as well.
12                I  do    want      to      express    my    deepest 13 appreciation to all of the staff who have contributed 14 to this Regulatory Guide.            It's quite comprehensive.
15 There are many updates that we will be going through 16 today.
17                There has been a significant amount of 18 work done by the staff, which you will hear about in 19 more depth -- in great depth, actually -- today in our 20 session.
21                We  look      forward        to    hearing  the      ACRS 22 members' views as we continue our track to collect 23 stakeholder perspectives on this draft guidance.
24                At  this      time,      I'm    going  to  turn      the 25 meeting over to Mark.          So, thank you very much.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
10 1                CHAIR PETTI:        Thank you.
2                Go ahead, Mark.
3                MR. BLUMBERG:              Thank      you  for        the 4 introduction.
5                We are pleased to be here today to provide 6 a presentation on a Draft Revision to Reg Guide 1.183, 7 known as DG-1389.        The names on the cover slide are 8 the names of the presenters for today's discussion.
9 These individuals represent the many individuals in 10 the steering committee and the working group that 11 developed DG-1389.
12                Let's get started on the presentation by 13 moving to slide 2.
14                Are the slides up?
15                MEMBER BIER:        No, we are not seeing any 16 slides in this.      There we go.            Thanks.
17                MR. BLUMBERG:        Okay.        Can you go to slide 18 2, please?
19                This slide provides the --
20                CHAIR PETTI:        Mark, hold on.
21                Please put it in Presentation Mode.                    It's 22 awful tiny on our very big screen.                    Thank you.
23                MR. BLUMBERG:          You're welcome.
24                This slide provides the agenda for our 25 presentation.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
11 1                  First,      we'll        give        you  introductory 2 information        containing          key        messages      for        our 3 presentation and some background information related 4 to the development of the Reg Guide 1.183 methods.
5                  Next, we will discuss several of the key 6 changes proposed in DG-1389.                In the presentation, we 7 will go into more details on several of these changes.
8 Those detailed discussions will be on proposed methods 9 to model the MSIV leakage pathway and on a new source 10 term for the maximum hypothetical loss of coolant 11 accident.        We will also discuss proposed new source 12 terms for the non-LOCA accidents.
13                  Next, we will discuss a change to the 14 fuel-handling accident methods.
15                  Lastly, we'll summarize some of the key 16 takeaways from our presentation and discuss what needs 17 to be done to complete the revise guidance, and then, 18 provide some time for discussion and any feedback that 19 you may have.
20                  Please go to slide 3 now.
21                  Before we get into our key messages, we 22 would like to acknowledge the 17 individuals involved 23 in      the    DG-1389  development.                As  Mike  said,        we 24 coordinated the proposed changes across management and 25 staff of multiple organizations within the NRC.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
12 1                  Our steering committee is comprised of 2 four Division Directors from the Office of Nuclear 3 Reactor Regulation and the Office of Nuclear Reactor 4 Research.
5                  Similarly, the working group is comprised 6 of staff from within those offices and the Office of 7 General Counsel.
8                  The steering committee and working group 9 provide        diversity    and    coordination        among    several 10 subject matter experts within the NRC.
11                  Let's now review our key messages for this 12 presentation.        Please go to slide 4.
13                  The first key message is that the NRC has 14 restarted in an effort to update Reg Guide 1.183, 15 Revision 0.        The purpose of Reg Guide 1.183 is to 16 provide methods and assumptions acceptable to the NRC 17 staff for showing compliance with the regulation known 18 as 10 CFR 50.67.
19                  As a result of a change to our Management 20 Directive in Handbook 8.4 on Backfitting, we are 21 proposing that Rev. 0 of Reg Guide 1.183 and Rev. 1 22 coexist, rather than have Rev. 1 supersede Rev. 0.
23                  Rather than finalizing a draft revision to 24 Reg Guide 1.183, known as DG-1199, the staff is 25 issuing a new draft revision known as DG-1389.                          1389 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
13 1 considers and addresses technical issues and public 2 comments related to the issuance of DG-1199.                    It also 3 incorporates new guidance based upon current issues 4 and lessons learned during the review of license 5 amendment requests using Reg Guide 1.183.
6                This restarted effort involved early and 7 frequent outreach to external stakeholders. The dates 8 and slides for three of these public meetings are 9 provided in bullet 3 on this slide.
10                Reg  Guide      1.183,        Rev. 1,  provides          a 11 preapproved method for modeling MSIV leakage.                    Use of 12 individual regulatory positions in isolation will need 13 additional justification.
14                Please go to slide 5.
15                Slide 5 discusses some of the objectives 16 of the proposed revision to Reg Guide 1.183.                          They 17 include the following:
18                Making the guidance more useful.
19                Incorporating                relevant        operating 20 experience.
21                Responding to changes in the regulatory 22 environment.
23                Incorporating            lessons        learned        from 24 previous reviews.
25                Providing guidance for changes in fuel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
14 1 utilization in accident-tolerant designs.
2                  And to support licensing of advanced light 3 water designs.
4                  And lastly, to incorporate insights from 5 new research activities.
6                  Please go to slide 6 now.
7                  CHAIR PETTI:        Mark?
8                  MR. BLUMBERG:          Yes?
9                  CHAIR PETTI:        Before you go on, go back.
10                  What is it about the NuScale SRM that 11 impacted this draft guide?
12                  MR. BLUMBERG:            If you can wait just a 13 little while, we'll be getting --
14                  CHAIR PETTI:        No problem.
15                  MR. BLUMBERG:          -- into detail.
16                  CHAIR PETTI:        Okay.        No problem.
17                  MR. BLUMBERG: And if we don't answer your 18 question, then we can bring it up at the end, and I'll 19 be happy to get into those details.
20                  DR. BLEY: And, Mark, this is Dennis Bley.
21 Could you back up one more?
22                  This is not a substantive issue, but in 23 Reg Guides, until a few years ago, there used to be a 24 section        called  "Regulatory          Positions."        You      guys 25 changed that and it has a new name.                      And I think, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
15 1 increasingly, referring to regulatory positions within 2 your own document that doesn't have a section named 3 "Regulatory    Positions"        will      become    more  and      more 4 confusing to people.
5                The  first      time      I    ran  across    it,      it 6 confused me because I didn't see anything labeled 7 "Regulatory Positions."          So, just a comment:            instead 8 of "positions," if you can come up with another word 9 that fits what you're calling the section now? That's 10 all.
11                MR. BLUMBERG:          Okay.        Thank you.
12                MEMBER HALNON:          Yes, Mark, this is Greg 13 Halnon.
14                One other question, and it's more probably 15 editorial. The coexistence of Rev. 0/Rev. 1, is it a 16 little unusual?      Is that an artifact of the document 17 control system, or did you consider just making it a 18 1.1a3a, or some other number?
19                MR. BLUMBERG:          No, it's more of an issue 20 that relates back to the SRM that we were talking 21 about before, the SECY-18-0049.                    It's a Management 22 Directive and Handbook on Backfitting and Finality of 23 Information Collection.
24                And when we went through this revised 25 change to the backfitting, we knew that we needed an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
16 1 effective program that would ensure backfitting and 2 forward-fitting for these facilities. And as a result 3 of looking at the previous positions, what we're 4 trying to do is we're saying that, rather than making 5 major changes that would override those that are 6 previously existing, in this particular instance, 7 we're proposing a completely revised method.
8                  Does that make sense?
9                  MEMBER HALNON:              Not entirely, but I'm 10 going to let it go.        I just felt like it could lead to 11 some other confusion down the road, but I'm sure 12 you've thought through that.                But if I come up with a 13 more coherent question that can make it clearer to me, 14 I will later on.
15                  MR. BLUMBERG:          And we can revisit it.
16                  MEMBER HALNON:          Thank you.
17                  MR. BLUMBERG:          Thank you.
18                  MEMBER BROWN:          This is Charlie Brown.
19                  I wanted to ask, springboarding off of 20 Greg's comment, how do they decide what they're going 21 to use?        Somewhere you've got to be telling people 22 which one you all accept more than one of the others.
23                  MR. BLUMBERG:        So, both provide different 24 methods.        In the instance of Rev. 1, as we'll go 25 through in this presentation, we'll be discussing the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309      www.nealrgross.com
 
17 1 benefits of using Rev. 1.                  And so, based upon their 2 needs -- so, say, for example, they needed a source 3 term for expanded burnup or enrichment, they may 4 choose to use Rev. 1 over Rev. 0 because it contains 5 a source term that's applicable to those changes.
6                    MEMBER BROWN: Okay. Let me just make one 7 observation relative to how you phrased that.                            When 8 you go through this, a lot of the -- I've forgotten 9 what        they're    called        --    the      tables  that      show 10 percentages of byproducts on whatever, they change.
11 You introduced a new accident called, I guess -- what 12 is it? -- maximum hypothetical accident.                        And then, 13 you threw in -- I can't remember the paragraph that 14 it's in, but you literally said, whatever you do when 15 you do your analysis, do not do that.                        Now, you've 16 added on and tacked on a separate part of it where it 17 says, but if you decide to, then go ahead and do it, 18 and you maybe can justify, whatever.
19                    So, that's at least three somewhat major 20 changes between the earlier version, Rev. 0, and this 21 one.        So, I'm just an electrical engineer.                        So, I 22 understand parts of -- physics seems to dictate source 23 terms,        and  it  just      seems      we've      got conflicting 24 directions on what percentages of source terms and how 25 they should be incorporated into the overall analysis.
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18 1                  So, I'm just echoing Greg and trying to 2 provide a little bit more meat from my perspective, or 3 from      one  Committee      member's        perspective,        on      the 4 differences between the two.                    So, I just throw that 5 out there.        It's a confusion to me.                And that's all I 6 wanted to make sure you understood.
7                  MR. BLUMBERG:          Okay.        Thank you.
8                  We'll be going through more details, and 9 hopefully, we can clear up the majority of that 10 confusion.
11                  DR. BLEY:        Hey, Mark, Dennis Bley again.
12                  Back to my editorial point.                    At places 13 during the document you refer to Regulatory Position 14 3, or whatever.        The Table of Contents, Section C is 15 "Regulatory positions," the old language; Section C 16 itself is labeled "Regulatory Guidance."                        So, maybe 17 I'm not the only one confused.                    That's all.
18                  MR. BLUMBERG:          I can't remember why that 19 particular change was made.                It seems like there were 20 changes in the formatting to Reg Guides.
21                  DR. BLEY:        Yes. Yes, I think that's it, 22 because        we've  had    several        other        Reg  Guides      come 23 through, and that language has been changed from 24 "Regulatory Positions" to "Regulatory Guidance."                            So, 25 somewhere you've got guidance telling you how to do NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
19 1 that, but it's not consistently carried out.
2                  MR. BLUMBERG: Okay. We'll take that note 3 back to our Reg Guide people and discuss the potential 4 for confusion with that.
5                  DR. SCHULTZ:        Mark?        Mark?
6                  MR. BLUMBERG:          Yes?
7                  DR. SCHULTZ:        This is Steve Schultz.
8                  Are you saying that, with regard to 1.183, 9 Rev. 0, that an applicant, a licensee can utilize that 10 to support extended burnup, increased enrichments, new 11 fuel designs, and that would be suitable for the staff 12 to receive such application under Rev. 0?
13                  MR. BLUMBERG:          So, Steve, the applicants 14 are always able to propose new methods to support 15 compliance with the regulations.                      Rev. 1 and Rev. 0 16 are      standalone    documents.                They    provide      those 17 preapproved methods. If one is to try to change those 18 methods, then additional justification is needed.
19                  So, with respect to Rev. 0, currently, 20 Rev. 0 does not apply to those extended burnups.                            It 21 only      goes  to  --    there's        a    footnote    within        the 22 regulatory guidance that goes out to, I think, 62,000 23 gigawatt days per metric ton uranium.                      The extended 24 burnups that are being sought are beyond that.                          What 25 Rev.      1  does  is    provides        a    source    term    that's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
20 1 applicable to that.            If an individual were to want to 2 use Rev. 0 for extended burnup, then that would 3 involve        more  justification,            rather    than    just        a 4 preapproved method.
5                    Does that make sense?
6                    DR. SCHULTZ: That's a good clarification.
7 Thank you.
8                    MR. BLUMBERG:          You're welcome.
9                    Okay. So, this part of the presentation 10 will discuss some background information.
11                    If we can go to slide 7, please?
12                    First, we would like to provide background 13 on the accident analysis at the NRC.                      The NRC exists 14 to protect the health and safety of the public from 15 accidental releases of fission products.                      One of the 16 ways the NRC and licensees determine what measures and 17 barriers        are  needed      is  to      perform    accident      dose 18 analysis.        Specifically, these analyses address the 19 situations where we are wrong about the success of the 20 facility's response to events or accidents.
21                    Dose analyses provide an effective way to 22 account        for  uncertainties          in    equipment  and    human 23 performance.        In particular, these analyses account 24 for the unlikely events that involve unknown and 25 unforeseen failure mechanisms or phenomena which are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
21 1 not reflected in either the PRA or the traditional 2 engineering analysis.
3                      The origin of the radiological consequence 4 analysis is based upon siting criteria contained in 5 10 CFR Part 100.                Part 100 criteria were used to 6 evaluate          the    defense-in-depth                provided    by      the 7 containment.          The source term for these analyses were 8 provided in a document known as Technical Information 9 Document 14844, or TID-14844, which was incorporated 10 into the Part 100 regulation.
11                      As a result of the TMI accident, the NRC 12 initiated severe accident research to learn from the 13 accident.          This research led to significant advances 14 in the understanding of the timing, the magnitude, and 15 the chemical form of the fission products released 16 from severe nuclear power plant accidents.
17                      When the revised source term was developed 18 using        that    research,        and      the      sequences      from 19 NUREG-1150, this revised source term was put into a 20 document known as NUREG-1465.                    That document provides 21 estimates of accident source terms that are more 22 physically-based and could be applied to the design of 23 future        light    water    reactors,          as    well  as  current 24 reactors.          Reg Guide 1.183, Version 0, adopted the 25 NUREG-1465 early in vessel fuel melt source term and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433              WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
22 1 provided guidance for operating power reactors.
2                  Can you go to slide 8 now, please?
3                  Reg Guide 1.183, Revision 0, was issued 21 4 years ago.        It provides assumptions and methods that 5 are        acceptable      for      performing            design      basis 6 radiological analyses using an alternative source term 7 that is an alternative to the TID-14844 source term 8 specified in Part 100.
9                  Because future research and changes in the 10 plant operation could occur, it was recognized that 11 there could be a need for a source term other than 12 1465.        So, within Reg Guide 1.183, we also included 13 guidance on acceptable attributes for source terms 14 developed in the future.
15                  Could you go to slide 9, please?
16                  In October of 2009, the staff incorporated 17 lessons learned from the use of Reg Guide 1.183 into 18 a draft revision known as DG-1199.                      Examples of other 19 reasons for revising Reg Guide 1.183 are also included 20 in this slide.        They'll be discussed in more detail 21 later in our presentation.
22                  We  received        150      public      comments        on 23 DG-1199.        These comments were used by the staff to 24 create the new revision known as DG-1389.
25                  Please go to slide 10.
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23 1                  The next two slides will discuss several 2 license amendments that were used to inform DG-1389.
3                  Since the issuance of DG-1199 for public 4 comment,      we  have    continued          to    process    license 5 amendments tied to Reg Guide 1.183.                      Several of these 6 license amendment requests are directly tied to the 7 non-LOCA gap fractions in Table 3 of Reg Guide 1.183.
8 Table 3 has limitations that are exceeded by modern 9 fuel utilization.            The license amendments request 10 listed here reinforced the need for expanding the 11 applicability      and    for    providing          new  non-LOCA      gap 12 fractions based upon modern fuel utilization.                            This 13 will      be  discussed    in    more      detail      later  in      the 14 presentation by Paul Clifford.
15                  If you'll go to slide 11 now, please?
16                  In 2019, the NRC received four alternative 17 source term amendments requesting increased allowable 18 MSIV leakage limits.            One of these license amendment 19 requests also requested removal of their main steam 20 leakage control system. As a result of this work, the 21 finalization      of  Reg      Guide      1.183      was  postponed, 22 allowing the NRC staff time to develop lessons learned 23 from these reviews.
24                  We have incorporated the lessons learned 25 from these reviews into DG-1389.                        This slide lists NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
24 1 those four license amendment requests to increase 2 their MSIV leakage.
3                Now I'd like to turn the presentation over 4 to John Parillo to discuss our next slide.
5                MR. PARILLO:        Thank you, Mark.
6                The thesis of the referenced differing 7 professional opinion is that licensees should be able 8 to credit safety-related systems to distribute the 9 deterministic substantial fuel melt source term within 10 the containment atmosphere at the times for which 11 these systems are designed to achieve operability.
12 Previous assumptions associated with the alternative 13 source term specified that mixing of the deterministic 14 source term should be delayed for two hours to allow 15 for the deterministic fuel source term described in 16 NUREG-1465 to develop.
17                The DPO suggests that any mechanistic 18 explanations    for    how      the      source    terms    occurs, 19 including the delayed activation of safety-related 20 systems, or the specification of main coolant system 21 pipe breaks, should be avoided.
22                The DPO recommends that the concept of the 23 maximum hypothetical accident be reinstated.                    I just 24 wanted to amplify that because I heard a comment that 25 we have added a new accident. Actually, something old NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
25 1 is      new    again.      The      term      "maximum    hypothetical 2 accident" in this Reg Guide is meant to describe how 3 we will analyze this accident.                        It is not a new 4 accident.
5                  The    maximum          hypothetical        accident, 6 historically, is that accident whose consequences, as 7 measured by the radiation exposure of the surrounding 8 public would not be exceeded by any other accident 9 whose occurrence during the lifetime of the facility 10 would appear to be credible.
11                  The MHA is described in several footnotes 12 in      regulation    as    "a    major      hypothetical    accident 13 resulting in substantial meltdown of the core with 14 subsequent release into the containment of appreciable 15 quantities of fission products."
16                  The DPO recommends that there should be a 17 clear distinction between the mechanistic design basis 18 loss of coolant accident used to show compliance with 19 Section 50.46 and the dose consequence analysis used 20 to demonstrate compliance with the regulatory dose 21 acceptance criteria.              This recommendation has been 22 incorporated into the proposed guidance by referring 23 to the regulatory dose consequence analysis described 24 in Appendix A of Draft Guide 1389 as the MHA LOCA.
25                  The adoption of DG-1389 of the concepts in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
26 1 this DPO resulted in the elimination of the Draft 2 Guide 1199 assumption that the initial source of BWR 3 MSIV leakage should be the reactor vessel's steam 4 dome.          Therefore,    the    source        term    multiplication 5 factors        endorsed    in      Draft      Guide      1199  to    ratio 6 containment atmospheric concentrations to simulate 7 reactor vessel steam dome concentrations have been 8 eliminated in the proposed guidance being discussed 9 today.
10                    Now, if there are no questions, I will 11 turn the presentation back to Mark to discuss our next 12 slide.
13                    MR. BLUMBERG:          Thank you, John.
14                    We'll    now      discuss        the    updates      we've 15 proposed in DG-1389.
16                    If you'll go to slide 14, please?
17                    Slide 14 provides a list of some proposed 18 updates to Reg Guide 1.183.                    These are changes that 19 were        originally      proposed          in      DG-1199.          Where 20 applicable, these changes were informed by public 21 comments previously provided on DG-1199.                        I'll now go 22 through some of these changes.
23                    Our proposed revision updates Reg Guide 24 1.183        to  add  new    guidance        and      remove  withdrawn 25 guidance.          For example, since Reg Guide 1.183 was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
27 1 issued, Reg Guide 1.194 for Meteorology was developed.
2 Since          meteorology      is      critical          to    the      dose 3 calculations, the Reg Guide 1.194 method was added in 4 DG-1389.
5                    As previously discussed, we will include 6 a source term which includes guidance for modern fuel 7 utilization.        We also incorporated changes due to two 8 regulatory issue summaries.                  For example RIS 2006.04 9 discusses 12 issues regarding experience the NRC has 10 had with the review of ASTs.                          These experiences 11 include those involving the MSIV leakage pathway and 12 decontamination          factors          for      the    fuel-handling 13 accident.          The proposed update to Reg Guide 1.183 14 includes revised methods to address some of these 15 issues.
16                    We    will      also        include      changes          in 17 terminology that have occurred since the issuance of 18 Reg Guide 1.183, Rev. 0. For example, the terminology 19 "EDEX," or Effective Dose Equivalent for External 20 Exposures, is now used, rather than "EDE," which was 21 known as the Effective Dose Equivalent.
22                    Next  is      the    removal        of  guidance        on 23 equipment qualification from Reg Guide 1.183.                              This 24 guidance is being transferred to a revision of Reg 25 Guide 1.189 which is for the qualification of Class 1E NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
28 1 equipment for nuclear power plants.                        This change is 2 contained in the proposed revision of Reg Guide 1.189, 3 which is known as DG-1361.                  Currently, the NRC staff 4 is addressing public comments on that draft guidance.
5                    Lastly, there is an update that's not 6 listed that I must mention.                        The NRC originally 7 created Reg Guide 1.183 for use by existing nuclear-8 powered        reactors      to    satisfy          regulations      under 9 10 CFR 50.34 and 50.67.              Reg Guide 1.183, Revision 1, 10 extends the applicability of the proposed guidance for 11 use by advanced and passive LWR designs in satisfying 12 the      requirements      for      dose      analysis      contained        in 13 10 CFR Parts 50 and 52 for safety and siting analysis.
14                    Would you please now go to slide 15?
15                    Reg  Guide      1.183,      Revision    0,  did      not 16 include an acceptable method for aerosol deposition in 17 the main steam lines. Some licensees proposed the use 18 of a calculation developed by the NRC staff to review 19 the      Perry    Alternative        Source        Term  Pilot    Project 20 submitted.          This document is known as AEB-98-03.
21 Other licensees used proprietary methods.
22                    The NRC staff saw some issues with some 23 calculations that used AEB-87-03.                        As a result of 24 these        issues,  the      NRC      staff        wrote  Regulatory 25 Information Summary 2006-04 to convey what the staff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
29 1 saw and needed to approve amendments that utilized 2 AEB-89-03.
3                In response to the issues we saw with 4 AEB-98-03, the NRC now has developed three deposition 5 models for the steam line.
6                The first is a holistic, state-of-the-art 7 model for modeling aerosol deposition in the steam 8 line and in the condenser.              That model is provided in 9 the Sandia National Laboratory's Document known as 10 SAND-2008-6601.        That model was also included in 11 DG-1199.
12                Two additional models were developed to 13 address the issues the staff saw with AEB-98-03.
14 These models are known as the Multi-Group Method and 15 the Numerical Integration Method.                    These models will 16 be discussed in more detail later in the presentation.
17                If you'll now go to slide 16, please?
18                In addition to the three new steam line 19 deposition        models,        the        staff      is    proposing 20 clarifications to Regulatory Positions for modeling 21 the      MSIV  leakage    pathway.            These    clarifications 22 include:
23                Modeling the steam lines as being intact.
24                Deleting        the        Regulatory        Positions 25 crediting the containment sprays when also crediting NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
30 1 aerosol removal in the steam lines.
2                    And thirdly, no longer crediting holdup in 3 aerosol deposition within the inboard steam line.
4                    Will you please go to slide 17 now?
5                    I  will      now    discuss          one  of  the    three 6 methods for the MSIV leakage deposition in the steam 7 lines.        That method is known as the SAND-2008-6601 8 Modified Method.
9                    Please go to slide 18 now.
10                    A    state-of-the-art,                  well-benchmarked 11 computer code called MELCOR was used to perform an 12 uncertainty analysis.                Within MELCOR, a code module 13 called MAEROS, which is an acronym for Multi-Component 14 Aerosol        Time  Evolution,          is    used      to  model        the 15 deposition of the aerosols throughout the MSIV leakage 16 pathway.        MAEROS models the growth in particle sizes 17 due to agglomeration or combining of aerosols and due 18 to      condensation.          It    also      models      other    removal 19 processes such as thermophoresis, diffusiophoresis, 20 and Brownian Motion.
21                    The MELCOR uncertainty analysis was very 22 similar to those currently endorsed in Reg Guide 23 1.183, Revision 0, for modeling containment aerosol 24 removal by natural processes and by sprays.                                  The 25 Modified SAND-2008-6601 Method does not use the steam NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
31 1 dome concentrations as the source of the MSIV leakage 2 for the first two hours of the accident.
3                  And now, I'll turn the presentation to 4 Elijah Dickson, who will discuss the two other methods 5 for modeling MSIV leakage deposition.
6                  MR. DICKSON:        Thank you very much, Mark.
7                  Please, the next slide. Slide 20, please.
8                  This      slide          highlights          our        work 9 incorporating lessons learned over the last 21 years 10 reviewing      alternative        source        term    boiling      water 11 reactor main steam isolation valve leakage pathways.
12 This effort, in updating 1.183, we decided to revisit 13 the      AEB-98-03  analyses,        which      Mark    has  mentioned 14 supported one of the first approved ASTs for the Perry 15 Pilot Plant.
16                  The work in regards to reevaluating the 17 AED      models  and  methods        and      the    Multi-Group        and 18 Numerical Integration Methods are presented in detail 19 in the first referenced memo at the bottom of this 20 slide.
21                  As Mark discussed, Rev. 0 does not contain 22 an aerosol deposition model for the main steam lines.
23 That is a model that you could use to compute removal 24 coefficients for the dose calculations. And what many 25 licensees have done is to follow the initial analyses NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
32 1 of AEB-98-03, in conjunction with the 2006 RIS.
2                    To address issues that were addressed in 3 the 2006 RIS, what many licensees would do is apply 4 more conservative settling velocity than what was done 5 for the Perry Pilot Plant.                  And some licensees also 6 developed what's termed as the 20 Group Method, which 7 is a Monte Carlo method to integrate the sudden 8 velocity distribution in that particular steam line.
9                    So, to date, we've performed a cursory 10 review of BWR SOAR.                Around 25 out of 30 of these 11 licensees have adopted some form of AED-98-03, and 12 four licensees have adopted or utilized this 20 Group 13 Method.
14                    In  our    reevaluation            of  AED-98-03,        we 15 reached out to the Office of Research, looking for 16 recommendations in how to correct the issues brought 17 up in the 2006 RIS, as well as to assess this 20 Group 18 Method. They performed a literature review for us and 19 updated the technical bases for AEB-98-03, applying 20 updated physics parameters.
21                    We settled on recommendations that are 22 discussed or provided in the European State-of-the Art 23 Report on Nuclear Aerosols, or the SOAR report, which 24 evaluated results from the PHEBUS test experiments, 25 giving        recommendations        for      our    aerodynamic      size NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
33 1 distributions ranging from 1 to 4 microns with a log-2 normal distribution.
3                  In regards to assessing how to integrate 4 the      settling    velocity        distribution        to  compute        an 5 effective removal constant that can be used in the 6 dose analyses, we assessed two integration methods, 7 one      being  the  20    Group      Method        as a  Monte      Carlo 8 integration, and the other being the Numerical Method.
9                  When assessing the industry-produced 20 10 Group Method, we were unsure if 20 groups was enough 11 to converge the results, and later found that you need 12 about 2,000 groups.                You need to discretize the 13 dataset into about 2,000 groups to properly integrate 14 the settling velocity distribution in the main steam 15 line. And so, we're now referring to it as the Multi-16 Group Method.
17                  To  verify        the    Multi-Group        Method,        we 18 developed the second method, the Numerical Integration 19 Method,        which    is      integration            by  parts,      which 20 effectively normalizes the distribution, and then, 21 allows you to compute an effective removal constant.
22                  There's many benefits to these updated 23 analyses.        We expect that the analyses themselves 24 should be simplified, and we should see consistency 25 between licensees in submitting license amendment NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
34 1 requests.
2                  We      are          using            state-of-the-art 3 recommendations from the SOAR report.                      And the models 4 themselves are really rather quite flexible.                              They 5 utilize plant-specific information and could be used 6 for        small  modular        reactor,          BWR-type      designs, 7 notwithstanding the source term distributions intended 8 for light water reactors.                  The SOAR report really 9 focused on large light water reactors, and we're 10 uncertain if the distribution of aerosols being 1 to 11 4 microns, as discussed in the SOAR report, would be 12 appropriate for BWRs of small modular design.
13                  And in the memo referenced here at the 14 bottom of this page, we also have some example cases 15 in      which    an  individual          can      go    and  rerun        the 16 calculations using the Multi-Group or this Numerical 17 Integration Method.
18                  And I'd like to note that the referenced 19 values used for the example calculations themselves 20 are not specifically being established as Regulatory 21 Positions. They are merely example calculations. For 22 the      purpose  of  this      Reg    Guide,        Rev. 0,    we've 23 established that an appropriate aerodynamic diameter 24 particle distribution being 2 microns.
25                  And    with      that,        I      will    turn        the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
35 1 presentation over to --
2                CHAIR PETTI:        So, I have a question, Ed, 3 before you go on.
4                MR. DICKSON:        Yes?
5                CHAIR PETTI:        So, the Multi-Group Method 6 or the integration by paths --
7                MR. DICKSON:        Yes.
8                CHAIR    PETTI:          --    gives    you  the      most 9 detailed result.      But you seem to imply at the top of 10 this slide that there's a 20 Group Method that's still 11 acceptable,      even    though        it      may    not    converge 12 adequately?
13                MR. DICKSON:        We've never endorsed the 20 14 Group Method.      We've reviewed them over the years.
15 And in updating this Reg Guide, we reviewed the 20 16 Group Method, and the, I would say that we made 17 modifications and enhancements to it to make the 18 Multi-Group    Method      an    acceptable          model    for      the 19 Regulatory Guide.
20                CHAIR PETTI:          But there have been LARs 21 that have used the 20 Group Method.
22                MR. DICKSON:        There have been, yes.
23                CHAIR PETTI:        Okay.        Thank you.
24                MR. BLUMBERG:        Just I'd like to point out 25 that, within those LARs, typically, the staff required NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
36 1 additional conservatisms to try to offset the issues 2 we saw with the 20 Group Method.
3                    CHAIR PETTI:        That was my next question.
4 Great.        Thanks.
5                    MR. BLUMBERG:          Yes.
6                    MR. DICKSON:          Yes, thank you, Mark.                I 7 appreciate that.
8                    And    with      that,        I      will  turn        the 9 presentation over to Steve Jones.
10                    MR. JONES:        Yes.        Good morning.        I am 11 Steve Jones from the Containment and Plant Systems 12 Branch of NRR.
13                    I'll be describing proposed changes to 14 guidance for use of the power conversion system to 15 reduce dose consequences from BWR main steam isolation 16 valve leakage following a postulated accident.
17                    Next slide, please.
18                    The approach establishes a low resistance 19 flow path to the main condenser from the main steam 20 lines just downstream of the MSIVs. The condenser has 21 sufficient volume to delay the leakage flow by the 22 main steam isolation valve, or MSIV, seats to provide 23 for holdup of radionuclides.                  In addition, the models 24 provide for radionuclide deposition within the main 25 steam lines, as Elijah and Mark discussed previously, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
37 1 downstream of the MSIVs, and a specified release rate 2 from the main condenser, once the leakage reaches that 3 point.
4                This pathway was originally defined in an 5 industry Topical Report and has been included in the 6 existing Rev. 0 version of Reg Guide 1.183 to provide 7 a      quantitative  reduction          in    dose  consequences.
8 Approximately half of the operating BWR fleet has 9 adopted that BWR Owners' Group methodology.
10                Next slide, please.
11                MEMBER BROWN: Can I ask a question before 12 you go on?
13                MR. JONES:        Oh, sure.
14                MEMBER    BROWN:          Go    backwards  with      the 15 slides.
16                Has that always been there as a passive 17 path, that you're just now allowing it --
18                MR. JONES:        Originally --
19                MEMBER BROWN:          Was it used, is what I'm 20 trying to say, before to evaluate the accidents?
21                MR. JONES:        Prior to the early 1990s, it 22 was not used.      The BWRs were originally designed with 23 leakage control systems, which involved active valves 24 to redirect the downstream portion of the main steam 25 lines, the portion downstream of the MSIVs, to a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
38 1 suction, to a fan that pushes that release onto the 2 standby gas treatment system and out the stack.
3                    However, those systems were found to be 4 problematic in their ability to both handle postulated 5 leakage amounts, and, you know, you can see, going 6 from a 1,000-pound system to a very low pressure 7 system like the standby gas treatment system, there 8 can be issues with the controls and having that system 9 function correctly.              So, most plants have gone to a 10 more      passive    reliance      on    this      use  of  the    power 11 conversion system itself to hold up the radionuclide 12 leakage.
13                    And as I indicated, I guess the staff 14 accepted the BWR Owners' group report in 1999 with a 15 Safety        Evaluation    Report.            I'll    be  getting        to 16 discussion of the conditions and limitations of that 17 in one of the future slides.
18                    MEMBER BROWN:          Okay.
19                    DR. SCHULTZ:            Steve,      this    is    Steve 20 Schultz.
21                    The earlier slide on clarifications for 22 the leakage pathway, slide 16 says, "No credit for 23 holdup or deposition within the inboard main steam 24 line."          Is that a new condition or does the BWR 25 Owners' Group methodology already not take credit for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
39 1 that?
2                    MR. JONES:        Well, the BWR Owners' Group 3 methodology just deals with the leakage that has 4 gotten past the MSIVs.                A deposition inboard of the 5 main      steam    isolation        valves        is    actually    inside 6 containment, and I would have to refer to Mark or 7 Elijah to address it, if you have any more questions 8 about that.
9                    MR. BLUMBERG: Sure. I can try to address 10 that.
11                    As far as enrollments to the same DM 12 method, when we did the Sandia 2008-6601 Method, we 13 found that there were expectations that there would be 14 turbulence in that section of the piping due to the 15 mixing        that's  credited        up      to    the  MSIV    leakage 16 isolation valves.          And so, the methods that, at least 17 with respect to utilization of the Stokes equation 18 that's used in AEB-98-03, in that volume, we didn't 19 see that it would be appropriate to credit deposition 20 in that volume using that Stokes equation because you 21 had forces other than buoyancy and gravity that would 22 be present.        The turbulence would potentially change 23 the      conditions    by    which      that      Stokes    equation        is 24 applicable.
25                    And then, secondly, that volume up to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
40 1 MSIVs is also part of the source term volume for 2 containment.        From a regulatory perspective, up to 3 those MSIVs are considered part of containment, an 4 extension of containment, because of the break that 5 occurs in the accident.              So, modeling it separately, 6 we didn't see it would be appropriate.
7                  So, for those two reasons, this is a 8 potential change for some licensees, if they were to 9 adopt this method in Rev. 1.
10                  DR. SCHULTZ:        So, that would have been a 11 change, if you will, between the initial application 12 and the results from the Sandia 2008 study?
13                  MR. BLUMBERG: It's really site-specific.
14 You would have to go back and look at the individual 15 licensees      because    there      were        differences    between 16 applications.        So, I can't really answer that for 17 everyone, but I do know that up to those MSIVs are 18 part      of  containment,        and    that      source term      is    a 19 releasant to containment. So, crediting deposition in 20 that volume, in addition to other deposition methods 21 played out or sprays, might not be appropriate.                          So, 22 that's one of those --
23                  DR. SCHULTZ:        This may be --
24                  MR. BLUMBERG:          Go ahead.
25                  DR. SCHULTZ:          This may be a change for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
41 1 some licensees?
2                    MR. BLUMBERG:          Yes, that's correct.
3                    DR. SCHULTZ:        Thank you.
4                    MR. JONES:        Okay.      Slide 23, please.
5                    Okay. As we discussed, Reg Guide 1.183, 6 Rev.      0,  actually    endorses        the      BWR  Owners'      Group 7 methodology.            And      a    part        of  that    includes 8 identification of the pathway, that is, what valves 9 would        need  to  be    operated        to    establish    a    low 10 resistance        flow  path      to    the    main    condenser      from 11 downstream of the MSIVs.
12                    And this change in Rev. 1 does not affect 13 the identification and establishment of the pathway 14 and its ability to direct that leakage flow to the 15 main      condenser.        However,          the    revised  guidance 16 includes streamlined justification of seismic capacity 17 as an alternative to the methods defined in the Safety 18 Evaluation Report in the Limitations and Conditions 19 section for the BWR Owners' Group Topical Report.
20                    And this slide attempts to identify two 21 flow paths, really, with respect to not physical flow 22 paths,        but  decision        points      with    regard    to      the 23 approach.        Information on the seismic capacity could 24 still use the Reg Guide 1.183, Rev. 0, method, or 25 leveraging risk information, the Staff's Technical NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
42 1 Assessment.
2                    Next slide, please.
3                    MEMBER    BROWN:          What      do  you  mean        by 4 "seismic capacity," for the uninitiated?
5                    MR. JONES:        Okay.
6                    MEMBER BROWN:          Does that mean this other 7 alternate path provides --
8                    MR. JONES:        Right.        For the majority of 9 BWRs, the main steam system is not safety-related 10 downstream of the MSIVs beyond the outboard seismic 11 restraints.          It's typically augmented quality.                        So, 12 better than typical power piping, but, in other words, 13 it's been subject to generally volumetric NDE to 14 verify        that  the    welds      are      correct      and  that      the 15 supports and things have been installed correctly.
16 And, in addition, it's typically designed to ASME 17 B31.1 power piping code.
18                    In some cases, some BWRs have upgraded 19 piping designs that use ASME boiler and pressure 20 vessel code for that piping.                    And the latest designs 21 often        include    stop      valves      that      are  intermediate 22 between the MSIVs and the turbine stop valves.
23                    MEMBER BROWN:          Thank you.
24                    MR. JONES:        Okay.      Next slide, please.
25                    Okay. The BWR Owners' Group report relied NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
43 1 predominantly on post-seismic event damage assessments 2 at non-nuclear sites to assess the seismic fragility 3 of piping and piping systems and develop a generic 4 seismic hazard curve applicable to those systems.
5                  The      associated            NRC      staff      Safety 6 Evaluation,        as    I      mentioned,          included      several 7 limitations and conditions on the use of that generic 8 information that may be unnecessarily burdensome to 9 provide the reasonable assurance of the pathway's 10 seismic        capability.            These        conditions      included 11 relatively detailed walkdowns to confirm configuration 12 and information on the piping supports.
13                  Next slide, please.
14                  The NRC staff has developed a Technical 15 Assessment        of  seismic        capability          that  considered 16 engineering        insights        and      updated      probabilistic 17 assessment of reliability following seismic events and 18 operating experience in sites from nuclear power plant 19 walkdowns following seismic events.
20                  The    staff        has      identified      a    median 21 fragility        for  components          in    the    pathway    to      the 22 condenser of at least 0.4 g.                    That value exceeds the 23 peak ground acceleration values for the safe shutdown 24 earthquakes at most U.S. BWR sites.
25                  The    staff        intends          to    release      that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
44 1 Technical Assessment as part of the Reg Guide package 2 and solicit public comments regarding that assessment.
3                  The use of this Regulatory Position to 4 establish seismic capacity would involve provisions of 5 the results of more limited assessments that could be 6 verified through an audit or inspection, rather than 7 providing      complete      evaluations            as  part    of      the 8 application submittal.
9                  MEMBER    HALNON:            Steve,    this  is      Greg 10 Halnon.
11                  MR. JONES:        Yes?
12                  MEMBER HALNON: I guess the seismic aspect 13 where, if the piping's not there, it can't retain it.
14 But how do you account for all the myriad of potential 15 failures a seismic event could cause, such as stream 16 trap      problems  that      will      bypass        directly  to      the 17 condenser and bypass a lot of the piping that we're 18 probably trying to take credit for to have deposition?
19 That's just an example, but there's probably many 20 others, in addition to the myriad of different designs 21 on the secondary side, based on who constructed it.
22 How do you account for all that?
23                  MR. JONES:        Well, that is part of the 24 identification of the pathway.                      And actually, the 25 predominant or the normal pathway selected, and that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
45 1 is identified actually, in the BWR Owners' Group 2 Topical, is use of the drain pathway from the main 3 steam lines.        And usually, there's a bypass valve 4 around the drain, the moisture traps, to direct that 5 leakage to the main condenser.
6                MEMBER HALNON: Okay. So, the piping that 7 we're crediting is the earliest, quickest path to the 8 condenser relative to the steam flow. Is that because 9 many of the preheaters are bypassed and other things 10 are happening on a turbine trip that prevent water 11 induction into the turbine?              So, much of that is shut 12 down.
13                MR. JONES:        Right.
14                MEMBER HALNON:            So, are you telling me 15 that it's a minimalist-type approach?
16                MR. JONES:          Right,        it's  a  minimalist 17 approach,      just    attempting            to    establish    a      low 18 resistance flow path to the main condenser.                    However, 19 we still expect that there would be some holdup within 20 the main steam lines themselves.                  It's a very limited 21 driving head.      We're talking on the order of a couple 22 hundred cubic feet per hour of leakage flow.                            And 23 there would be some cooling and possibly condensing of 24 that leakage within the main steam lines after some 25 period of time.
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46 1                Again, I mean, looking at the release 2 assumptions,    there    would      be    some    time  delay        to 3 establish this leakage path and provide that flow path 4 directly to the main condenser.
5                MEMBER HALNON:          Okay.      The next question:
6 how do we account for maintenance issues on the 7 condenser itself with high-oct gases and potential 8 small tube leaks, and stuff? Is that assumed to be at 9 the maximum tech spec allowables, and all that stuff?
10                MR. JONES:        The Reg Guide 1.183 provides 11 a defined leakage rate from the main condenser.                      It's 12 not intended to be leak-tight following the event.
13                MEMBER HALNON:          Okay.
14                MR. BLUMBERG:          This is Mark Blumberg of 15 the staff.
16                In general, the way that is modeled for 17 this particular accident is what goes in comes out.
18                MEMBER HALNON:          Okay.      So, eventually, it 19 does get out?
20                MR. BLUMBERG:          Right.
21                MEMBER HALNON:          Okay.      I got it.
22                MR. JONES:        Okay.      Thank you, Mark.
23                DR. SCHULTZ:            Steve,      this  is    Steve 24 Schultz.
25                From your previous slide, note regarding NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
47 1 the original requirement for the investigation of the 2 components with outside contractors, and so forth.                              I 3 presume that, on this slide, the docketed information 4 that licensees will need to provide does not include 5 that,        based  upon      what      was      presented      to      the 6 Subcommittee last July and the overall evaluation of 7 experience bases on seismic.                    Is that true?
8                    MR. JONES:        Right, that's correct.                I'll 9 be getting to that in a little more detail in the next 10 slide.
11                    DR. SCHULTZ:        Oh, thank you.
12                    MR. JONES:        Thank you.          Thanks.
13                    MR. HSUEH:        This is Kevin Hsueh, NRR.
14                    So, I guess this slide, I saw there was an 15 earlier question about using the NuScale block valve 16 SRM. And so, this is the Technical Assessment kind of 17 used        in    terms    of      applying          the    risk-informed 18 principles.          And so, this is how we are kind of 19 following up with the design, how we apply risk-20 informed principles.              So, this is one example.
21                    MR. JONES:        Okay.      All right.      Thank you, 22 Kevin.
23                    Getting back to this slide, the regulatory 24 guidance establishes two paths for confirmation of 25 seismic capacity: a deterministic dynamic analysis of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
48 1 seismic response to the safe shutdown earthquake, 2 using the code of record, which would typically be, as 3 I mentioned, ASME B31.1.
4                This is not necessarily a safety-related 5 assessment, but a dynamic assessment that the licensee 6 references and makes available, either through audit 7 or inspection, to verify the integrity of the piping.
8                Alternatively, the licensee could follow 9 a more risk-informed evaluation method with gradation 10 of the elements of that method, based on the site 11 ground motion response spectra.
12                For    the      risk-informed            method,        the 13 evaluation      would  confirm        margin        in the    existing 14 evaluation to the code of record.                  That would be more 15 of a static evaluation. Considering insights from the 16 individual plant examination for external events; 17 also, verification of flexible joints at major branch 18 connections and terminal ends to verify that there's 19 no      likely  locations        of    rupture        that  would        be 20 inconsistent with the staff's probabilistic assessment 21 of      seismic  fragility.          And      variable    degrees        of 22 sampling of the pathway during walkdowns to establish 23 construction      consistent        with      design    and    adequate 24 supports.
25                When the site seismic hazard exceeds the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
49 1 median fragility, the guidance would call for more 2 extensive        walkdown        sampling          and    confirmatory 3 calculations        to  be      completed          to    verify    seismic 4 integrity commensurate with the increased seismic 5 hazard and considering the extent of quality assurance 6 applied to the piping design and fabrication during 7 initial construction. The intent is to provide a more 8 streamlined and risk-informed assessment consistent 9 with the safety importance of the holdup function.
10                  This approach differs significantly from 11 the Draft Interim Staff Guidance previously presented 12 to the Subcommittee last fall.                          The ISG did not 13 provide        methods  to    change      the      radiological        dose 14 consequence analysis of record.                        However, for the 15 change to Reg Guide 1.183 in Revision 1, the staff is 16 seeking confidence in the reliability of the pathway 17 to the main condenser commensurate with the dose 18 reduction credibly provided by these components.
19                  If there are no further questions, I will 20 now turn the presentation over to John Parillo to 21 address MSIV leakage rates.
22                  CHAIR PETTI:        Yes, so hold on.
23                  MR. JONES:        Okay.
24                  CHAIR PETTI:        I just had a question --
25                  MR. JONES:        Sure.
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50 1                CHAIR PETTI:        -- on the ISG.        This is the 2 ISG that we, as a full Committee, reviewed here a few 3 months ago.
4                MR. JONES:        Yes.
5                CHAIR  PETTI:          And    now,  you're    saying 6 there's a change to that.
7                MR. JONES:        No, there's no -- I'm just 8 differentiating what we're doing with Reg Guide 1.183 9 compared to the ISG.        The ISG is more of an aid to the 10 reviewer in assessing what level of assurance they 11 need for some of the assumptions going into the dose 12 consequence analysis, rather than any change to the 13 actual analysis of record and the assumptions that go 14 into that.
15                However,      the      Reg      Guide  1.183,        this 16 revision is looking for detailed information about the 17 design and pathway established.                  And as a result of 18 having that information, actually modified the dose 19 consequence analysis to take credit for that pathway 20 as a means of reducing the release to offsite or 21 control room personnel.
22                CHAIR PETTI:          And I'm not sure I fully 23 understood. As I recall, when we talked about the 24 ISG, I thought we were told that that was going 25 completely into Reg Guide 1.183 as is.                      But now, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
51 1 you've said that there's additional information and an 2 additional option?
3                  MR. JONES:        No, the --
4                  CHAIR PETTI:            Is that the way to think 5 about it?
6                  MR. JONES:          The ISG is not intended to 7 enter the Reg Guide at all.                    It would be eventually 8 adopted into the Standard Review Plan, because it's 9 really, essentially, staff guidance.                      But it does not 10 affect the method that the licensees use to perform 11 their analysis of record for the dose consequences, 12 which        is  the    purpose          of      the      --  you      know, 13 alternatively, which is the purpose of the Reg Guide, 14 is to provide guidance to the licensees when they're 15 developing their analysis.
16                  CHAIR      PETTI:                But      they're        not 17 inconsistent, I guess is a better way to say it?
18                  MR. JONES:          They're different, I guess.
19 The ISG is intended to indicate that we have some 20 confidence that, even if a pathway is not specifically 21 identified, the power conversion system is still going 22 to be there.        It doesn't vaporize.                And therefore, it 23 would        provide  some        level      of    holdup,    but      it's 24 unquantified and it doesn't -- it just is intended, as 25 I    mentioned,      for    staff      use      in    assessing      their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
52 1 information needs regarding assumptions and other 2 inputs to the analysis of record.
3                  CHAIR PETTI:            I guess, you know, I'm 4 probably obtuse here.            I'm just confused.            We will be 5 writing a letter on an ISG.                I don't want to write a 6 letter on an ISG that is somehow inconsistent with 7 what the underlying philosophy and technical basis is 8 on 1.183.        That's what I'm trying to get --
9                  MR. JONES:        I guess it's the commonality 10 is, for both, we're relying on the probabilistic 11 seismic        assessment,      and    that      we    have  a  lot      of 12 confidence that the power conversion system, including 13 the steam lines, the drain piping, other equipment, 14 will actually provide holdup volume for actual changes 15 to the analysis of record that, in fact, credit that 16 holdup in the main condenser -- in other words, the 17 delay of the release, allowing more fission products 18 to decay.        If that credit is actually taken, then the 19 licensee needs to provide sufficient information, 20 based on the seismic hazard, to justify that they will 21 establish a flow path, and that that would be a 22 reliable means of directing leakage flow to the main 23 condenser.
24                  From  the      ISG    perspective,        what      we're 25 telling the staff is that, even though the power NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
53 1 conversion    system    is    not    safety-related          and      not 2 seismic, there is substantial confidence that there 3 will be surfaces available for deposition and some 4 lower confidence pathways that eventually get to the 5 main condenser, where other volumes that may allow for 6 holdup of the -- that will most likely allow for 7 holdup of a release.
8                And  therefore,          the      quality    or    --    I'm 9 sorry, I'm having trouble finding the right word.
10 But, you know, their confidence in the accuracy of the 11 assumptions need not be quite as great, since the 12 assumption in the actual analysis of record is that 13 there is a direct release at the MSIV.
14                So,  does      that    --    is    that  helpful        in 15 clarifying the distinction between the ISG and the Reg 16 Guide?
17                CHAIR    PETTI:          Yes,      I    understand        the 18 purpose of the two documents.                I just want to assure 19 myself that what you're telling applicants in the Reg 20 Guide and what you're telling the staff to look for in 21 evaluation are not inconsistent.                      That's all I'm 22 asking.
23                MR. JONES:        They are not inconsistent, 24 but, for the Reg Guide, there's a lot of work that the 25 applicant or a licensee would need to do to establish NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
54 1 that this is a viable path and that it should be 2 credited.        Our intent here is just to relax a little 3 bit from the degree of effort that's required to 4 establish that path compared to the BWR Owners' Group 5 Topical Report and our associated Safety Evaluation.
6                    CHAIR PETTI:          Okay.        I see a hand up.
7 Mike?
8                    MR. FRANOVICH:          This is Mike Franovich.
9                    I think you got the gist of what the 10 purposes of the ISG vis the Reg Guide, and it really 11 comes        down  to,  essentially,            if    the  licensee        is 12 crediting these pathways in their analysis of record 13 to stay within the limits of the regulations, they're 14 going to have to do a little bit more demonstration to 15 provide        us  assurance        that    that's      appropriate        for 16 crediting.
17                    The ISG is more focused on, well, the 18 licensees -- and I'm generalizing a little bit here --
19 if they are not going to rely on these pathways, they 20 are going to do an analysis without the pathways, 21 essentially, and we're looking at where they are in 22 terms of dose estimates.                As Steve said, physically, 23 the condenser and piping will be there. How much more 24 information or how should we be looking at that, when 25 we're making an alternate decision that the analysis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
55 1 that they're providing, which doesn't rely on those 2 pathways, is within the 5-rem limit of the regulation?
3                    So, that's the kind of gist of it.                      So, 4 there's more work the licensee has to do.                                They 5 wouldn't really be using the ISG for an analysis of 6 record.        They're going to be using the Reg Guide.
7                    But I understand your point about, are 8 these two documents, from a technical standpoint, are 9 they      compatible?          I    would      say,    yes,  they      are 10 compatible.          There's just some more rigor that's 11 needed if a licensee's own analysis of record is going 12 to depend on the pathways.
13                    CHAIR PETTI:          Okay.        So, let me ask a 14 different        question.          Does    the      Reg  Guide  require 15 greater rigor than the LARs that have been accepted 16 that have credited the condenser to date?
17                    MR. JONES:        No, the intent is to reduce 18 the      level    of  rigor      with    respect        to  the  seismic 19 analysis.
20                    CHAIR PETTI:        Right.        Okay.
21                    MR. JONES:          There are still walkdowns 22 involved, but they're more limited and focused on 23 where the high risks are.                And one of those is, as I 24 pointed out, verification of flexible joints at the 25 major branch connections, as any constrained joints NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
56 1 might be more subject to cracking during a seismic 2 event to an extent, that could cause bypass of what 3 we're really trying to accomplish here.
4                  And then, of course, actually looking at 5 the flow, walking down parts of at least the flow path 6 to      the    main  condenser        to      establish  that      it's 7 adequately supported.
8                  CHAIR PETTI:        Thank you.
9                  MEMBER BROWN:        Yes, this is Charlie Brown 10 again.
11                  And the basis for allowing less rigor is?
12                  MR. JONES:        It's predominantly based on 13 our assessment of the seismic fragility of the median 14 pipe systems; our operating experience from a number 15 of seismic events that have directly impacted nuclear 16 power plants.        Between North Anna, the KK -
17                  MEMBER BROWN:          Let me --
18                  MR. JONES:        I'm sorry.
19                  MEMBER BROWN:          Let me phrase that.          So, 20 you're fundamentally basing it on actual end results 21 of seismic events where the systems have maintained 22 their integrity?
23                  MR. JONES:        Right.
24                  MEMBER BROWN:          Did I say that right?
25                  MR. JONES:        That's correct.      That's part NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309      www.nealrgross.com
 
57 1 of it, but another part of it is, as I mentioned, the 2 code of record used to design the piping and augmented 3 QA      applied  to  the      fabrication          and  construction.
4 There's a lot of elements that address that.                      But, in 5 total, we're predominantly looking at a passive system 6 that has demonstrated a high degree of reliability 7 over the life of these units.
8                  MEMBER BROWN:            But you're expanding it 9 beyond the exact plants that actually experienced the 10 seismic events, based on an assessment that their 11 design bases are similar, and therefore, it should be 12 acceptable?
13                  MR. JONES:        Right.        That's correct.
14                  MEMBER HALNON:          So, just carrying on with 15 what Charlie -- something struck me.                    This is Greg.
16                  Looking at a post-trip situation with a 17 lot of steam flow going to the condenser, there have 18 been some events where flow-accelerated corrosion has 19 reared its ugly head and has had significant leaks.
20 How do things like that, where there's potential 21 degradation        on  the      secondary          side  that's        not 22 necessarily picked up -- is that in any way looked at 23 it?        Or is that part of the walkdown process?                        In 24 programmatic, I guess, from the standpoint of you're 25 looking at their programs, making sure they're up to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
58 1 snuff.
2                  MR. JONES:        Well,        I    guess,  from      the 3 standpoint        of  flow-accelerated              erosion,    we    don't 4 expect that in these drain lines that flow directly to 5 the main condenser.
6                  MEMBER HALNON:          But, again, that's where 7 they have manifested themselves in time. I mean, some 8 of them have done it on the reheat system, but there 9 have been -- I believe the FAC programs cover those 10 volumes, don't they?
11                  MR. JONES:            Yes,        I  believe        the 12 erosion/corrosion program does apply, but I'm not 13 actually familiar with any that have actually occurred 14 in these small -- we're talking very small-diameter 15 drain lines, like 1-inch or so.
16                  MEMBER HALNON:            Yes, I agree.            I think 17 it's larger lines that the erosion/corrosion volumes 18 look at, but they do look at areas.
19                  MR. JONES:        Right.
20                  MEMBER HALNON: I was thinking more of the 21 margin.        I mean, we talk about walkdowns.                  You can be 22 walking down a secondary system and everything looks 23 good.        But if there's internal erosion or corrosion, 24 that you're not going to see it, and a seismic event 25 would normally not affect the high-integrity pipe, it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
59 1 may cause a leak.          I'm just wondering, is that even 2 discussed in any of the review?
3                  MR. JONES: No, we have not addressed that 4 specifically. We are looking for margin in the piping 5 design.        These pipelines are, you know, generally, as 6 I mentioned, small diameter.
7                  My    familiarity            with      the    operating 8 experience is that erosion/corrosion is generally are 9 in larger-diameter pipes that experience two-phase 10 flow.        We don't expect any flow, you know, any phase 11 change to be occurring in this piping.                      And, in fact, 12 normally,        they're    isolated        on    a    number  of    these 13 plants.
14                  MEMBER HALNON:            Well, there will be a 15 phase change somewhere, whether it's at the very --
16                  MR. JONES: Well, I mean, we're looking at 17 establishing this an hour or so after the event.                            So, 18 it's not an --
19                  MEMBER HALNON:          Much lower pressure.
20                  MR. JONES:        Right.
21                  MEMBER HALNON:          Okay.
22                  MR. JONES:      And it should be closed.                  The 23 piping should have experienced an hour or so of 24 cooldown, and those would be the conditions.                                Now 25 we're more cooled-down and we'll be looking at that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
60 1 particular plant condition.
2                  MEMBER HALNON: Okay. Well, weakened pipe 3 would occur when it's online, and then, manifest 4 itself in a seismic event later on.                    But, yes, I get 5 the lower pressure.          And for some reason, the 3-inch 6 line hits my mind as relative to the erosion/corrosion 7 program.      So, we may be talking different lines, but 8 I'll look it up, and if I have another question, I'll 9 get back to you.
10                  MR. JONES:        Okay.      Thank you.
11                  MEMBER BROWN:          It's Charlie Brown again.
12                  Is the basis for this, the basis for the 13 less rigor, is that documented in either the Reg Guide 14 or the ISG, or some other supporting document?
15                  MR. JONES:        Yes.      There's a --
16                  MEMBER        BROWN:                There's      written 17 documentation?
18                  MR. JONES:              There's        a    Technical 19 Assessment. There is a Technical Assessment that will 20 be issued for public comment in conjunction with the 21 regulatory guide.
22                  MEMBER BROWN:            Which explains why less 23 vigor is satisfactory?
24                  MR. JONES:        Yes.
25                  MEMBER BROWN:          Okay.
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61 1                  MR. JONES:        And it predominantly relates 2 to those three factors, or three major factors really:
3 the margin and the code of record applied to the main 4 steam system; the seismic fragility evaluation that's 5 been done, and operating experience from other plants.
6 I'm sorry, I should have said four. And the walkdowns 7 that help establish the integrity of the system.
8                  MEMBER BROWN:          Okay.        Thank you.
9                  MR. JONES:        Thanks.
10                  MEMBER    BIER:          I    have    some  followup 11 questions or comments.              Vicki here.
12                  First, I think part of the confusion or 13 angst about this is just the choice of words; that if 14 you say, "less burden," everybody might say, okay, 15 fine.          If you say, "less rigor," everybody gets 16 worried about, is this going to be less accurate as 17 well?        So, that's just kind of a suggestion.
18                  But the other thing that I would worry 19 about, again, without having seen all the analyses, 20 and whatever, is evidence from a small number of 21 earthquakes from a Bayesian perspective may be very 22 weak evidence about whether you could expect those 23 successes to be replicated in future earthquakes.
24                  It's one thing if you go in with an 25 assumption that an earthquake is guaranteed to cause NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
62 1 problem, and then, you see a couple of events where 2 nothing terrible happened.            You could say, okay, it's 3 obviously not guaranteed; the probability is a lot 4 less.
5                But if you go with in a perception that, 6 oh, maybe half the earthquakes are going to have big 7 problems, and then, you see two or three where it 8 didn't happen, it's not clear that, just from a 9 statistical perspective, you learn much.
10                Again, you know, there may be a lot of 11 other analysis that supports that, but, you know, I 12 mean, I have not seen it and may not be qualified to 13 comment on it.        But, statistically, two or three 14 successes seems like pretty weak evidence to me.
15                MR. JONES:        Okay.          Dr. Vasavada      can 16 address your questions. I believe he has his hand up.
17                DR. VASAVADA:          Yes, thanks, Steve.
18                This  is    Shilp      Vasavada.        I'm  at      the 19 divisional risk assessment in NRR.
20                To address that question, I'll say that 21 the operating experience and the walkdowns for the 22 three plants that Steve mentioned, or especially two, 23 North Anna and Kashiwazaki Kariwa, they were one of 24 the important, not the only places, for whatever is 25 being proposed in the new provision.
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63 1                  We  also        had      a,      and    the    Technical 2 Assessment is that the fragility evaluation that we 3 looked at, actually, fragility data for bolted pipes 4 and other confidence, which will be part of this --
5 but      from  the  1990s      all    the      way    to  the    current 6 contemporary seismic PRAs that were developed post-7 Fukushima, that is also a central piece of this 8 particular      evaluation.          So,      it's    not  just    about 9 dumped; it's also a seismic fragility evaluation.
10                  And based on that, what was determined was 11 a lower bound in comparison to multiple failure modes.
12 It's even less, if you can think about it, than the 13 offsite power facility, which is expected to fail, 14 offsite power that's expected to fail for seismic.
15                  So,    it's      not      just        one  piece,        the 16 walkdowns.        It supports the fragility assessment, 17 which        is the  central        piece,        and    the  Technical 18 Assessment carries all of that information.
19                  Thank you.
20                  MEMBER BIER:          Thank you.
21                  MR. JONES: Okay. If there are no further 22 questions, I'll pass it on to John Parillo to address 23 the MSIV leakage rates.
24                  MR. PARILLO:          Thank you, Steve.
25                  As discussed, the Draft Guide contains NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
64 1 three BWR MSIV leakage deposition models, as well as 2 a streamlined procedure to allow credit for an intact 3 power conversion system proving a pathway to the 4 condenser.
5                  The staff's intention is that the revised 6 guidance will provide licensees with a methodology 7 that will have a more predictable regulatory outcome.
8 Currently, the two licensees with the highest total 9 MSIV leakage values in the BWR fleet, 400 and 350 10 standard cubic feet per hour, credit a pathway to the 11 condenser        in  their      dose      consequence      analysis        of 12 record.        However, there are several licensees with 13 total MSIV leakage values exceeding 200 standard cubic 14 feet per hour that do not credit a pathway to the 15 condenser.
16                  The NRC staff is concerned that providing 17 a streamlined procedure to allow credit for a pathway 18 to      the    condenser    may    result        in  licensees      with 19 relatively high MSIV leakage values submitting license 20 amendments requesting to credit a pathway to the 21 condenser, in conjunction with proposed MSIV leakage 22 values exceeding the current fleet maximum values.
23                  The NRC's staff's position is that it is 24 not unreasonable to encourage BWR licensees to limit 25 their maximum allowable MSIV leakage, since these NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
65 1 valves are, in essence, containment isolation valves.
2 Leakage values more than 200 standard cubic feet per 3 hour per valve may be an indication of a substantial 4 valve defect, which should be evaluated and corrected.
5                  A recent review of the current maximum 6 total allowable MSIV leakage values relative to the 7 maximum-per-valve values indicates a fleetwide average 8 ratio approximately two.                  That is, typically, the 9 total allowable MSIV leakage is a factor of two higher 10 than the maximum-per-valve value.
11                  In addition, the total MSIV leakage value 12 of 400 standard cubic feet per hour approximates the 13 total containment leakage for all other containment 14 isolation valves.
15                  Finally,        these      proposed      maximum      MSIV 16 leakage,      MSIV  limitations,            should    be  viewed        in 17 relationship to the original per-valve limit of 11.5 18 standard cubic feet per hour.
19                  If there are no questions, I'll turn the 20 presentation over --
21                  CHAIR PETTI:            Sorry.        John, just one 22 question.
23                  MR. PARILLO:          Sure.
24                  CHAIR PETTI:          Does any of this -- as I 25 recall, when we heard about the topic through the ISG, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
66 1 there was also this discussion of, you know, the 2 worker dose that gets received in trying to make sure 3 you meet these leakage valves, and this idea that 4 workers were taking dose when there was sort of a 5 safety margin because you could trade off of the two.
6                    Is that still part of the thinking here?
7 Because        I  haven't    heard      anybody        talk about      that 8 aspect.
9                    MR. PARILLO:        Yes, that has always been a 10 consideration in balancing the risks of a hypothetical 11 accident versus the actual dose that is incurred in 12 maintenance of these valves.                    Yes, we didn't mention 13 that, but that is always a consideration for any of 14 the license amendment requests for increasing MSIV 15 leakage.
16                    CHAIR PETTI:        Okay.        Thank you.
17                    DR. SCHULTZ: John, this is Steve Schultz.
18                    The second bullet that you have talks 19 about the maximum of 400 standard cubic feet per hour 20 or below.        But, then, the next sentence says, "Higher 21 values will be considered on a case-by-case basis."
22                    You seem in your presentation to have 23 suggested that, you know, we're really going to try to 24 hold to 400.          Where does this come out?                Is it, in 25 fact, true that higher values will be considered on a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
67 1 case-by-case basis?
2                    MR. PARILLO:          Well, Steve, you know, we 3 kind of struggle with this.                    We don't want to make 4 this      a  hard-and-fast        rule.          We    don't  have      that 5 capability. But we just wanted to convey to licensees 6 that, should they propose leakages higher than what is 7 currently the maximum in the fleet, that they should 8 expect to receive a high level of scrutiny.                          I'm not 9 sure how else I could explain that, but --
10                    DR. SCHULTZ:        I would have used the same 11 words.
12                    The next question I had -- thank you, John 13 -- the next question is, and it goes back to our 14 original discussion on Rev. 0 versus Rev. 1, this 15 seems like an area where a BWR licensee might want to 16 use this particular segment of Rev. 1 to perform their 17 evaluation.        Is that something that could be carved 18 out of Rev. 1 and applied to a licensee who's using 19 Rev. 0 for their other evaluations?
20                    MR. PARILLO:        Well, I would probably want 21 to defer that to Mark.
22                    But my sense of it is that that would be 23 an      acceptable    approach,        but      Mark      may  have      some 24 additional concerns on that.                    That's getting back to 25 this idea of having both Rev. 0 and Rev. 1 coexisting.
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68 1                  DR. SCHULTZ:        Yes, that's why I asked the 2 question.      Thank you.
3                  MR. BLUMBERG:          I'm sorry, would you mind 4 repeating the question again?                  I apologize, I didn't 5 catch it all.
6                  DR. SCHULTZ:        Thank you, Mark.
7                  The question is, could a licensee who's 8 using Rev. 0 take advantage of BWR MSIV leakage value 9 reevaluation under the guidance of Rev. 1 without 10 changing their overall evaluations of other non-LOCA 11 events?
12                  MR. BLUMBERG:            So, once again, all the 13 different permutations that could occur, we have not 14 gone through all those scenarios.                    What we have gone 15 through is we've created a guidance that includes a 16 methodology that we find acceptable for meeting the 17 regulations. Like with any methodology, if a licensee 18 proposes      differences      to  that,        then  they  have      to 19 provide additional justification why they think that 20 that continues to meet the regulations.
21                  So, I don't know that on the fly I can 22 give you an answer, but I can just tell you that, if 23 you go outside the method, then you've got more work 24 to do.
25                  DR. SCHULTZ:        Okay.        Thank you.
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69 1                  MR. BLUMBERG:          You're welcome.
2                  So, I don't know; on the schedule right 3 now, it shows 10 o'clock is a break.                      We are at the 4 slides that we expected to get to at that point.                              I 5 don't know if you would consider a break or you want 6 us to continue on.
7                  CHAIR PETTI:          Yes.        No, I had the same 8 thought because we're going to transition a little to 9 talk about other aspects of the Reg Guide.
10                  So, why don't we take a break for 15 11 minutes and be back at 10:25 Eastern time?
12                  (Whereupon, at 10:07 a.m., the foregoing 13 matter went off the record and went back on the record 14 at 10:25 a.m.)
15                  CHAIR PETTI:        Okay.        Hopefully, everyone 16 is back from our break.
17                  Mark, let's continue.
18                  MR. BLUMBERG:          Okay.        We're going to be 19 starting with Dr. Elijah Dickson.
20                  Elijah?
21                  MR. DICKSON:        Thank you, Mark.
22                  I'd like to preface with slide 28 and 29.
23                  We're talking about the MHA LOCA source 24 term        here  for  ATF      high      burnup      and  increased 25 enrichments.      That is the source term to containment.
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70 1                  Can  you    please      move      on  to  slide      29, 2 please?
3                  This slide highlights our work in updating 4 Reg Guide 1.183 to meet, in part, the Nuclear Energy 5 Innovation      and  Modernization            Act,      or  NEIMA,      that 6 requires the NRC to be prepared to license ATF-type 7 designs.      To do so, the staff has identified several 8 improvements to Rev. 0 that can benefit from off-the-9 shelf research and models which expand and encompass 10 near-term AFT designs.            Those would be chromium-built 11 and chromium-cladded fuel designs with burnups up to 12 68 gigawatt days per metric ton uranium peak rod 13 average and enrichments up to 8 percent.
14                  Since      NUREG-1465,            there    have        been 15 substantial        improvements          in      the      computational 16 resources and data available to predict radionuclide 17 release and behavior during severe accidents. And the 18 intention      of  Rev. 1    is  to      cite      the  source      term 19 recommendations from the Sandia National Lab report, 20 SAND-2011-0128, which is patterned after NUREG-1465, 21 which was published in 1995, and forms the technical 22 bases for Rev. 0.
23                  This is a well-documented analysis with 24 two additional supporting technical documents for BWRs 25 and PWRs, respectively.            It has been thoroughly peer-NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
71 1 reviewed.        In short, it's a well-analyzed and vetted 2 update to the accident source term.
3                    We find that, for the MHA LOCA source 4 term,        fission  product        inventory          releases    to      the 5 containment.        They are essentially unchanged due to 6 increases in burnup and enrichment.                          However, an 7 important        distinction        to    make      is  that  there      are 8 variations between the two source terms -- that is, 9 NUREG-1465 and the 2011 Sandia work -- in regards to 10 fission product fractional releases and the timing.
11 These changes are due to differences in the accident 12 scenarios, as modeled in the MELCOR code, as well as 13 modeling improvements.
14                    For instance, we have increased model 15 fidelity.        That is, we can model the core geometry 16 much more finer.              There's state-of-the-art severe 17 accident modeling practices that we've learned since 18 NUREG-1465; specifically, the work that was done in 19 SOARCA. And we have updated test data from the PHEBUS 20 fission product terminal program, as well as separate 21 effects        tests.      All      of    this      leads    to  a    better 22 understanding of severe accident progression, which is 23 captured in this update to Reg Guide 1.183.
24                    In addition to the work done by Sandia 25 National Lab, we did some research with the Office of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
72 1 Research      in  regards        to    assessing        fragmentation, 2 relocation, and dispersal for this in the containment 3 source term, focusing on the particular aspects of 4 radionuclide released fractions and timings.                          It is 5 expected that, for FFRD, it's expected to have limited 6 impact on the MHA LOCA source term, which is a result 7 of a substantial fuel melt of the core. This work can 8 be found in the memo describing this research, the 9 second memo at the bottom of this page.
10                  We do, also, have ongoing research being 11 executed right now with Sandia National Lab to update 12 this 2011 Sandia accident source term analysis to 13 accommodate higher burnups and increased enrichments.
14 And this work is ongoing right now, and it's not 15 expected to be fully completed before the issuance of 16 Reg Guide 1.183, Rev. 1. So, this would initiate some 17 work in regards to updating it again to a possible 18 Rev. 2 to incorporate this work probably several years 19 down the line.
20                  And with that, if there's no questions, 21 I'd move the presentation on to Dr. Paul Clifford.
22                  DR. SCHULTZ:            Elijah, this is just a 23 question about terminology -- well, not terminology, 24 but      what  you've    chosen      in      the    first  sub-bullet.
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73 1 tons, gigawatt days per metric ton, for a maximum 2 assembly. Traditionally, the focus is on the peak rod 3 average.      And I'm just curious as to why you've 4 incorporated a maximum assembly value. Is that coming 5 out of some of the research or something like that?
6 Because I'm not sure it's generically applicable.
7                MR. DICKSON:        Yes, I would have to pass 8 that question on to the Office of Research.                          James 9 Corson or Hossein Esmaili are on the line.                        Perhaps 10 they can answer your question.
11                DR. SCHULTZ:        I just think it raises the 12 issue of what limit should we be thinking about?
13 Assembly average now, if we get assembly average, or 14 a rod average?
15                MR. DICKSON:        With my experience with the 16 vendors, they use these interchangeably.                        Some use 17 peak rod average; some use max assembly.
18                DR. SCHULTZ:              In      their  licensing 19 documents?
20                MR. DICKSON:          Yes, discussions with the 21 vendors who develop the fuel, yes.
22                MR. ESMAILI:        James, are you on the line?
23                MR. BLUMBERG:        So, I can maybe help some.
24 This is Mark Blumberg from the staff.
25                In the regulatory guidance, we cited as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
74 1 peak average.
2                    DR. SCHULTZ:        Peak rod average rod --
3                    MR. BLUMBERG:          Sixty-eight gigawatt days 4 per metric ton uranium peak rod average.
5                    MR. CORSON:        Yes, this is James Corson 6 from the Office of Research.
7                    I think this might just be a typo on the 8 slide.        Because 59 gigawatt days max assembly average 9 burnup, that was the assembly average burnup from the 10 Sandia SAND-2011-0128 calculations.
11                    DR. SCHULTZ:        Okay.
12                    MR. CORSON:                And      that    actually 13 corresponded to a peak rod average of 62 in those 14 particular calculations, but --
15                    DR. SCHULTZ:          That's what I would have 16 thought.
17                    MR. CORSON:          Yes, yes.          So, the memo, 18 though, that is referenced here does discuss why like 19 68 peak rod average is appropriate, even though the 20 calculations only went to 62.
21                    DR. SCHULTZ:        Okay.        Thank you.
22                    MR. CLIFFORD:          Okay.        If  there's        no 23 further questions, my name is Paul Clifford. I am the 24 Senior Technical Advisor for Reactor Fuels in the 25 Division of Safety Systems.
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75 1                And we're kind of changing gears now. And 2 now, we will talk about radiological release fractions 3 to be used in predicting doses for the non-LOCA design 4 basis accidents. And these consist of two components.
5 There is a steady-state component, which is the amount 6 of each radionuclide released from the pellet into the 7 void space of the rod during normal operation as a 8 result of temperature-driven diffusion.                      And then, 9 there is, for certain types of accidents, there is a 10 transient release, almost a burst release, as a result 11 of fuel fragmentation.
12                So, in my presentation, I'll be talking 13 about several updates to the guidance relative to Rev.
14 0, which was, I believe, released in the year 2000.
15                First, we're expanding the applicability 16 of the guidance to support modern fuel utilization and 17 fuel rod burnups up to 68 gigawatt days per metric ton 18 uranium rod average burnup.
19                We're    providing          separate      BWR  and      PWR 20 steady-state    releases,      and    you'll      see  why    that's 21 important later on the presentation.
22                We're introducing a new component, and 23 that is the transient fission gas release, which is 24 based upon, as I mentioned earlier, not like the slow 25 diffusion base release, but, really, a burst release NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
76 1 as a result of fuel fragmentation under prompt power 2 increasing accidents, like control rod ejection or 3 control blade drop.
4                  And  finally,        we    will    talk  about        an 5 acceptable analytical procedure that's provided in the 6 guidance which could be used by the licensees to 7 calculate        plant-specific              steady-state        release 8 fractions.
9                  So, moving on to slide 32, this figure 10 depicts the expanded range of applicability.                              The 11 Y-axis is rod average power for the peak rod, and rod 12 average burnup on the X-axis. You'll see the red line 13 is indicative of what you would find, if you looked 14 at, say, a BWR corroborating limits report for the 15 TMOL.        We've taken the general trends, where you see 16 a knee relative early, towards the end of the first 17 cycle, followed by a second inflection point latter on 18 in rod life.        We've expanded those both out in burnup 19 and up in power to provide flexibility for fuel 20 managers.
21                  The same with the PWR, you'll see the blue 22 line here, which goes out to about 35 gigawatt days at 23 12.2 kilowatts a foot, and then, drops down to 68, 24 which is our target burnup here.
25                  Now, how does this compare to Rev. 0? So, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
77 1 in Rev. 0, there was a footnote, footnote 11, I 2 believe,        which    provided          a      limitation      on      the 3 applicability of the Table 3 non-LOCA steady-state 4 release fractions.              And it stated that they were 5 acceptable, provided your rod power was below 6.3 6 kilowatts a foot beyond 54 gigawatt days.                        So, that 7 limitation is kind of shown in this blue box.                              You 8 have to be below this power burnup envelope when 9 you're in just this limited range of rod life.
10                    As we'll get into in the next slides, 11 you'll        see  that      the      releases        for    different 12 radionuclides are driven at different times in life.
13 So, you know, providing a small window like this is 14 not adequate to really, truly define the range of 15 applicability.
16                    And what was mentioned earlier by Mark, 17 one of the first slides in this overall presentation, 18 was that we've been busy issuing or reviewing and 19 approving        license      amendment          requests    by    various 20 utilities which find themselves outside of this blue 21 box.      In other words, they're operating above the 6.3 22 at this high burnup region.
23                    So, we've issued -- they've come in with 24 license amendment requests with unique methods for 25 calculating release fractions, and we've reviewed and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
78 1 approved these.          So, there was definitely a need to 2 expand both upwards and outwards of these lines to 3 provide          flexibility        to      support          modern        fuel 4 utilization.
5                    So, I ran this code called FAST.                  FAST is 6 the NRC's confirmatory fuel rod thermal-mechanical 7 model.        It superseded FRAPCON and FRAPTRAN that many 8 of you may be familiar with. It represents the state-9 of-the-art modeling of the fuel's response during 10 normal operation and has transic capabilities built 11 into it also.          And it's well-documented and well-12 validated against the latest and greatest empirical 13 database.
14                    The FAST code was also updated to include 15 the ANS-5.4 standard release model.                        The ANS model is 16 used        to    predict      releases          of      the  short-lived 17 radionuclides.          And we'll get into that in the next 18 slide.
19                    So, using this updated version of FAST, we 20 ran      sensitivity      cases        to      determine      what's        an 21 appropriate set of tables and release fractions to 22 provide the industry for use.
23                    Unlike what we talked about earlier, which 24 was kind of a deterministic fuel melt-driven source 25 term that's used for MHI, these steady-state releases NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
79 1 are very sensitive to rod power and fuel rod design.
2 So,      changes    in    either      of    these      inputs  makes        a 3 significant difference on how much of each of the 4 radionuclides are released at different times in life.
5                  And  so,      in  order        to    provide  bounding 6 values, we had to identify which fuel designs were the 7 most limiting, and then, separately, a develop a 8 bounding power history that would, then, be useful to 9 the industry.          In other words, we don't want to 10 provide a rod power envelope which licensees couldn't 11 use because they're outside of it.                        So, we tried to 12 bound both, but with being sensitive to the fact that 13 we can't be overly conservative in our rod power 14 history; otherwise, we're going to put very large 15 release fractions.            And if we do the opposite, then 16 they're not as useful.
17                  So, the most limiting fuel designs are the 18 BWR 10x10 lattice-type and the PWR 14x14 lattice 19 design.        In general, though, the larger-diameter fuel 20 rods at a given power level have higher operating 21 temperatures.          Higher operating temperatures means 22 more efficient gas release from the pellet into the 23 rod plenum.
24                  So, there were several competing effects.
25 As I mentioned, by expanding the operating domain to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
80 1 provide flexibility for fuel management, that promotes 2 higher releases.
3                  Also, we applied modeling uncertainties, 4 because there is a large scatter in data.                            When we 5 validate        these    models,        we      also      quantify        the 6 uncertainties        in    the      models'            capabilities          of 7 predicting things like fuel temperature and fission 8 gas release. So, we want to provide something -- when 9 you're doing a safety analysis, especially a dose 10 calculation to the public, you want to have something 11 that accounts for those modeling uncertainties.                              So, 12 when      you  apply    the      modeling        uncertainties,          that 13 promotes even higher releases.
14                  But, to offset that, the application of 15 the new ANS Standard Release Model provided a lot of 16 benefit relative to the model that was used for the 17 Rev. 0 of the Reg Guide 1.183, which goes back to the 18 1982      standard.      Application          of    this  latest        and 19 greatest model is also a significant decrease in 20 iodine-131 and other short-lived radionuclides.
21                  So, let's get into some results of the 22 calculations.
23                  DR. SCHULTZ:        Excuse me, Paul.
24                  MR. CLIFFORD:          Yes.
25                  DR. SCHULTZ: Before you leave that slide, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
81 1 on    the      previous    slide,      I    believe      slide  32,      you 2 presented the power versus burnup for those limiting 3 curves.
4                    MR. CLIFFORD:          Uh-hum.
5                    DR. SCHULTZ:        And do those correspond to 6 what you would expect to see for these limiting fuel 7 designs that you've evaluated?                      Do you see what I'm 8 saying?
9                    MR. CLIFFORD:            Yes, yes.        No, no, I 10 understand.          Right.
11                    So,  generally,          the      larger  fuel      rods, 12 there's less linear feet of fuel, because they're a 13 larger        diameter,    and      you    generally      have    higher 14 operating kilowatts per foot.                    Of course, it depends 15 on the power density of the reactor and how many 16 assemblies there are, but, generally, that's what 17 happens.
18                    So, when you see a plant go from a 10x10 19 to a configuration to, say, an 11x11, BWR, you've 20 added 21 percent more linear feet of fuel.                        So, your 21 linear heat rate goes down.                    So, you actually would 22 move away from these.                You would be closer to -- you 23 know, you'd be further down here.
24                    So, yes, these power operating histories 25 are representative of those two fuel designs.
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82 1                    DR. SCHULTZ:          Okay.        Thank you.
2                    MR. CLIFFORD:          Yes.
3                    That's      a    problem        with    the  animation.
4 Okay.
5                    BWR, let's talk about some results.                        So, 6 there's different trends in the release of different 7 radionuclides which are important.                        So, here in this 8 figure on the top left -- we're at slide 34 now --
9 here, we show this red line is what we just talked 10 about.        This is the bounding rod power history with 11 the two knees in the curve for the BWRs.                          And that's 12 shown here.
13                    And    then,      on    this      axis,    we  have      the 14 predicted          best-estimate            release        fraction          for 15 krypton-85.          Krypton-85 is a long-lived noble; half 16 life of just over 10 years.
17                    Now what's imported about that is, if 18 you're        generating      fission        gas      back  here    at      the 19 beginning of life, and it's working its way through 20 the matrix to the grain boundaries, from the grain 21 boundaries out into the void space, by the time you 22 get further on in life, up here, that's still active 23 krypton-85.          It hasn't gone anywhere.                    It hasn't 24 decayed.          So,  as    a    result,        you're    continuously 25 accumulating more and more fission gas as it gets NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
83 1 released.
2                    So, as you would expect, later on in life, 3 is when you get your maximum concentration of long-4 lived radionuclides within the void volume.                          And so, 5 that's the trend you're seeing here.
6                    Now, you may ask, why is it diminishing?
7 It's      slightly    diminishing          as    you    get  further        on 8 because there is such a dramatic decrease in rod 9 power.        With this decrease in rod power, there's a 10 significant decrease in fuel temperature.                          As such, 11 you're not really releasing anything.                      And since this 12 is a ratio of what's produced versus what's released, 13 the      release    isn't      changing,          but    the  production 14 continues to go up.            So, you'll start seeing a slight 15 decrease here.        So, that's the trend you would see in 16 the long-lived.
17                    And then, when you get to the short-lived, 18 it    has    a  different      trend.          So,    the  same      power 19 operating limits; the same fuel rod. And here, you're 20 calculating the release rate, which is shown in these 21 yellow squares, for iodine-131.                    Now iodine-131 has a 22 half-life of eight days. So, the active iodine that's 23 produced at the beginning of life, it's not active 24 iodine by the time you get further on in life.
25                    So, the population or amount of active NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
84 1 iodine-131 in the gap is much more dependent on the 2 instantaneous flux; in other words, the production 3 rate.          It  becomes      very      dependent        on  both      fuel 4 temperatures, which govern the release, and then, 5 production rate, which keeps the amount of active 6 iodine available for release.
7                  So, what you're seeing here is the maximum 8 combination of that occurs at the maximum production 9 rate, which is at this first knee of the curve.                            And 10 so, it's building up.            It's building up here, and you 11 may ask, why is it different here than here?                          That's 12 because your burnup has increased; your diffusion 13 coefficients        have    changed,          and    also,  your      fuel 14 temperature        has  changed,        with      the    degradation        of 15 thermal conductivity.              So, as that decreases, fuel 16 temperatures are actually going up, even though you're 17 at the same part.
18                  And there's other competing effects, like 19 irradiation swelling, closure of the gap, things like 20 that. But this turns out to be the maximum point with 21 respect        to  the  production            over      the  release        to 22 calculate the amount of active iodine that's in the 23 gap.
24                  And so, combining the highest release 25 fractions for each of their points in life, you end up NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
85 1 with a table for the BWRs.
2                  Now, how does this compare to what's in 3 Rev. 0? For iodine-131, Rev. 0 had a fraction release 4 of 0.08.        So, even though we pushed this limit up, 5 this knee of this curve was pushed up relative to Rev.
6 0, we were able to compensate for that by using the 7 new ANS standard.          So, there actually ended up being 8 a decrease in the active iodine, which is the most 9 important      of  the    radionuclides              when it  comes        to 10 calculating the effects of radiation on the human body 11 because of the appetite of your thyroid for iodine.
12                  Krypton-85 was 0.10 in Rev. 0, and that 13 has gone up to 0.32.            And that is, again, because we 14 pushed this envelope way up and we pushed it out in 15 burnup.        So, the combination of those two, since it 16 accumulates continuously, is you're getting a much 17 larger value.        But the nobles don't have as a big of 18 an impact.
19                  And we'll get into some more discussions 20 about the alkalis, which is like cesium-134. This was 21 at 0.12, and now it's at 0.16.
22                  So, that's the releases for the BWRs. Now 23 let's go --
24                  DR. SCHULTZ:        Paul?
25                  MR. CLIFFORD:          Yes?
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86 1                  DR. SCHULTZ: Before you leave this slide, 2 you mentioned that the curves are the past best-3 estimate values, as you've shown them.
4                  MR. CLIFFORD:            Right, this value right 5 here.
6                  DR. SCHULTZ:        In the table, the table is 7 derived with some focus on conservatism and other 8 information, as you mentioned, from ANS-5.4?
9                  MR. CLIFFORD:          Correct.        So, to get an 10 upper tolerance value for the long-lived, you're using 11 the empirical database for which the models of FAST 12 were tuned.        So, it's a model-specific fission gas 13 release uncertainty that's being applied for the long 14 term. For the short-lived, there is uncertainty terms 15 dictated by ANS-5.4.
16                  DR. SCHULTZ:        Got you.          I figured there 17 was a transition there. That explains it. Thank you.
18                  MR. CLIFFORD: Okay. So, let's move on to 19 the PWRs.      So, on the top left, it's the same trends.
20 With        respect  to    long-lived,            they    continuously 21 accumulate.        When we get to the short-lived, once 22 again, you're seeing the values peak at the first knee 23 of the curve for the PWRs, noting that the PWR has a 24 knee that curves a lot further out.                      We used the same 25 power level, but, instead of dropping and starting at NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
87 1 20 gigawatt days, it goes out to 35.                      So, you'll see 2 that that has a first order effect on the results.
3                  So, the iodine-131 went from 0.13 for the 4 BWRs to the 0.07 for the PWRs.                    This was really the 5 driving factor, why we wanted to have separate tables 6 for the Bs and the Ps, because, intrinsically, they 7 have      different  operating        regimes,        and there's        no 8 reason why we should, essentially, the BWRs for a 9 different regime than a PWR.                      That was why that 10 decision.
11                  So, we're up at 0.07, but we're still 12 below the Rev. 0 values, once again, due mostly to the 13 introduction of the new ANS standard.                    The krypton-85 14 has just gotten even a little higher because of the 15 shape and the slope of these lines, and the same with 16 the alkalis.
17                  So, in the next section, I wanted to talk 18 about our treatment of cesium.                  The ANS-5.4 standard, 19 either the earlier standard, the '82 standard, or the 20 2011 standard, they provide a means to infer the 21 release fractions of cesium, based upon the predicted 22 release fractions of krypton-85.                        And essentially, 23 they recommend that, since the diffusion coefficient 24 is considered to be twice that for cesium versus the 25 noble metals, that you increase the predicted release NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
88 1 fraction of krypton-85 by the square root of the 2 diffusion coefficient. So, that would be a multiplier 3 of 1.414.        So, in other words, whatever you predict 4 for krypton-85, you multiply that by 1.414, and that's 5 your estimated releases for cesium-134 and cesium-137.
6                    Now, as you saw from the previous values, 7 the value of krypton went way up because of we were 8 pushing the fuel envelope so aggressively.                        And as a 9 result, we ended up with values of cesium that just 10 seemed completely unreasonable and unrealistic.
11                    So, when the ANS model was developed, 12 there was not a lot of data; there was not a lot of 13 like        post-irradiation          examination,          destructive 14 examinations of the chemical compositions of what was 15 in the gap region, you know, targeting cesium versus 16 nobles.        But there have been recent research programs 17 that have done this.
18                    So, looking at these various research 19 programs, we identified there were notable differences 20 between        the  physical      properties          of  the  chemical 21 effects between nobles and cesium.                        Most important, 22 the cesium tends to react with other constituents to 23 form      relatively    immobile        compounds,        which,      then, 24 accumulate along the grain boundaries or can get 25 trapped within the porous, high burnup structure that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
89 1 we refer to as the rim.
2                    And below 1200 C, 2200 F, the cesium is in 3 a liquid form, and generally, forms a film along the 4 grain boundaries.              And that's the temperature you 5 would see in higher burnup fuel, and that's the 6 temperature you would see towards the outer region of 7 the fuel pellet, you know, like at the rim region.
8                    And when it is released, the studies show 9 that the cesium actually can react with the I.D.
10 zirconium cladding.                And that can form, once again, 11 less mobile, stable, non-gaseous compounds. So, that, 12 again, hinders the release.
13                    Because, for krypton-85, what you're doing 14 is you're taking these chemically inert noble gases, 15 which can readily diffuse, you know, from the fuel 16 matrix        onto  the  grain      boundaries,          and  then,      get 17 released into the plenum region in there.                              They're 18 just, essentially, hanging out in the plenum region of 19 the fuel rod, and if you get any perforation in the 20 cladding, there's essentially a puff release.                                  So, 21 they're          available        to    release            because    they're 22 relatively          chemically        inert;        whereas,      the    cesium 23 generally gets trapped along the way.                          And even if it 24 gets out of the pellet, it kind of gets trapped again.
25 So, it's not as readily released.
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90 1                  And so, from looking at all these studies, 2 there were recommendations that were made of the 3 various studies, looking at the ratio of release 4 between the nobles and the cesium. So, we saw between 5 2.2-to-1        release    to    a  10-to-1        release  from      the 6 different studies.            So, conservatively, we chose a 7 lower bound of a 2-to-1 release ratio between cesium 8 and the nobles.
9                  So,  what      you    end      up  with    is      this 10 relationship.          So, you would take your predicted 11 krypton-85 releases and you would multiply them by .5.
12 That's,        obviously,    a    dramatic        change  compared        to 13 multiplying it by 1.414.                  So, this is, it's more 14 realistic        using  better      engineering,        and  this        is 15 generally a release that we identified.
16                  CHAIR PETTI:        So, Paul?
17                  MR. CLIFFORD:          Yes?
18                  CHAIR PETTI:            A question:        given this 19 potential of the clad to hold up fission products if 20 they're cesium, other clads, non-Zircaloy clads then, 21 there would be some more work required to get a 22 database on those, I would think.
23                  MR. CLIFFORD:            Absolutely.        That        is 24 absolutely correct.
25                  As I mentioned, the steady-state releases NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
91 1 are      extremely    sensitive        to      rod    design  and    power 2 history. Any changes that are introduced really would 3 invalidate these. When you're predicting the releases 4 from fuel as a result of a core-wide melt, it really 5 doesn't matter, the operating history, how you got to 6 where you were. You're melting the fuel and releasing 7 everything in the pellet.
8                    CHAIR PETTI:          Right.
9                    MR. CLIFFORD:          Whereas, here, we're just 10 talking about kind of finds its way out of the pellet.
11 And so, it's very sensitive to fuel temperatures and 12 operating history, and then, pellet diameters and 13 other constituents which would cause the noble, cause 14 the various radionuclides to get trapped could all 15 change.
16                    A good example would be a dopant.                  If you 17 added certain dopants into the fuel, those dopants 18 could        interact    differently            with    the  different 19 radionuclides,          which        could      cause    differences        in 20 release.        Maybe sometimes it increases the diffusion 21 coefficient; maybe sometimes it traps it along the 22 grain boundaries.            So, it's very sensitive to that.
23                    And  so,      as    a    result,        the  limits        of 24 applicability are kind of set at the UO2, and we're 25 kind of in an unusual place.                      Because in the past, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
92 1 everybody was using UO2, and it was all roughly the 2 same within the same ASTM specifications.                        And they 3 were all using zirconium tubing, roughly, within the 4 same ASTM specifications, except for some minor tweaks 5 here and there.
6                  And    so,      the    fuel      behavior    was      very 7 similar. Now we have ATF coming, where we're going to 8 have a bunch of deviance, and especially when you 9 start        talking  dopants        and      different    theoretical 10 densities and grain sizes of the different fuel and 11 different materials for cladding.                      They're all going 12 to behave differently.
13                  And I'm not saying we can't model that, 14 and I'm not saying we can't collect the data and 15 validate models and do that.                  But the problem is, all 16 that data is proprietary.
17                  So, I think we're moving to a space in 18 non-LOCA where the NRC's not going to be able to 19 provide a one-size-fits-all because I don't have 10 20 proprietary fission gas release models that I can, 21 then, exercise, and then, publish the results.                                  I 22 mean, I could do it internally to check what the 23 licensees are doing, but I can't publish those results 24 because it's based on proprietary information that I 25 don't own.
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93 1                    And so, that will get us to really one of 2 the next sections, and I think it will share inputs of 3 the analytical procedure we provided.
4                    So, next, we want to move away from the 5 diffusion-based, steady-state release, you know, from 6 the pellet into the void volume of the rod. And we'll 7 kind of get into kind of a transient burst fission gas 8 release, which is not dictated by diffusion. It's not 9 dictated by, hey, you have a transient; you're running 10 at a temperature for a couple of seconds, and that's 11 causing more fission gas.                This is really fission gas 12 that's        released  as      a  result        of  grain  boundary 13 separation and fragmentation of the pellet, and the 14 high burnup structure.
15                    So, there have been many tests which have 16 been conducted, prompt-pile excursion-type testing.
17 And they have collected and reported fission gas, 18 which was released during the accident itself.
19                    And I know this figure is a little busy, 20 but there's a lot of good information.                        Let me just 21 run through it real quick.
22                    So, what we've chosen to do is plot the 23 data as a function of measured transient fission gas 24 release.        So, this is the amount of fission gas that 25 was released during the prompt-pulse test.
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94 1                    And on the X-axis is the change or peak 2 enthalpy increase.            So, it's the change in the fuel 3 enthalpy.        So, it's the amount of energy that was 4 deposited during that brief power excursion.
5                    So, you have different symbols.              You have 6 tests that were conducted at the CABRI reactor in 7 France, the NSRR reactor in Japan, the BIGR and IGR 8 reactors in the Russian Federation.                      You also have 9 designations for PWR rods and BWR rods, and finally, 10 you have a designation for MOX. So, everything that's 11 not labeled MOX would be standard UO2.
12                    And, in addition, we've color-coordinated 13 the results -- green being very low burnup fuel, below 14 30, all the way up to purple, which would be greater 15 than 70.        So, you've got all this data, and there is 16 a large spread in the data, but that's not unexpected.
17 You actually see a pretty good spread in the data for 18 steady-state        fission      gas    release.        So,  it's      not 19 surprising to see a spread in the data for transient 20 fission gas release.
21                    And what the general trends are is, as you 22 deposit        more  energy,        you,      basically,    have      more 23 fragmentation of the pellet, which results in more 24 fission        gas  release.        So,      as    you move  in      this 25 direction, you're seeing more transient fission gas NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
95 1 release.
2                    And if you look kind of going straight up 3 vertically, and if you pick up 100 delta calories per 4 gram, you kind of see a general trend where the higher 5 burnup        rods  for  the      same      energy      deposition        are 6 releasing more fission gas.
7                    As I mentioned, there is some spread in 8 the data, but there are general trends.                          And we've 9 chosen to use engineering judgment and kind of try to 10 capture the leading edge of this correlation, which we 11 feel is a function of both the increased or the 12 deposited energy, which is delta H and burnup, and we 13 came up with these correlations here.
14                    Okay. Moving on to --
15                    CHAIR PETTI:        So, Paul, just for clarity 16 in terms of how this would actually be applied, the 17 whole closedown, seeing these increases, and in these 18 events        there's    a  certain        number        of  rods    around 19 whatever rod ejection event is being postulated.                                Is 20 that --
21                    MR. CLIFFORD:        Yes, it's a very localized 22 event. It generally only occurs in a relatively short 23 axial region of the core, and it's a very localized 24 radially region of the core.                    So, you could see only 25 a few assemblies, and then, only a few feet of each NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
96 1 assembly, and then, based on proximity to the actual 2 rod      that gets  ejected,        you're        going  to    see      a 3 distribution of delta Hs.
4                So,  the      modern      fuel,        the modern      core 5 physics parameters -- the modern core physics models 6 are certainly capable of calculating changes in fuel 7 enthalpy across several different bundles, as a result 8 of an ejected rod; whereas, you're seeing relatively 9 large quantities -- like you're saying, well, 20 10 percent fission gas release, that's a lot. But you're 11 generally not -- it's only being applied to a few 12 axial segments of the rod.              It's not over the entire 13 length of the rod.          So, it's not, when you actually 14 apply it on a rod segment basis, based upon individual 15 delta Hs across the different rods, I don't see it as 16 being a major impact.
17                MEMBER HALNON:          So, Paul, this is Greg.
18                Hey, how do these trendlines get drawn?
19 Just if I put my thumb up to it, I would have made a 20 much flatter line for the red line.
21                MR. CLIFFORD:          Yes.      Yes, I mean, to try 22 to look at this from numerical analysis and do a data 23 reduction, and try to come up with a two sigma, it 24 gets very complicated. It was really just engineering 25 judgment trying to capture the leading edge.                      Yes, I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
97 1 agree,        you  could        draw      these        lines    slightly 2 differently, but that's just the way we chose to draw 3 these lines.
4                    MEMBER HALNON: Could you do a sensitivity 5 analysis to see if you drew it drastically different, 6 but the results would be at least reasonably the same?
7                    MR. CLIFFORD:            So, we did not.            We did 8 not.      And really, it's difficult.                  It's difficult to 9 do a sensitivity analysis on the impact of these 10 fission        gas    release        correlations          because        the 11 transients are very plant-specific and very cycle-12 specific.        And in the margins to the actual offsite 13 and control room doses, it varies widely throughout 14 the industry.        So, if you're saying, well, if I added 15 an additional 3 percent fission gas release as a 16 result of incorporating these various lines to the 17 different rods in the core, how does that impact the 18 offsite doses?          It's impossible to do that.
19                    MEMBER HALNON:          Okay.      So, a steeper line 20 is more conservative, then, to the end result?
21                    MR. CLIFFORD: Correct. A steeper line is 22 more conservative.              So, that's what we tended to go 23 for.
24                    MEMBER      HALNON:            Okay.      That's      good.
25 Thanks.
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98 1                    MR. CLIFFORD:          Okay.        So, the transient 2 fission        gas  release,        as    we    mentioned,    it's        an 3 additional component. So, you have the fissile -- I'm 4 sorry -- you have the radionuclides which are in the 5 fuel rod void and ready for release when the cladding 6 fails.        And now, you have this additional component 7 which you need to add, add to the overall source term 8 when doing your dose consequence assessments.                              As I 9 mentioned previously, it's sensitive, we feel, to both 10 fuel burnup and deposit of energy, and these are the 11 numerical values of the equation that are in the Reg 12 Guide.
13                    Now    the      releases          are  based        upon 14 measurements of the noble gases.                        So, there needs to 15 be some adjustment made for kind for the short-lived, 16 more        volatile      radionuclides,              like  iodine-131.
17 Actually, PNNL had done a sensitivity calculation many 18 years ago, when we went out with DG-1199, and they 19 came up with a relationship between the nobles and the 20 short-lived half-lives radionuclides, like iodine-131.
21 So, there was a correction factor of multiplying it by 22 0.333.
23                    DR. SCHULTZ:          Paul, just to go back to 24 what you said on the previous slide, that is, if one 25 looks at the accidents, as Dave mentioned, that have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
99 1 a relatively localized effect of a transient power 2 increase, that your judgment is that the additional 3 dose associated with that event is likely to be small, 4 a small effect on the overall evaluation?
5                  MR. CLIFFORD:              That    would      be      my 6 assessment, that's correct.
7                  DR. SCHULTZ:              And      then,  the      other 8 statement you made -- and I didn't interrupt you at 9 that point -- but you said that, I think you used the 10 term      that  there's      wild      differences        amongst        the 11 evaluations of doses that are calculated across the 12 licensees' reactors.
13                  MR. CLIFFORD:          Uh-hum.
14                  DR. SCHULTZ:        That kind of surprises me, 15 that that would be the case.
16                  MR. CLIFFORD:        So, the licensing basis of 17 each plant is considerably different.
18                  DR. SCHULTZ:        Yes.
19                  MR. CLIFFORD:          It really depends on how 20 old the plant is; when they were licensed; what 21 they've        done since      then;      how    many    licenses      they 22 request.
23                  DR. SCHULTZ:        Okay.
24                  MR. CLIFFORD:            But they have altered a 25 source term.        Maybe they don't alter a source term.
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100 1                    DR. SCHULTZ:        Okay.
2                    MR. CLIFFORD: But they may have different 3 wind        patterns.          You      know,        there's    different 4 atmospheric          dispersion        factors          in  their      dose 5 calculation.
6                    DR. SCHULTZ:        That's okay.          I understand 7 what you've said now.              It's over the decades, and so 8 forth?
9                    MR. CLIFFORD:          Yes.
10                    DR. SCHULTZ:        Over the methodologies that 11 have been applied.
12                    Thank you.
13                    MR. CLIFFORD:          Okay.        So, the next topic 14 is the analytical procedure.                  As I mentioned earlier, 15 and as Mark mentioned at the beginning, we've actually 16 reviewed and approved several applications that have 17 come      in,  most  notably,        from      Duke    Energy.        They 18 submitted almost their entire fleet, where they could 19 not reach that range of applicability that's in Rev.
20 0, that 6.3 kilowatts per foot, about 54 gigawatt 21 days.        They just couldn't meet it.
22                    So, they had to come in and say, "I want 23 to calculate some new values, but I don't even know 24 where to start."            Because like how am I going to 25 replicate something that the NRC did 20 years ago that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
101 1 wasn't even well-documented?
2                  So, we've looked at different -- and it 3 wasn't just Duke.            There were other utilities which 4 have also run into that limit.
5                  And as a result, we felt, hey, why not 6 publish        an  acceptable        analytical          procedure        for 7 calculating these values?                    And this significantly 8 reduces        regulatory        uncertainty            because    now      the 9 licensees can say, okay, well, maybe I can't live with 10 what the Reg Guide proposes, or maybe I don't want to 11 live with them; it's not like I can't, but maybe I 12 want to seek some margins in my dose calcs.
13                  I mean, there's a big difference between 14 predicted steady-state releases for the same power 15 level        between  a    14x14      CE    fuel      rod  and  a    17x17 16 Westinghouse fuel rod. So, if I'm a licensee and I've 17 got a 17x17, I don't want to use the 14x14.                          So, I'm 18 going to go off and calculate my own.
19                  And    maybe      my    operating        domain      looks 20 differently.        Maybe my fuel management schemes are 21 significantly different.                So, the knee of that curve 22 moves dramatically, and I can capture some of that 23 margin by using something more realistic.
24                  And as the NRC, we can't publish every 25 variant because there's so many variables here between NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
102 1 fuel      rod  design    and      operating          history.          We're 2 providing a generic bounding assessment that someone 3 can use, if they don't want to go off and do their own 4 calculation.        Like it's a default.
5                  But  if    people        want      to    do  their      own 6 calculations, which I believe most of the industry 7 will, then we're providing this means for them to do 8 that, and it needs for us to review it.
9                  So, each of the vendors have approved fuel 10 thermal-mechanical codes which are well-validated and 11 verified and have quantified uncertainties.                        And they 12 can use those methods, incorporating the 5.4 standard, 13 the ANS-5.4 standard to calculate the short-lived 14 ones.        And that provides them with a tool.
15                  And this provides them with a lot of 16 flexibility, but it also supports things like high 17 burnup.        If a vendor has an approved code that's 18 approved after 72 gigawatt days, then they can use 19 that code to calculate release fractions that are good 20 up to 72.
21                  So,    you      know,        as      they    get      more 22 information,        and  they      quantify        their    models        and 23 validate their models, they can use that to, then, 24 predict better release fractions, and use that in 25 their dose calculations.
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103 1                    And as I mentioned, the benefits are the 2 licensees can use plant-specific values, recover some 3 margin.        They can use that margin for high burnup, or 4 wherever else they need that margin.                      And once again, 5 it really reduces regulatory uncertainties.                          So, we 6 don't have people doing it all different ways that we 7 wouldn't find acceptable, once it's finally submitted 8 to us.        So, that's the basis for what we issued the 9 analytical procedure.
10                    And here's my last slide here.                    I just 11 kind of wanted to talk briefly about high burnup and 12 increased enrichment in ATF.
13                    So, DG-1389 provides the revised steady-14 state fractions that are applicable to an expanded 15 operating domain, which helps support high burnup 16 applications and increased enrichment, because you're 17 moving the power levels both up and out.
18                    There's a significant relaxation in the 19 short-term iodine-131, in the long-term cesium-137 20 radionuclides          which      were      achieved      through      more 21 rigorous engineering, meaning applying the latest and 22 greatest ANS releases and, also, going off and looking 23 at some available research data, and making some 24 reasonable assumptions on holdup of cesium.
25                    The  Reg    Guide      provides      an  acceptable NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
104 1 analytical          procedure      for    calculating          application-2 specific          release    fractions,          and      that  definitely 3 supports high burnup in ATF applications.
4                    And once again, we're going to end up with 5 a lot of different proprietary fission gas release 6 models out there.            And it's difficult for the NRC to 7 use that information and publish results. It's really 8 going to be, as we deviate from UO2 and we start 9 adding        a  lot  of  things      to      the    fuel  or  changing 10 materials for the cladding, the vendors, it's going to 11 be their responsibility to generate proprietary -- let 12 me rephrase that.            They're going to have to publish 13 results          using  their        proprietary          models.            And 14 certainly,          there's      a    lot    of      flexibility        that's 15 provided to the industry by providing that guidance.
16                    And finally, combined with the Reg Guide 17 1.236 coolability limits, the transient fission gas 18 release components provide a means to address FFRD for 19 the      postulated      PWR    control        rod    ejection    and      BWR 20 control rod drop access.                  And that, again, supports 21 high burnup and increased enrichment applications.
22                    So, that's the conclusion of my slide 23 presentation.          Are there any questions before we move 24 on to Elijah and fuel-handling accident?
25                    DR. SCHULTZ: Paul, this is Steve Schultz.
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105 1                On    the        second          bullet,    the        ATF 2 applications, I mean, there's several different types 3 of applications, ATF applications, that have been 4 proposed.      But the Reg Guide, the Rev. 1 is focusing 5 on a couple of those in terms of the overall -- the 6 current evaluations that the staff has done.                    Is that 7 correct?
8                MR. CLIFFORD: Not entirely. So, the fuel 9 melt MHA source terms have deemed appropriate for high 10 enrichment and different fuel designs such as chromia-11 doped and chromium-coated.
12                DR. SCHULTZ: So, it's "such as." So, the 13 other designs that have been proposed aren't included 14 at this point?
15                MR. CLIFFORD:          Correct.      And neither of 16 those designs are approved for the non-LOCA. In other 17 words, the Table 3 and 4, I believe, which is the non-18 LOCA steady-state releases, are not applicable to 19 doped fuel or coated cladding right now.
20                DR. SCHULTZ:        And your second sub-bullet, 21 the relaxation and flexibility may be achieved, how is 22 the staff going to review and validate the submittals 23 of the licensees?
24                MR. CLIFFORD:            So, well, the licensees 25 would need to submit an LAR because they're making a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
106 1 change, a fundamental change, to the assumptions in 2 their dose analyses.              So, we would expect that they 3 would use their approved codes, such as GALILEO or 4 PRIME or PAD5, and along with the ANS-5.4 model, 5 follow        the procedure        that's      in    the appendix        and 6 calculate design-specific or plant-specific release 7 fractions.          And then, we would review and approve 8 those.
9                    DR. SCHULTZ:        And you'd use that with the 10 results from FAST?
11                    MR. CLIFFORD:            I mean, we could.                We 12 certainly could run some confirmatory calculations 13 with FAST.        I'm not sure it would be needed all the 14 time because we've done so many cases already with 15 FAST.        So, we understand the sensitivity.
16                    So, if someone came in and said, look, I'm 17 moving that knee of the curve back and down a little, 18 and, oh, by the way, instead of 7 percent iodine 19 releases, I'm now getting 6 percent, or 5 percent, 20 we'd be like, well, that's obviously in the right 21 direction.
22                    DR. SCHULTZ:          Understood.        That's good.
23 Thank you.
24                    MR. CLIFFORD:          Okay.
25                    CHAIR PETTI:          I just want to note that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
107 1 we're running a little behind.
2                  MR. CLIFFORD:          Don't blame me.
3                  CHAIR PETTI:        We still have industry, and 4 we have significant public comment.
5                  Keep going.
6                  MR. DICKSON: Okay. I'll make this short.
7                  With the interest of updating Reg Guide 8 1.183, we decided to revisit the Appendix B fuel-9 handling analyses.            This slide highlights our work 10 that incorporates lessons learned for the last 22 11 years, or 21 years, of reviewing the fuel-handling 12 accident analysis for the alternative source term.
13                  To give you a little bit of history, the 14 FHA can require significant staff resources, as well 15 as from the licenses, and in some cases become a 16 limiting      accident,      despite        a    low safety      risk 17 significance.
18                  What we decided to do is revisit the old 19 studies supporting the current FHA analysis, much of 20 which stems back from work that was done back in the 21 early '70s.
22                  The old model retained, Rev. 0 retained 23 much of the old models from Reg Guide 1.125, which is 24 a very, very old Reg Guide, and assumed that the 25 release to the environment is a puff release, where 5 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
108 1 percent of the gas gap for iodine is available for 2 release.
3                    With the new model, we developed a new 4 model that's based off of improvements in our current 5 understanding of the reactor fuel in physics, iodine 6 chemistry, and the environmental conditions under 7 which fuel-handling operations are taking place.
8                    Also, this would lend to changing the 9 chemical speciations from the two different models.
10 So, the current Rev. 0 model assumes that iodine is in 11 a gaseous form.        The new model recognizes that, under 12 fuel-handling operations in which the pool water is 13 between 100 and 180 degrees Fahrenheit, that cesium 14 iodide is actually -- the predominant form of iodine 15 is in cesium iodide form.                  Under these conditions, 16 it's in a solid form.
17                    So, the revised model assumes that the 18 entire fuel pin gap inventory of iodine is available 19 for release.        Ninety-five percent of it -- I'm sorry, 20 can we go to the next slide, slide 43?                    There we go.
21                    So, the revised model assumes that the 22 entire fuel pin gap inventory, 100 percent of that 23 iodine is available for release to the pool within two 24 different phases.              The first is the initial puff 25 phase.        The second would be the re-evolution phase of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309      www.nealrgross.com
 
109 1 cesium iodine, given the low pH of the spent fuel pool 2 or the reactor cavity.
3                  In general, the alternative fuel-handling 4 accident      models    compute        estimated          radiological 5 consequences that are around 90 to 95 percent lower 6 than that of the current models.                  So, we did identify 7 that there is considerable margin due to the scrubbing 8 effects of iodine in the spent fuel pool, as well as 9 considering the chemical speciation of iodine being in 10 a solid form under these conditions.
11                  And    with      that,        I      will    pass        the 12 presentation on to Mr. Mark Blumberg.
13                  MR. BLUMBERG:          Thank you, Elijah.
14                  Can we go to slide 44, please?
15                  So, since we provided you the DG, a couple 16 of issues have come to our attention that we're 17 considering      addressing.            This      slide    shows      those 18 potential changes that we are considering 19                  The first of these is in response to an 20 issue that was recently identified.                          It is being 21 considered because the currently proposed Regulatory 22 Position for release fractions might benefit from 23 additional clarification to directly state that the 24 applicability of these particular release fractions do 25 not pertain to fuel with iron, chromium, aluminum NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
110 1 alloy cladding.
2                The second is with respect to atmospheric 3 dispersion factor modeling. We are considering adding 4 guidance to address facilities that need to determine 5 offsite atmospheric dispersion factors for distances 6 less than 1200 meters.            This change provides some 7 guidance to address those situations.
8                I want to emphasize that these changes at 9 this point have not been added to DG-1389, but we 10 wanted to make the ACRS and the public involved in 11 this meeting aware that we are considering these two 12 additional changes.
13                Can we go to slide --
14                MEMBER HALNON:          Mark, I'm sorry, this is 15 Greg.
16                Would that include the new 1.249 Reg Guide 17 for ARCON?
18                MR. WHITE:        Yes.
19                MR. BLUMBERG:          Yes.
20                MR. WHITE: That's what it's referring to.
21                Sorry, this is Jason White.
22                MEMBER HALNON:          Okay.        I didn't see the 23 number up there.      That's why I was wondering.
24                MR. WHITE:        Yes, we didn't reference the 25 number because it hasn't been issued yet.                    It hasn't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
111 1 been approved.          So, it's still --
2                    MEMBER HALNON:            It's supposed to be this 3 month, I think.
4                    MR. BLUMBERG:          It's imminent, I believe.
5                    MEMBER HALNON:            Right.
6                    MR. BLUMBERG:            Okay.        Can we go now to 7 slide 45?
8                    So, I wanted to summarize for you some of 9 the key takeaways from our presentation.
10                    Currently, Reg Guide 1.183 does not have 11 an acceptable main steam line aerosol deposition level 12 provided in it.
13                    To  address        future        changes    for      this 14 particular pathway, DG-1389 provides guidance for 15 three acceptable means to align aerosol deposition 16 levels.
17                    DG-1389 also provides guidance to support 18 a streamlined approach for licensees to credit an 19 alternative pathway for BWR MSIV leakage to the main 20 condenser.
21                    The staff are prepared to review near-term 22 ATF designs and burnups up to 68 gigawatt days per 23 metric        ton  uranium        peak      rod      average    and    U2-35 24 enrichments up to 8 percent.                      Future updates to the 25 guide are expected to accommodate higher burnups and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
112 1 enrichments.
2                    The  range      of    applicability        has      been 3 expanded to support modern fuel utilization.                                The 4 impact        of  this    expanded          operating      domain          on 5 radiological        source      terms      has    been    minimized        by 6 applying more rigorous engineering judgment.
7                    Lastly, an acceptable analytical procedure 8 for        predicting      plant-specific                radiological          or 9 radionuclide release fractions has been included and 10 provides flexibility in margin recovery.
11                    I'll now turn the presentation over to 12 Sean for our last slide.
13                    MR. MEIGHAN:          Hi.        My  name  is      Sean 14 Meighan.        I'm the Project Manage for the update to 15 Regulatory Guide 1.183, and I'll be quickly reviewing 16 the next steps for the update to the Regulatory Guide.
17                    So, looking forward, the next step would 18 be to issue the Draft Regulatory Guide 1389 for 60-day 19 public comment, and that's going to occur fairly soon, 20 in April of 2022, this year, so next month.
21                    The next step after that is staff review 22 and disposition of the public comments.
23                    And  then,      we    will      update  the      Draft 24 Regulatory Guide, as necessary, in response to the 25 public comments.
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113 1                The next step after that would be the 2 Office of General Counsel review of the final draft, 3 followed an ACRS full Committee meeting.
4                And    then,      in      the      November-December 5 timeframe of 2022, we expect the issuance of Reg Guide 6 1.183, Revision 1.
7                And  that's        the      next    steps,    looking 8 forward.
9                So, I turn it back to Research, or ACRS, 10 for the rest of the meeting.
11                Thank you.
12                MEMBER HALNON:          Hey, Sean, this is Greg 13 Halnon.
14                When we get to the full Committee, and it 15 may be a "no, never mind," but I'd like to hear it at 16 least the effect of the flow accelerated corrosion 17 programs on the seismic rigidity and other conditions 18 that may be latent in the secondary system that we're 19 taking credit for for holdup.                    Like I said, just 20 consider it, and it may be a "no, never mind," but 21 when we get to full Committee, I'd like to at least 22 hear that piece of it.
23                MR. MEIGHAN:        Okay.        So, I'll with Steve 24 and Shilp to ensure that that aspect is addressed in 25 the full Committee meeting.
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114 1                  Go ahead.        I heard somebody hopping in.
2                  (No response.)
3                  CHAIR    PETTI:              Members,      any      other 4 questions?
5                  (No response.)
6                  Thank you, Staff.
7                  And now, let's turn -- I don't know who it 8 is.      NEI is going to speaking.
9                  MS. PIMENTEL:          Hi.          This is  Frankie 10 Pimentel from NEI.          Can everyone hear me?
11                  CHAIR PETTI:        Yes, we can.
12                  MS. PIMENTEL: Okay. I want to be able to 13 share my screen also.            Okay.
14                  MEMBER REMPE: Is it Michael that needs to 15 quit sharing his screen, in order to enable Frankie to 16 share hers?
17                  MR. EUDY:        Frankie, this is Mike.                I've 18 got your slides.          If you're having trouble, I can 19 probably just display yours for you.
20                  MS. PIMENTEL: Okay. That would be great.
21 Thank you.
22                  MR. EUDY:      All right.          Bear with me for a 23 second.
24                  (Pause.)
25                  MS. PIMENTEL:            So, while we wait, Greg NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
115 1 Broadbent will be presenting our slides.                          So, I'll 2 just provide our opening remarks.
3                  So, good morning to everyone, and thank 4 you,        Subcommittee    Chair        David        Petti  and      other 5 Committee Members, for allowing us to participate in 6 today's meeting, and provide comments on the proposed 7 draft of Reg Guide 1.183.
8                  The proposed revision to this Reg Guide is 9 important to the industry to support ATF objectives.
10 Dose analyses are a long lead-time item for the 11 analyses and the NRC review. So, clarity as timely as 12 possible on this proposed guidance is important to us.
13 We don't want unnecessary delays, but we recognize 14 that some changes may be needed to be able to use the 15 guidance provided in this revision.
16                  Some fuel suppliers and utilities have 17 described the dose consequences resulting from the 18 Draft Reg Guide 1.183 as challenging.                        Based on the 19 proposed      draft,  select      sites        have    performed      dose 20 calculations using the revised release fractions, 21 timing, and elemental groupings, assuming no increase 22 in    burnup.      For    these      select        sites,  dose      rates 23 increased, significantly so for BWR units. While this 24 isn't a systematic evaluation of the fleetwide impact, 25 these results indicate that the current licensing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
116 1 bases source terms will require revision.
2                While industry desires how the proposed 3 Revision 1 of Reg Guide 1.183 establishes methods for 4 achieving 68 gigawatts days per MTU on peak rod 5 average burnup, industry views the proposed revision 6 as introducing unnecessary conservatism to the dose 7 consequences for PWRs and BWRs in doing so.
8                The  intent      of    our    presentation    is      to 9 provide additional insights from our initial concerns, 10 from our review of the proposed revision.                          Greg 11 Broadbent, Senior Staff Engineer in Nuclear Analysis 12 for Entergy will be providing our presentation.
13                Thank you.
14                MR. BROADBENT:          All right. Can you hear 15 me?
16                CHAIR PETTI:        Yes.
17                MR. BROADBENT: Okay. So, you can back it 18 up a slide. That's the slide we'll start with.
19                But, first of all, I'll say, good morning.
20 I'm Greg Broadbent.          I work for Entergy as a Senior 21 Staff Engineer in the Corporate Nuclear Analysis Group 22 in Jackson, Mississippi.
23                And I'll say that the industry recognizes 24 that the staff has expended significant resources on 25 developing this Draft Reg Guide, and our goal is to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
117 1 ensure that this effort results in a product that all 2 currently licensed plants and applicable new plants 3 can apply well into the future.
4                  So, starting with slide 2, which you're 5 on, the proposed changes in the Draft Reg Guide are 6 significant.        The impact of a transition to the 7 proposed guidance on the plant licensing bases can be 8 wide-ranging,      including        offsite        and control      room 9 doses, environmental qualification, containment or 10 drywell spray timing, and suppression pool chemistry.
11                  These proposed changes are considered as 12 significant as those introduced with the original 13 transition from TID-14844 source term to the AST.                          As 14 such, the intended and unintended impacts of these 15 changes and the release fractions and timing should be 16 evaluated and shared with the industry before the Reg 17 Guide        is issued,    similar        to      those  re-baselining 18 analyses performed for the AST introduction some 20 19 years ago.
20                  As I'll discuss later, the conservatisms 21 and changes in the current version of this proposed 22 Reg Guide may preclude its use by many plants that 23 need to implement ATF design concepts with extended 24 fuel burnups and enrichments.
25                  All right.        Slide 3.
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118 1                  A primary concern for the industry is the 2 significant change in the release fractions for the 3 maximum        hypothetical      accident.            The  BWR    release 4 fractions for the MHA nearly doubled for the iodines, 5 which currently comprise a large fraction of the 6 calculated offsite and control room doses.
7                  Contributing to this is the eightfold 8 increase in tellurium, which eventually decay in the 9 iodine.        The bases for these large increases are not 10 well-understood,        but      appear      to    be  driven    by      the 11 scenario selection and pool scrubbing assumptions in 12 Sandia's evaluation and report SAND-2011-0128.
13                  The core damage scenarios applied in the 14 development of the proposed release fractions are more 15 heavily        weighted    to    station        blackout      sequences 16 compared to those previously used in NUREG-1465, in 17 spite of the industry's recent efforts in minimizing 18 SBO risk with post-Fukushima modifications.
19                  It's also unclear from the Sandia report 20 whether the beneficial effects of suppression pool 21 scrubbing have been credited to reduce the airborne 22 release fractions for these SBO scenarios.
23                  In addition, the proposed models for BWR 24 steam line deposition appear to be more conservative 25 than those currently approved for many BWRs.                            And I NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
119 1 think Mark had mentioned that in his presentation.
2                  As a result of these conservatisms, BWRs 3 may find it difficult to meet the proposed Reg Guide 4 without significant licensing basis changes or plant 5 modifications. These changes could include reductions 6 in MSIV or containment leakage rates, or tighter 7 control room boundaries that may result in higher 8 occupational exposures to meet these more stringent 9 limits with little or no safety benefit to show for 10 it.
11                  All right.        Next slide, please.
12                  Considering the potential difficulties for 13 some plants in implementing the proposed guidance, the 14 industry wants to ensure there's clear acceptance of 15 remaining with the current guidance.                        Specifically, 16 what        analyses  or    inputs        would        need  additional 17 justification        or    modification              for    applications 18 submitted under the current Reg Guide revision?                              Are 19 the currently approved non-LOCA gap fractions and main 20 steam line deposition models still acceptable, or 21 would a plant be forward-fit to the updated guidance, 22 particularly with regard to steam line models?
23                  For  those      plants        remaining      with      the 24 current guidance, we would also like to ensure there's 25 a clear path for selective implementation of certain NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
120 1 aspects of Revision 1, such as the revised fuel-2 handling accident. We recognize that Section 1.2.2 of 3 the proposed guidance describes this approach.
4                  All right.        Next slide, please.
5                  Appendix B documents a revised approach to 6 analyzing the radiological consequences of a fuel-7 handling accident. This model was based on a thorough 8 review of decades of historical information and the 9 current understanding of aqueous iodine chemistry.
10 This model was not developed based on an industry 11 proposed application, but, instead, was internally 12 developed by the NRC.
13                  The industry wants to acknowledge the 14 staff's        effort,  and      based    on    our  review  of      the 15 proposed model, we agree with the staff that this 16 approach is a more accurate characterization of the 17 low consequences expected from this event.
18                  Slide 6.
19                  And this is just with regard to the DPO 20 that was mentioned back in November on ISG-2021-01.
21 There was some discussion of a DPO regarding potential 22 inadequacies associated with credit for steam line 23 deposition.        And the industry wants to be proactive 24 and work to ensure that any valid NRC concerns are 25 addressed.        However, there's been no communication NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
121 1 with the industry regarding the details or status of 2 this DPO.        And we would like to know if this DPO will 3 affect the revised seismic analysis in the proposed 4 guidance or the schedule for releasing the proposed 5 Reg Guide.
6                  Any questions for me?
7                  Oh, and I'd like to just mention, with 8 regard FAC, the Flow Accelerated Corrosion issue that 9 came up recently, the BOP systems do have models for 10 flow accelerated corrosion.                  We have a FAC program at 11 all our sites.          We have models that identify areas 12 where        FAC  is  more    likely      to    occur. These      are 13 typically areas where there's two-phase flow and the 14 droplets tend to generate the FAC on the inside of the 15 piping.
16                  And    we      do    inspect          at some    certain 17 frequency, as recommended by our models, that piping 18 to      ensure    that    the      minimum        wall  thickness        is 19 maintained.        And obviously, if that piping does fail 20 during operation, it will be something that we would, 21 obviously, recognize, and we would potentially have to 22 shut down to fix.
23                  So, I just to, at least from the industry 24 perspective, address that FAC issue.
25                  So, any questions for me?
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122 1                  MEMBER BALLINGER:              This is Ron Ballinger.
2                  I guess the question is, the FAC program 3 identifies        in  principal          areas        where  you've        got 4 decreased        wall    thickness,            not      necessarily      wall 5 thickness reduction that would threaten the sort of 6 operational integrity.                But do these results get 7 factored into the analysis that you're doing here? In 8 other words, you're now starting out with a pipe which 9 has a different wall thickness.
10                  MR. BROADBENT: Yes, I'm not sure what the 11 answer to that is.              You know, we ensure that we do 12 maintain that minimum wall thickness, and I would 13 think that the requirements would be to only maintain 14 that wall thickness.
15                  MEMBER HALNON:            This is Greg.
16                  That's why I thought we should have the 17 staff consider it and bring us back the answer to it.
18 It's small piping, typically, but, still, think about 19 it, and then, we'll go forward.
20                  Greg, I guess this is just a lingering 21 question in my mind. NEI, is Entergy a member of NEI?
22 Are      you  speaking    for    the    industry?        Or  are      you 23 speaking for Entergy?
24                  MR. BROADBENT:            I'm in the NEI Working 25 Group        that's  addressing          this,        and  I'm  also      the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
123 1 Chairman of the BWR Owners' Group Committee that is 2 associated with this Reg Guide.
3                MEMBER HALNON:          Okay.
4                MR. BROADBENT:          So, I'm speaking for the 5 industry.
6                MEMBER HALNON:          Okay.      Thank you.
7                CHAIR PETTI: Any other comments, Members?
8                MEMBER BROWN:          Yes.      I just went off and 9 looked at the implementation part of the Reg Guide.
10 Normally, that's a little bit shorter and there's no 11 backfit and no forward.            You know, it's very clear 12 statements. And this is a very long statement that is 13 somewhat fuzzy, like the staff does not intend or it's 14 not a declarative statement, is all I'm saying.                        So, 15 I'm not so sure whether this supports the industry's 16 concern or inflames the industry's concern.                    I'm not 17 trying to say that in a inflammatory manner.                          It's 18 just that it's a very vague, long implementation, 19 based on what I've seen in other Reg Guides, where 20 we've so no backfit is there.                  And I understand the 21 industry's concerns.
22                So, I think somehow this needs to be 23 defined a little bit more conclusively, if you look at 24 the NEI's statement, or the industry's statement on 25 the earlier slide, if I go back to it, where the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
124 1 overall impact should be evaluated prior to issuing 2 it.
3                  It seems to me the backfit considerations 4 are going to be open to the public, open to some 5 comments.        So, I'm just raising that now, so that you 6 can address that, and stuff, in the future.                      This is 7 for the staff.
8                  MR. EUDY:        Did the staff understand the 9 question? This is Mike Eudy. I'm just trying to take 10 notes.
11                  MEMBER BROWN:          It was really a comment, 12 based on reading the implementation paragraph on page 13 34.
14                  MR. EUDY:        Yes, Section D?
15                  MEMBER BROWN:          Yes, Section D.
16                  MR. EUDY:        I can pull it up, if you'd 17 like.
18                  MEMBER BROWN:          If you can put it on the 19 screen, that would be fine.                        That way, at least 20 somebody can see what it says.                        It's pretty long, 21 though.        It's most of the backfit.
22                  MR. EUDY:        Yes.        Just for perspective, 23 it's pretty much our boilerplate for all Part 50.52 24 Reg Guides, this section.                It's much shorter than it 25 used to be.        But this is up-to-date with the current NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
125 1 MD 8.4 and our considerations of backfit and forward-2 fit.
3                  MEMBER BROWN:            Well, do you think this 4 satisfies the industry concern, that they don't have 5 to abide by that and make plant changes, based on the 6 issuance of this Reg Guide, and go back and reassess 7 all      the  statements      that      they      made  during      their 8 presentation?
9                  MR. EUDY:        Well, the main point is that, 10 you know, Reg Guides used to be just forward-fit. And 11 now, it's not.      It's, basically any applicant can use 12 any regulatory guidance that's out there, even if it 13 predates this one.          They would just need to justify 14 it.
15                  So, we're pretty clear.                We're trying to 16 be pretty clear on that.                And if they believe that 17 there's an issue, like if we bring up this Reg Guide 18 during a review and say, "Well, you're not consistent 19 with Reg Guide, Revision 1," they have remedies for 20 that, based on the language in MD 8.4.                      So, really, 21 you know, unless a previous                revision of a Reg Guide 22 is withdrawn for maybe safety reasons, you know, they 23 can still use previous revisions.
24                  Now this Reg Guide, as discussed earlier 25 -- someone can jump in, if they'd like -- how they're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
126 1 going to coexist together, it's a little complicated 2 with this particular Reg Guide.
3                MR. BLUMBERG:        Yes, this is Mark Blumberg 4 of the staff.
5                Every  plant      has      their    own design      and 6 licensing basis.      So, making general statements about 7 what can be and what can't be done is very, very 8 difficult.
9                But there are differences between the two 10 sets of guidances and what they're applicable for.
11 And so, in those instances, a plant that wants to use 12 Rev. 0, but they want to apply it to fuel that has 13 higher burnups than the applicability for Rev. 0, then 14 they need to justify why that's appropriate. It's not 15 just a slam dunk, you know, I've got new fuel, and 16 now, I want to use Rev. 0, because Rev. 0 was not 17 designed for those higher burnups.
18                MEMBER BROWN:          Has Rev. 0 been used for 19 any plants that want to use it for higher burnups?
20 Has that been through staff review at any point yet 21 for BWRs?      Or PWRs, if that's applicable?
22                MR. BLUMBERG:            Paul, would you like to 23 address that for the non-LOCA?
24                MR. CLIFFORD:        Well, we have not received 25 any applications for burnup extension.
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127 1                  MEMBER BROWN: So, it's an open issue? Is 2 that a correct assumption or statement by me?
3                  MR. BLUMBERG:          What do you mean by "open 4 issue"?
5                  MEMBER BROWN:        Well, I mean, if it hasn't 6 been applied, then the staff has not had to address 7 the use of Rev. 0 for higher burnups for existing 8 plants.
9                  MR. BLUMBERG:        That's a correct statement 10 for higher burnups.
11                  MEMBER BROWN:          So, industry's considered 12 thoughts are that this could be a problem for them.
13 Because it sounds like there's a lot of work for them 14 to adapt this on a forward-fit basis, the higher 15 burnups.      Is that --
16                  MR. BLUMBERG: The intent of this revision 17 to the Regulatory Guide was to provide a methodology 18 to address that, to, basically, create less work.
19 That is the intent of the guidance, is for efficiency 20 and effectiveness of regulation.
21                  MEMBER BROWN: Okay. All right. So, it's 22 going to go out for public comment, and we'll see what 23 falls out.        I guess that's the conclusion I'd draw 24 from that.
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128 1 me, let me know.      I'm just kind of looking at it from 2 a non-source-term development person.
3                I'm done.
4                MR. BLUMBERG:          Okay.
5                MEMBER HALNON:            So, Greg, back to Greg 6 from Entergy.
7                MR. BROADBENT:          Yes, I've got it.
8                MEMBER HALNON: You mentioned earlier that 9 this was going to discourage the BWRs from trying to 10 get their MSIV leakage.          Do you think that or do you 11 have a feel for whether or not, if Rev. 1 existed on 12 its own and there was no Rev. 0, that we would not see 13 any more ASTs coming in for leakage?
14                MR. BROADBENT:              That would be likely.
15 Yes, if they play with the new release fractions for 16 BWRs, you know, doubling the iodine release -- I mean, 17 iodine is the main contributor to doses and most of 18 our analyses. So, if you double that release, you're 19 significantly increasing the outside and control room 20 calculated doses.
21                And then, you throw in changes to the 22 steam line deposition models, as I think previously 23 mentioned, some of the earlier revisions are not 24 considered acceptable to the staff anymore.                      That 25 would make it very difficult for BWRs to come in with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309    www.nealrgross.com
 
129 1 any changes to MSIV leakage rates.                        In fact, they 2 probably would have to be decreased.
3                MEMBER HALNON:            So, given the fact that 4 the numbers are higher or the release fractions are 5 higher, are the numbers reasonably correct?                        Or are 6 they just overly conservative, and you can identify 7 where that conservative point is that caused this to 8 happen, and can justify why it's not necessary?
9                MR. BROADBENT: So, the report, the Sandia 10 report that developed them, has very little detail 11 regarding how they were developed.                    It doesn't really 12 describe the model or anything that was used.
13                But what it does list, for example, the 14 events that went into developing them, and they're 15 significantly different for BWRs than for previously 16 applied in NUREG-1465, and it's different than even 17 what was applied for PWRs. If you look at that Sandia 18 report -- 2011-0128, I believe is the number -- most 19 of the core damage sequences are associated with 20 station blackout for BWRs.              And we've done a lot of 21 work in the last 20 years to reduce our station 22 blackout risk.      So, if that's scenario selection was 23 risk-informed, it's going the wrong way.
24                And    station        blackout          releases      occur 25 through the SRVs and through the pool.                        So, there NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
130 1 would be a significant amount of scrubbing associated 2 with those releases before the source terms would 3 become airborne. And it doesn't appear as though that 4 scrubbing effect was considered in the analyses that 5 were performed.
6                    So, we think those are the two impacts 7 that are driving the increase in BWR releases.                                We 8 don't think it's the increase of burnup itself.
9                    MEMBER HALNON:          Thank you.
10                    MEMBER    BROWN:            Can    I  ask  one      more 11 question?        I'm not much of an expert on the station 12 blackout.          I mean, I know what it is, but that's 13 assuming        you  have    no    power.          And  could  that      be 14 mitigated by plant modification which would decrease 15 the risk of having the station blackout?
16                    MEMBER HALNON:              No, all SBOs are not 17 created equal.
18                    MEMBER BROWN:          I figured that.
19                    MEMBER HALNON:          It depends on what feeds 20 into it, diesel generators configuration, and --
21                    MR. BROADBENT:          Yes, I think what these 22 scenarios that were assumed, you know, the NRC needs 23 to generate some core damage.                    And so, we have makeup 24 capability for station blackouts through our RCIC 25 reactor        cooling,    reactor        core      isolation    cooling NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
131 1 system, RCIC, R-C-I-C.              And I would assume that those 2 analyses probably assume that that trips at some point 3 in time and you get a late-term-type core uncovery, 4 and therefore, fuel failure.
5                  But  what      I  was      pointing      out  was      the 6 selection        of  core      damage      sequences        was    heavily 7 weighted towards station blackout.                        In fact, nearly 8 all      the  scenarios      that      went      into    developing        the 9 release fractions for boiling water reactors were 10 station-blackout-related sequences.
11                  MEMBER BROWN:          Okay.        Thank you.
12                  CHAIR PETTI: Okay. Let's turn now, then, 13 to public comments.
14                  Since we have received a request to speak, 15 we'll start with Brian.              Are you on the line?
16                  MR. MAGNUSON:          Yes.
17                  CHAIR PETTI:          Go ahead.
18                  Unmute yourself.              Unmute yourself.
19                  MR. MAGNUSON:              Hello.        This is Brian 20 Magnuson.        Can you hear me?
21                  CHAIR PETTI:          Yes.
22                  MR. MAGNUSON: Okay. As I stated, my name 23 is Brian Magnuson.            I'm a Lead Emergency Management 24 Specialist        at    Constellation,                formerly      Exelon 25 Generation.        I'm a former NRC-licensed Senior Reactor NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
132 1 Operator and Operation Shift Manager at Quad Cities 2 nuclear power plant.                I am acting expressly as a 3 member of the public.
4                  First of all, I appreciate the opportunity 5 and the invitation to attend this public meeting.
6                  I submit my prior public comments that are 7 posted        in ADAMS    and      understand          that  documented 8 responses from the NRC are now forthcoming.
9                  I also submit my petition for rulemaking 10 regarding Reg Guide 1.183, Rev. 0.                      It's PRM-50-122.
11                  In my prior public comments, I referenced 12 SAND-2008-6601, which determined that the BWR MSIV 13 source term methodologies provided in Reg Guide 1.183, 14 Rev. 0, are, quote, "non-conservative and conceptually 15 inaccurate."
16                  My prior comments expounded on this Sandia 17 document and identified other examples in which Reg 18 Guide 1.183 methodologies violate the laws of physics.
19                  It  appears        that      Draft      Guide  1389      may 20 correct some of the errors.                      However, if the NRC 21 allows Rev. 0 to coexist with Rev. 1, then the NRC 22 will still allow nuclear plants to ignore the laws of 23 physics in accident dose calculations.                      The practice 24 of Rev. 1 coexisting with Rev. 0 seems imprudent.
25                  I submitted documented pre-meeting public NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433          WASHINGTON, D.C. 20009-4309            www.nealrgross.com
 
133 1 comments to the NRC and the ACRS on March 15th, 2022.
2                My initial comments for this meeting focus 3 on the Reg Guide 1.183 and the Draft Guide 1389 4 assumption that primary containments will not fail in 5 a    maximum  credible      accident        or,    as  the  NRC      is 6 redefining, maximum hypothetical accident.
7                This    is      contrary        to    the  intent        of 8 TID-14844 and the purpose of assessing worst-case 9 credible accidents in source term dose calculations.
10                In some of the severe accident studies 11 referenced in today's presentation, the NRC clearly 12 recognizes that containments will fail in credible 13 nuclear accidents. Many of us watched the containment 14 at Fukushima catastrophically fail.
15                Therefore, the use of lesser accidents to 16 determine the regulatory capability of nuclear power 17 plants to prevent or mitigate radiological releases 18 seems imprudent and illogical.
19                There was just some discussion on station 20 blackout or long-term station blackout accidents. And 21 I'll add to that the floods capabilities and the use 22 of hardened containment events that were required by 23 the NRC in Order EA-13109; that those source terms are 24 considerably greater than the DBA LOCA used in the 25 current Reg Guide and the proposed Draft Guide.
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
134 1                    Additionally, I believe the NRC's use of 2 risk-informed insights and regulations conflict with 3 the      regulatory    premise        of      the    maximum  credible 4 accident source term that's used in the design basis 5 of each nuclear power plant.
6                    It appears the NRC's risk-informed insight 7 regulations are sometimes used to allow nuclear powers 8 to      evade    deterministic          regulations,        particularly 9 General Criterion 19, because the NRC recognizes, and 10 I quote, "For certain plants, the margin for meeting 11 acceptable dose limits in the control room or certain 12 DBAs        is  very    small."              The      concern    is      the 13 implementation of the new ICRP 103 values because 14 these        would  exceed      regulatory          dose  limits.            A 15 subsequent        evaluation        confirmed          this  concern        was 16 accurate.        Therefore, the NRC rejected ICRP 103.
17                    This motivation epitomizes my concerns 18 with Reg Guide 1.183 and Draft Guide 1389.                              I am 19 concerned that the NRC's efforts are overly focused on 20 mitigating known nuclear power plant design basis 21 efficiencies,          providing            additional,        sometimes 22 questionable, allowance to ensure nuclear power plants 23 can comply with deterministic regulations, circumvent 24 deterministic regulations, while bringing them into 25 compliance. This does not comport with the NRC's duty NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433            WASHINGTON, D.C. 20009-4309          www.nealrgross.com
 
135 1 to protect people and the environment.
2                That's the end of my comments for now, 3 unless anybody has any questions.
4                CHAIR PETTI:        Thank you for your comment.
5                Anybody else?          Identify yourself, unmute 6 yourself, and make your comment.
7                (No response.)
8                Okay. Not hearing anything, I think we 9 are at the end of our meeting.                  So, I would like to 10 adjourn our Subcommittee meeting.
11                Thank you all.
12                (Whereupon, at 11:56 a.m., the meeting was 13 adjourned.)
14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433        WASHINGTON, D.C. 20009-4309        www.nealrgross.com
 
Revision of Regulatory Guide 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors ACRS Subcommittee Meeting Mark Blumberg, Senior Reactor Engineer, NRR/DRA (Technical Lead)
Elijah Dickson, Reactor Scientist, NRR/DRA Steve Jones, Senior Safety and Plant Systems Engineer, NRR/DSS Shilp Vasavada, Senior Reliability and Risk Analyst, NRR/DRA John Parillo, Senior Reactor Engineer, NRR/DRA Paul Clifford, Senior Technical Advisor, NRR/DSS Sean Meighan, Reactor Scientist, NRR/DRA (Project Lead)
March 16, 2022
 
Agenda
* Key Messages
* Background
* RG 1.183 Updates Proposed in DG-1389
* Modified SAND 2008-6601 Method
* Re-evaluation of AEB-98-03 with Multi-group/Numerical Integration Method
* Technical Assessment on Seismic Analysis for Alternative Drain Pathway
* ATF, HBU, Increased Enrichment (LOCA)
* Steady-State and Transient Release Fractions for Non-LOCA Analyses
* Revised Fuel Handling Accident (FHA)
* Draft Revisions to Current DG-1389
* Key Takeaways
* Looking Forward
* Discussion/Feedback 2
 
NRC Management and Staff Coordination Steering Committee Mike Franovich, NRR/DRA        NRR/DSS:
Joseph Donoghue, NRR/DSS      Paul Clifford Mohamed Shams, NRR/DANU        Steve Jones Louise Lund, RES/DE NRR/DANU:
Working Group                  Michelle Hart NRR/DRA:
Mark Blumberg (technical lead)
Sean Meighan (project lead)    RES/DE:
John Parillo                  Michael Eudy Richard Clement Jerry Dozier                  OGC:
Elijah Dickson                Mary Frances Woods Sean Meighan Shilp Vasavada 3
 
Key Messages
* The NRC staff is revising RG 1.183, Rev 0 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.
* RG 1.183 Rev. 0 and Rev. 1 will co-exist as a result of SRM-SECY-18-0049, Management Directive and Handbook 8.4, Management of Backfitting, Issue Finality, and Information Collection.
* The NRC staff has elected not to finalize a draft revision to RG 1.183, known as DG-1199, and is issuing DG-1389 as a replacement. DG-1389 considers and addresses technical issues and public comments related to the issuance of DG-1199.
* NRC staff made efforts for early outreach to external stakeholders by holding 3 public meetings focused on the revision of RG 1.183.
* November 19, 2020 (Slides ML20296A425)
* March 5, 2021 (Slides ML21056A058)
* May 13, 2021 (Slides ML21124A121)
* RG 1.183 Rev. 1 provides an acceptable method of modeling BWR MSIV leakage. Use of individual regulatory positions in isolation will need additional justification.
4
 
Key Messages (Contd)
* The objectives of the revision are to:
  - make the guidance more useful by considering feedback and comments from licensees;
  - incorporate relevant operating experience as well as recent post-Fukushima seismic risk insights and walkdowns;
  - respond to change of regulatory environment (e.g., backfit guidance SRM-SECY-18-0049 & NuScale SRM-SECY-19-0036);
  - incorporate lessons learned from recent NRC staff reviews of Alternative Source Term (AST) and BWR Main Steam Line Isolation Valve (MSIV) leakage LARs;
  - ensure sufficient guidance is in place for licensing advanced light-water reactors (LWRs), accident tolerant fuel (ATF), high-burnup, and increased enrichment fuel; and,
  - incorporate insights from new research activities.
5
 
===
Background===
6
 
===Background===
* Origin: Footnote to 10 CFR 100.11(a) is a performance-based rule to evaluate the defense-in-depth provided by the containment.
  - TID-14844 Source term provided guidance which assumed the source term is instantaneously available in the containment.
* Radionuclide behavior observed during the TMI accident did not appear at all similar to the TID-14844 source term.
  - NRC initiated research efforts in the area of severe accidents which culminate in publication of NUREG-1150.
  - NUREG-1465 source term was derived from the sequences in NUREG-1150.
* RG 1.183 Rev. 0 adopted the NUREG-1465 early in-vessel fuel melt source term.
7
 
Background (Contd)
* NRC staff developed RG 1.183 Rev. 0 (July 2000) to support implementation of 10 CFR 50.67, Accident Source Term.
  - Applicable to nuclear power reactor applicants and licensees adopting 10 CFR 50.67.
* Limited range of applicability on Non-LOCA release fractions.
  - Identified the significant attributes of an acceptable accident AST based on NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (1995).
  - Provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST.
8
 
DG-1199
* In October 2009, the NRC issued for public comment DG-1199 as a proposed Rev. 1 of RG 1.183.
* Staff received 150 public comments
* The reasons for revision of RG 1.183 in DG-1199 were:
  - Providing additional guidance for modeling BWR MSIV leakage,
  - Expand applicability of Non-LOCA release fractions to support modern fuel utilization,
  - Extending the applicability of the proposed RG for use in satisfying the radiological dose analysis requirements contained in 10 CFR Part 52 for advanced LWR design and siting,
  - Providing additional meteorological assumption guidance.
9
 
Modern Fuel Utilization
* Since DG-1199 was issued for public comment, NRC issued several license amendments to support modern fuel utilization.
  - Oconee Units 1, 2, and 3 (2019)
  - Shearon Harris (2018)
  - H.B. Robinson (2017)
  - Catawba Units 1 and 2, McGuire Units 1 and 2, Oconee Units 1, 2, and 3 (2016)
  - Diablo Canyon Units 1 and 2 (2015)
* Reinforced need for expanded Non-LOCA release fractions 10
 
2019 License Amendment Requests
* In 2019, NRC received several AST LARs requesting increased BWR MSIV leakage
* As a result, work on DG-1199 was postponed to allow NRC staff to incorporate lessons learned, from evaluation of the LARs, into the revised RG 1.183:
  - James A. FitzPatrick Amendment No. 338 for AST, July 21, 2020 (ML20140A070)
  - Quad Cities Nuclear Power Station, Units 1 & 2 - Amendment Nos. 281 and 277 to increase allowable MSIV leakage, June 26, 2020 (ML20150A328)
  - Nine Mile Point Nuclear Station, Unit 2 - Amendment No. 182 to change allowable MSIV leak rates, October 20, 2020 (ML20241A190)
  - Dresden Nuclear Power Station, Units 2 & 3 - Amendments Nos. 272 and 265 to increase allowable MSIV leakage, October 23, 2020 (ML20265A240) 11
 
Changes from Disposition of Recent Differing Professional Opinion (DPO-2020-002) Panel Report1
* A significant change in how radiological consequences are performed.
* Crediting Safety Related Systems with a deterministic fuel melt source term.
* Reinstatement of the Maximum Hypothetical Accident (MHA).
* Removing mechanistic explanations for the deterministic source term.
* The LOCA will be defined as an event resulting in the loss of the ability to cool the core.
* The MHA LOCA will be defined as an unspecified event resulting in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.
* Appendix A will describe acceptable assumptions for the evaluation of the MHA LOCA.
* Elimination of the assumption of a 2-hour delay in the distribution of the fuel melt source term into the containment and the elimination of the R-Factors in the Sandia Method.
1 For more information see the DPO Case File for DPO-2020-002 (ML21067A645).
12
 
RG 1.183 Updates Proposed in DG-1389 13
 
General Updates
* Staff plans to include changes proposed in DG-1199 as modified by public comments.
  - Incorporate new or remove withdrawn regulatory guidance (e.g., RG 1.194 atmospheric dispersion modeling)
  - Guidance for modern fuel utilization (non-LOCA gap fractions).
  - Changes due to Regulatory Information Summaries (i.e., 06-04, 01-19).
  - Lessons learned from license reviews (i.e., clarify DFs and containment isolation as used in the FHA).
  - Clarify TEDE calculation terminology (i.e., EDEX vs. EDE).
  - Remove environmental qualification guidance from RG and refer to RG 1.89.
14
 
BWR Main Steam Line Aerosol Deposition Models
* RG 1.183 Rev. 0 does not provide an aerosol main steam line deposition model for the LOCA main steam line leakage pathway.
* DG-1389 provides guidance for the three deposition methods as acceptable in regulatory positions to be used in conjunction with revised seismic analysis.
* The first model utilizes select recommendations from the report, SAND 2008-6601, Analysis of Main Steam Isolation Valve Leakage in Design Basis Accident Using MELCOR 1.8.6 and RADTRAD.
* Two additional models were developed as the staff addressed issues identified in RIS 2006-04, Experience with Implementation of Alternative Source Terms.
  - Reconstitution of the AEB-98-03 settling velocity modeling parameters and developed the: (1) multigroup method, and (2) numerical integration method, to address changing settling velocity distributions in downstream piping sections.
15
 
BWR Main Steam Line Aerosol Deposition Models (Contd)
* Clarifications for BWR MSIV leakage pathway
  - All steam lines are intact, assuming no break.
  - Delete regulatory position crediting containment sprays when crediting aerosol removal within the main steam lines.
  - No credit for holdup or aerosol deposition within the inboard main steam line.
16
 
Modified SAND 2008-6601 Method 17
 
Modified SAND 2008-6601 Method
* A MELCOR uncertainty analysis of the main steam line model was developed to calculate the aerosol deposition in the steam line and condenser.
* MELCOR is a well vetted, benchmarked state-of-the-art computer code that holistically models the pertinent phenomena for the LOCA.
* The MELCOR models considered the range of physical parameters (steam line configurations and designs) and aerosol physics parameters.
* The model considers following deposition mechanisms: gravitational aerosol settling, diffusiophoresis, thermophoresis, Brownian motion.
* The model also considers the effects of agglomeration, condensation and hygroscopicity.
* The uncertainty analysis used is very similar to those currently endorsed in RG 1.183 for modeling containment aerosol removal by natural processes and by sprays (See Regulatory Positions A-3.2 and A-3.3).
* The Modified SAND 2008-6601 method eliminates the use of steam dome concentrations as the source of BWR MSIV leakage for the first 2 hours of the accident evaluation.
18
 
Re-Evaluation of AEB 98-03 with Multi-group/Numerical Integration Method 19
 
Re-Evaluation of AEB 98-03 with Multi-group/Numerical Integration Method
* AEB 98-03 is utilized by many licensees in conjunction with RIS 2006-04, Experience with Implementation of Alternative Source Term to address staff concerns.
: 1. Applied more conservative AEB-98-03 settling velocities.
: 2. Developed the 20-Group Method to account for the entire settling velocity distribution which the NRC has never endorsed.
* Re-evaluate AEB 98-03 in coordination with the Office of Research.1
        - Simplified AEB-98-03s model to just one known distribution (e.g., aerodynamic mass median diameter (AMMD)).
        - European State of the Art Report on Nuclear Aerosols (SOAR) .2
* Assessed methods for integrating the settling velocity distribution. 1
        - (1) Multi-group (Monte Carlo Integration) and (2) Integration by parts.
1 - See ADAMS Accession No. ML21141A006. Note that while referenced in the draft RG, this document does not establish regulatory positions. For example, input parameters such as the AMMD, assumed in example calculations and statements regarding the validity of the existing 20-group method are not endorsed in the RG.
2 - Allelein, et. al., State-of-the-Art Report on Nuclear Aerosols, NEA/CSNI/R(2009)5 20
 
Technical Assessment on Seismic Analysis for Alternate Drain Pathway 21
 
What is the Alternate Pathway in RG 1.183?
* Pathway for BWR MSIV leakage to main condenser via drain lines
* Holdup in main condenser -> fission product decay -> lower dose
  - RG 1.183 defines release rate from condenser (not leak tight)
* Credit for pathway results in quantitative change in calculated dose
  - Licensees request credit with docketed supporting information 22
 
Information for Alternate Pathway Credit Alternate Pathway Credit Information on Pathway Identification, Availability, and Use                            Information on Seismic (No changes proposed compared to RG                  Capacity of Alternate Pathway 1.183, Rev. 0)
Proposed RG 1.183, Rev. 1:
RG 1.183, Rev. 0:        Leveraging Staffs Technical SE for BWROG Topical              Assessment (ML20085J042) 23
 
BWROG Topical Report Approach in Rev. 0
* BWROG NEDC-31858P (1993)
  - Non-nuclear seismic experience information
  - Generic seismic hazard curves
* NRC staffs safety evaluation issued in 1999 (ML010640286)
  - Included 9 limitations and conditions on use including:
* Applicability of generic seismic hazard curves
* Walkdowns of main steam piping by independent contractors 24
 
Proposed Approach in Rev. 1
* Based on NRC staffs technical assessment (ML20085J042)
* Presented to ACRS Subcommittees on July 23, 2021 (ML21223A034)
  - Based on operational insights, seismic capacity data, and risk insights over past 20 years
  - Identifies lower bound median fragility to encompass potential failure modes of alternate pathway SSCs
* FRN will solicit public comments on technical assessment concurrent with DG
* Docketed information to justify that plant is within the bounds of the technical assessment 25
 
Proposed Approach in Rev. 1 (Contd)
* Available dynamic analyses against plant-specific safe shutdown earthquake (deterministic)
OR
* Risk-informed evaluation
  - Margin in code of record of alternate pathway SSCs
  - Sample confirmatory calculations for flexibility at major branch connections
  - Walkdowns sample alternate pathway SSCs
  - Gradation based on latest site-specific ground motion response spectrum
* NRC staff considers proposed approach to be streamlined
* Credit for pathway impacts licensees analysis of record (AOR)
  - Licensee requests quantitative credit and includes it in its dose assessment
  - Draft Interim Staff Guidance (DRA-ISG-2021-01, ML21278A372) does not change the proposed AOR
  - NRC staff needs high confidence for accepting inputs that impact the AOR 26
 
Consideration of BWR MSIV Leakage Values
* RG 1.183 Rev. 1 provides 3 different BWR MSIV leakage models which can be used in conjunction with a streamlined strategy to credit a pathway to the condenser.
* NRC has approved BWR MSIV leakage of 200 scfh or below per MSIV with a total MSIV leakage of 400 scfh or below. Higher values will be considered on a case-by-case basis with sufficient justification.
* Maintaining BWR MSIV leakage at or below certain values is based on the following considerations:
  - Leakage in excess of 200 scfh per MSIV could be indicative of substantial valve defects
  - These values represent maximum values in existing fleet
  - 400 scfh is on the order of total containment leakage
  - Comparison to original design value of 11.5 scfh per MSIV 27
 
ATF, HBU, Increased Enrichment (LOCA) 28
 
ATF, HBU, Increased Enrichment (LOCA)
* Approach: for MHA-LOCA, utilize readily available information to support near-term licensing activities.
* Updating RG 1.183 Rev. 0 Tables 1, 2, and 4 (MHA-LOCA) with hybridized accident source term tables from SAND 2011-0128, Accident Source Terms for Light Water Nuclear Power Plants Using High-Burnup of MOX Fuel.
        - Expanded to encompass near-term ATF design concepts1 fuel burnup extension to 59 GWD/MTU max assembly-averaged discharge burnup (~68 GWd/MTU peak rod-average) and 235U enrichments up to 8.0 wt%.
        - Considered impact of FFRD for the Appendix A assumptions2.
        - Provide conditions and limitations of the report applicability for regulatory purposes.
* On going research efforts are underway to update the SAND 2011-0128 accident source term to accommodate higher burnup and increased enrichments for LOCA releases. However, completion of the updated analyses may not be finished before the update to the regulatory guide.
1- NRC Memorandum, Applicability of Source Term for Accident Tolerant Fuel, High Burn Up and Extended Enrichment, dated May 13, 2020, ADAMS Accession Number ML20126G376 2- NRC Memorandum, Letter Report on Evaluation of the Impact of Fuel Fragmentation, Relocation, and Dispersal for the Radiological Design Basis Accidents in Regulatory Guide 1.183 (ADAMS Accession No. ML21197A067) 29
 
Steady-State and Transient Release Fractions for Non-LOCA Analyses 30
 
Updates to Guidance
* Expand applicability of guidance o Modern fuel utilization o High burnup (68 GWd/MTU rod average burnup)
* Separate BWR and PWR steady-state releases
* Burnup-dependent transient fission gas release (FGR) correlations for prompt power increase accidents
* Analytical procedure for calculating revised steady-state release fractions 31
 
Expanded Range of Applicability Rev.0 Range of Applicability 32
 
FAST Steady-State Release Fractions
* FAST-1.0.1 along with ANS-5.4 (2011) standard release model used to calculate bounding steady-state release fractions o Results sensitive to fuel rod design and operating history o  BWR 10x10 and PWR 14x14 designs o Competing effects:
o  Expanded power operating domain promotes higher releases o  Application of model uncertainties promote higher releases o  ANS-5.4 (2011) standard promotes lower I-131 releases relative to ANS-5.4 (1982) 33
 
BWR Steady-State Release Fractions 34
 
PWR Steady-State Release Fractions 35
 
Alternate Treatment for Cesium Release
* ANS-5.4 standard release model recommends a diffusion coefficient for cesium equivalent to twice that of Nobles (Release Fraction)Cs-134, Cs-137 = (Release Fraction)Kr-85 * (2.0)0.5
* Key observations from various research programs:
o Notable differences in physical properties and chemical effects o Cesium reacts with other constituents to form relatively immobile compounds which accumulate along grain boundary and rim structure o Below 1200°C, cesium is a liquid and forms a film along grain o When released, cesium may react with the zirconium cladding to form more stable (i.e., non-gaseous) compounds o Nobles are chemically inert gas which readily diffuse to grain boundary, form interconnected bubbles, and then released 36
 
Alternate Treatment for Cesium Release (Contd) o Ratio of measured noble gas and cesium releases varies between 2.2/1 to 10/1 from the different studies
* Based on these key observations, a conservative lower bound of 2/1 was selected (Release Fraction)Cs-134, Cs-137 = (Release Fraction)Kr-85 * (0.5) 37
 
Transient Fission Gas Release
* Fission gas released during prompt power excursion testing on irradiated fuel rod segments has been reported 38
 
Transient Fission Gas Release (Contd)
* TFGR is an additional component of radiological source term and should be considered in dose consequences
* TFGR is sensitive to local fuel burnup and deposited energy
* Correlation adjustment to account for differences in diffusion coefficients and half-life (i.e., 1/3 for short-lived nuclides) 39
 
Analytical Procedure
* An acceptable analytical procedure provided for calculating plant-specific steady-state release fractions o Significant reductions in steady-state release fractions achievable using plant-specific fuel rod designs and power operating domains o Fuel vendors may use their approved fuel rod thermal-mechanical codes
* Provides flexibility and dose margin to accommodate more aggressive fuel utilization, HBU, etc
* Reduces regulatory uncertainty 40
 
HBU, IE, and ATF Applications (Non-LOCA DBAs)
* DG-1389 provides revised steady-state release fractions applicable to an expanded operating domain o  Supports HBU/IE applications o  Relaxation to both short-term (e.g., I-131) and long-term (e.g., Cs-137) radionuclide releases achieved through more rigorous engineering
* DG-1389 provides an acceptable analytical procedure for calculating application-specific release fractions o  Supports HBU/IE and ATF applications o  Relaxation and flexibility may be achieved
* DG-1389 TFGR correlations + RG 1.236 coolability limits provide a means to address FFRD for postulated CRE/CRD accidents o  Supports HBU/IE applications 41
 
Revised Fuel Handling Accident (FHA) 42
 
Revised Fuel Handling Accident
* FHA can require significant staff resources, and in some cases has become a limiting accident, despite its low safety and risk significance.
* Revisited the original studies from the early 1970s forming the technical basis for the FHA and ACRS recommendations1.
* Model improvements established from the current understanding of reactor fuel pin physics and iodine chemistry under the environmental conditions in which fuel handling operations are taking place.2
    - Two-phase model:
: 1. Initial gas gap release - Elemental, I2, and Organic Iodine, CH3.
: 2. Re-evolution - Assumes CsI dissociates into the pool water and re-evolves as I2 over the accident duration.
* Concluded considerable margin exists regarding the scrubbing effects of iodine in the spent fuel or reactor pool and that the current staff DBA FHA fission product transport model can be refined while still maintaining conservatism.
1 - 473rd ACRS Full-committee meeting, held June 7, 2000, the committee members expressed concern with the staffs treatment of iodine speciation and modeling the release as a puff release. 43 2 - Memo from RES to NRR, Closeout to Research Assistance Request for Independent Review of Regulatory and Technical Basis for Revising the Design-basis Accident Fuel Handling Accident, November 23, 2019 (ML19270E335)
 
Draft Revisions to Current DG-1389
* 3.2      Release Fractions
        - Applicability of Source Term for Accident Tolerant Fuel, High-Burnup and Extended Enrichment, (Ref. 25) in part, documents the applicability of SAND 2011-0128 for burnups up to 68 gigawatt-days per metric ton of uranium (GWd/MTU) peak rod average and enrichments up to 8 percent for certain near-term ATF designs which include chromium-coated cladding and chromia-doped fuel, but do not include Iron-Chromium-Aluminum alloy cladding.
* 5.3      Atmospheric Dispersion Modeling and Meteorology Assumptions
        - RG 1.145 and RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, issued June 2003 (Ref. 38), should be used if the FSAR /Q values are to be revised or if values are to be determined for new release points or receptor distances or release characteristics that affect plume rise. In addition to calculating control room /Q values, the modeling methodology outlined in RG 1.194 may be modified to estimate offsite
            /Q values at offsite boundaries out to distances of 1,200 m as long as the procedures are consistent with the Regulatory Positions in Part C of RG 1.145.
Note: These draft revisions have been prepared and are being released to support ongoing public discussions. These revisions have not been subject to NRC legal reviews and approvals, and its contents are subject to change and should not be interpreted as official agency positions.
44
 
Key Takeaways
* DG-1389 provides guidance for three acceptable main steam line aerosol deposition models.
* DG-1389 provides guidance to support a streamlined approach for licensees to credit an alternate pathway for BWR MSIV leakage to the main condenser.
* The staff are prepared to review near-term ATF designs and burnups up to 68 GWd/MTU peak rod-average and U-235 enrichments up to 8.0 wt%. Future updates to the guide are expected to accommodate higher burnups and enrichments.
* The range of applicability has been expanded to support modern fuel utilization. The impact of this expanded operating domain on radiological source terms has been minimized by applying more rigorous engineering judgement.
* An acceptable analytical procedure for predicting plant-specific radionuclide release fractions has been included and provides flexibility and margin recovery.
45
 
Looking Forward
* Issue DG-1389 for 60-day public comment - April 2022
* Staff review and disposition of public comments
* Update draft RG 1.183 Rev. 1, as necessary
* OGC review of final draft
* ACRS Full Committee Meeting
* Issuance of RG 1.183 Rev. 1 - Nov/Dec 2022 46
 
Discussion/Feedback 47
 
Acronyms ACRS - Advisory Committee on Reactor    external exposures)                        MSL - main steam line Safeguards                              EOL - end of life                          NRR - Office of Nuclear Reactor Regulation AEB - accident evaluation report        FAST - Fuel Analysis under Steady-State and OGC - Office of General Counsel AMMD - aerodynamic mass median          Transients                                  PNNL - Pacific Northwest National diameter                                FFRD - fuel fragmentation, relocation, and Laboratory AOO - anticipated operational occurrence dispersal                                  PWR - pressurized water reactor AOR - analysis of record                FGR - fission gas release                  RADTRAD - RADionuclide, Transport, ATF - accident tolerant fuel            FHA - fuel handling accident                Removal, and Dose Estimation AST - alternative source term            FRN - Federal Register Notice              RCS - reactor coolant system BWR - boiling water reactor              FSAR - final safety analysis report        RES - Office of Research BWROG - BWR Owners Group                GMRS - ground motion response system        RG - regulatory guide CFR - Code of Federal Regulations        GSD - geometric standard deviation          RIS - Regulatory Issue Summary Cs - cesium                              GWD/MTU - gigawatt day per metric ton Scfh - standard cubic feet per hour DANU - Division of Advanced Reactors and uranium                                    SE - safety evaluation Non-power Production and Utilization    HBU - high burnup                          SSC - systems, structures, and components Facilities                              IE - Increased Enrichment                  TEDE - total effective dose equivalent DBA - design basis accident              Kr - krypton                                TFGR - transient fission gas release DE - Division of Engineering            KW/ft - kilowatt per foot                  TID - technical information document DF - decontamination factor              LAR - license amendment request            TMI - Three Mile Island DG - draft guide                        LHGR - linear heat generation rate DPO - Differing Professional Opinion    LOCA - loss of coolant accident DRA - Division of Risk Assessment        LWR - light water reactor DSS - Division of Safety Systems        MHA - maximum hypothetical accident EDE - effective dose equivalent          MOX - mixed oxide EDEX - effective dose equivalent (for    MSIV - main steam isolation valve 48
 
Backup Slides for Modified SAND 2008-6601 Method 49
 
MELCOR Analysis of Full Plant DBA 51
 
SAND 2008-6601 Method (contd)
Aerosol size distribution evolves in time, depending on sources, agglomeration and removal processes
* MAEROS sectional model of Gelbard section mass
* Particles grow in size
                                                              - Agglomeration Time 1                - Water condensation
* Particle fallout by gravitational settling
* Particle deposition processes 1
                                                              - Thermophoresis 0.1                        10        100 aerosol diameter [µm]                  - Diffusiophoresis
                                                              - Brownian motion
* Cs chemisorption in RCS modeled
                                                              - Iodine from CsI re-volatilizes when section mass reheated Time 2 Validation Experiments PANDA, DEHBI, CVTR, FALCON, 0.1        1              10        100 LACE-LA4, ABCOVE, Wisconsin flat aerosol diameter [µm]                                      plate 52
 
SAND 2008-6601 Method (contd)
Pre-Recorded MELCOR Conditions pressure temperature gas composition aerosol mass and size distribution steam        inboard inboard intermediate outboard outboard dome          steam  MSIV      steam    MSIV    steam  condenser volume Line              Line              Line
* Reduced nodalization
* Boundary conditions from full-plant MELCOR analysis
* Monte Carlo sampling used to derive distributions 53
 
Illustrative Example of Results from Deposition Analysis inboard                    between valves outboard                    condenser 54
 
SAND 2008-6601 Results MSL and Condenser Removal Coefficients MSL section              0 - 10 (hr) 10+ (hr) in-board                            0*      0*
between MSIVs                      1.8      1.0 out-board                          1.0      0.7 condenser                          0.015    0.012 Note. removal coefficients are given in 1/hr
* Not credited because of thermal bi-directional flow, turbulence, containment boundary
* Based upon Recommendation 6.4 of SAND 2008-6601 for post-reflood conditions (taken from Table 6-1)
* Lambdas decrease as remaining aerosols becomes smaller 55
 
Backup Slides for Steady-State and Transient Release Fractions for Non-LOCA Analyses 56
 
Challenges to 75 GWd/MTU
* Approved fuel rod thermal-mechanical performance model validated up to 75 GWd/MTU o Quantified model uncertainties
* Defined fuel rod power operating domain o PWR and BWR fuel management scoping studies needed to define realistic fuel rod power history o Predicted released sensitive to power history 57
 
Impact of Burnup Extension AOO Fuel Temperature -> Fuel Melting EOL RIP -> Clad Liftoff Short-lived Volatile (I-131) = 0.07 Long-lived Noble (Kr-85) = 0.40 Short-lived Volatile (I-131) > 0.10 Long-lived Noble (Kr-85) > 0.50 58
 
Extent of Empirical Database
* PNNL completed a critical assessment of the various models within the FAST fuel performance code.
o Most models validated up to 75 GWd/MTU
* ANS-5.4 Release Model limited to 70 GWd/MTU
* TFGR database limited to
  ~68 GWd/MTU 59
 
Use by Non-LWRs
* Guidance which is not LWR-specific could provide useful information on radiological consequence analyses and development of source terms
  - Section 2: Attributes of an Acceptable Accident Source Term
  - Section 4: Dose Calculation Methodology
  - Section 5: General Considerations 60
 
ACRS Meeting on Proposed Draft RG 1.183 Frankie Pimentel, Sr. Project Manager - Engineering & Risk, NEI Greg Broadbent, Senior Staff Engineer - Corporate Nuclear Analysis, Entergy March 16, 2022
  ©2022 Nuclear Energy Institute
 
Initial Concerns from Review of Proposed Draft RG 1.183 The overall impact on sample BWR/PWR plants should be evaluated prior to issuing
* Industry strongly recommends that NRC staff provide information indicating that aspects of the updated guidance, such as release fractions and timing, have been considered for impacts to resulting analyzed doses
* Proposed changes are as significant as when the guidance transitioned from the TID Source Term to the Alternative Source Term
* As proposed, the conservatisms and changes incorporated in this revision precludes its use by many plants that are interested in implementing near-term ATF design concepts, fuel burnup extension to 68 GWd/MTU (peak rod average), and 235U enrichments up to 8.0 wt%
                                                                              ©2022 Nuclear Energy Institute 2
 
Initial Concerns from Review of Proposed Draft RG 1.183 Majority of BWR plants may not be able to adopt Rev. 1 without significant licensing basis changes or plant modifications
* The new BWR/PWR core inventory release fractions and timing may have considerable impact on the margin to the Control Room exposure criteria BWR iodine release fraction nearly doubles for LOCA/MHA BWR tellurium (parent of iodine) release fraction increases significantly Basis of the changes in the BWR release fractions is not understood Core damage scenarios heavily weighted by SBO Suppression pool scrubbing apparently not credited
* New BWR main steam line model applies more conservative assumptions than previous models
* Lowered CMT and MSIV leakage rates
* Higher occupational doses to support lower leakage requirements
* Reduced control room in-leakage allowances
                                                                              ©2022 Nuclear Energy Institute 3
 
Initial Concerns from Review of Proposed Draft RG 1.183 Option to maintain compliance to Rev. 0
* Clear validation regarding continued acceptance of Rev. 0
* Any additional justification needed non-LOCA gap fractions MSL deposition model (Rev. 0 is silent on acceptable models)
Option for selective implementation of Rev. 1
* Describe acceptable paths for selective implementation
* Acceptable on an accident basis (e.g., fuel handling accident only)
                                                                              ©2022 Nuclear Energy Institute 4
 
Feedback from Review of Proposed Draft RG 1.183 Revised Fuel Handling Accident
* Industry recognizes the new FHA model developed by the Staff Based on thorough review of decades of historical information and current understanding of iodine chemistry Internally motivated - not industry initiated
* Resulting model is a more accurate characterization of the low consequences of this accident
                                                                      ©2022 Nuclear Energy Institute 5
 
Other Communication Regarding DPO Pertaining to MSIV Leakage
* No updates have been provided on the status of the DPO discussed during the November 2021 ACRS meeting on ISG-2021-01
* Will the resolution of the DPO affect the changes regarding the Revised Seismic Analysis of the Alternative Drain Pathway introduced in the proposed revision to RG 1.183
* Will the resolution of the DPO affect the schedule for issuing the revision to RG 1.183
                                                                        ©2022 Nuclear Energy Institute 6
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments RE: ACRS Subcommittee Meeting on Revision of Regulatory Guide 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors March 15, 2022 In preparation for tomorrows ACRS Subcommittee Meeting, I submit the following pre-meeting public comments:
While I appreciate the invitation and opportunity to attend another public meeting, I do not believe the NRC has been responsive to my prior RG 1.183/DG-1199 public comments. The NRC posted my public comments (and questions) in ADAMS (ML20343A064 and ML20351A321), but has not provided documented responses. In effect, this negates my efforts and the intent of NRC public meetings.
As recently informed, I understand that documented responses are now forthcoming.
My initial comments on DRAFT REGULATORY GUIDE DG-1389, Proposed Revision 1 to Regulatory Guide 1.183 are posted (bulleted) below relevant sections.
Sincerely, Brian D. Magnuson magnuson28@msn.com 1020 Station Blvd. #212 Aurora, IL 60504 Lead Emergency Management SpecialistConstellation (formerly Exelon Generation)
Former NRC Licensed Senior Reactor Operator/Operations Shift Manager Acting expressly as a member of the public.
DRAFT REGULATORY GUIDE DG-1389, Proposed Revision 1 to Regulatory Guide 1.183 B. Discussion This revision of the guide (Revision 1) addresses new issues identified since the guide was originally issued. These include (1) using the term maximum hypothetical accident (MHA) 1 loss-of-coolant accident (LOCA) to define the accident described in the applicable regulations in Section A above . . .
Footnote 1: The maximum hypothetical accident (MHA) (also referred to as the maximum credible accident) is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible. As used in this guide, the term LOCA refers to any accident that causes a loss of core cooling. The MHA LOCA refers to a loss of core cooling resulting in substantial meltdown of the core with subsequent release into containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite hazards evaluated based on design basis containment leakage.
1
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments
* The term maximum hypothetical accident (MHA) appears misleading. Because Regulatory Guides are not regulations, it seems inappropriate to use an NRC Regulatory Guide to bound applicable regulations to the DG-1389 definition of the MHA. Otherwise stated, the NRC does not have the authority to redefine the applicability of federal nuclear safety regulations (or the current licensing basis (CLB) of nuclear power plants) by using Regulatory Guide 1.183. For example, the requirements of 10 CFR 50, Appendix A, General Design Criterion19 are not limited to a LOCA or the MHA described in DG-1389.
* The term maximum credible accident (MCA) was developed by the Atomic Energy Commission (TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites) to quantify the threat of nuclear power plants in terms of the worst-case possible accident release of radiological material to the environment. In theory, by quantifying the worse-case radiological release, nuclear power plants would be built safe distances from populations. Subsequently, plant designers used the maximum credible accident to design safety systems, such that they would prevent or mitigate the quantified threat, thereby, ensuring compliance with source term regulations. The AEC authors of TID-14844 developed, what they thought was, a conservative example to describe the maximum credible accident, but they intentionally did not define it, stating:
o  Consideration of these as well as other aspects of [radiological] hazards evaluation involves so many different situations and such complex technological problems that it would be quite impossible to anticipate and answer all questions that will arise.
o  Designers of reactors are expected to examine all significant aspects of the hazards and safety problem they believe are appropriate to the particular situation with which they are dealing. In any case, the designer and/or, applicant bears the responsibility for justifying all the assumptions and methods of calculation used in a [radiological] hazards evaluation. The fact that aspects of the problem are not considered in the example set forth here, does not in any way relieve the designer and/or applicant of the responsibility for carefully examining, in his particular case, every significant facet of the hazards and safety problem.
* As stated in RG 1.183 Revision 0, Since the publication of TID-14844 [1962], significant advances have been made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents. These advances were made as the result of severe accident studies conducted by or for the NRC.
Conclusions from studies include the following:
o  As in the DBA-LOCA class, the doses from "melt-through" releases (involving thousands of curies) generally would not exceed even the most restrictive PAG beyond about 10 miles from a power plant. The upper range of the core-melt accidents is categorized by those in which the containment catastrophically fails 2
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments and releases large quantities of radioactive materials directly to the atmosphere because of over-pressurization or a steam explosion. These accidents have the potential to release very large quantities (hundreds of millions of curies) of radioactive materials. There is a full spectrum of releases between the lower and upper range with all of these releases involving some combination of atmospheric and melt-through accidents. These very severe accidents have the potential for causing serious injuries and deaths. [(1978) NUREG-0396 (EPA 520/1-78-016)]
o The accident at TMI demonstrated the reality of the risk, previously only theoretically assessed, of accidents that result in substantial degradation and melting of the core. This risk arises from the fact that core-degradation accidents can lead to containment failure and the eventual release of large amounts of radioactivity to the environment. [(1980) NUREG-0660 (Vol. 1), NRC Action Plan Developed as a Result of the TMI-2 Accident]
o In examining the source terms . . . one must assess the probability of escape from the containment under routine use conditions or in any postulated accident situation. [(1983) NUREG/CR-3332, A Textbook on Environmental Dose Analysis]
o Conclusion 9: Containment performance (survival, failure, or bypass), which is described by input parameters in all current source term codes, is a major factor affecting source terms. [(1986) NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Terms"]
o The consequences of severe reactor accidents depend greatly on containment safety features and containment performance in retaining radioactive material.
The early failure of the containment structures at the Chernobyl power plant contributed to the size of the environmental release of radioactive material in that accident. Maintaining the integrity of the containment can affect the source term by orders of magnitude. The NRC's 1986 reassessment of source term issues reaffirmed that containment performance "is a major factor affecting source terms." [(1990) NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants]
o Shortly after the accident at Three Mile Island, the NRC initiated a program to review the adequacy of the methods available for predicting the magnitude of source terms for severe reactor accidents. After considerable effort and extensive peer review, the NRC published a report entitled "Reassessment of the Technical Bases for Estimating Source Terms," NUREG-0956. As expected, the magnitude of the source term varies between different accident progression bins depending on whether or not containment fails, when it fails, and the effectiveness of engineered safety features (e.g., BWR suppression pool) in mitigating the release.
However, within an accident progression bin, which represents a specific set of accident progression events, the uncertainty in predicting severe accident 3
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments phenomena is great. [(1990) NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants]
o  With regard to the lack of treatment of some severe accident phenomena and containment failure modes, it would not be appropriate to attempt to identify or analyze all containment failure modes or scenario pathways. It is the intent of the latest sequence selection, reported in NUREG/CR-4624, to analyze at least the most risk significant of these pathways. [(1990) NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants]
* These NRC studies conclude that containments will fail in credible nuclear accidents.
Regretfully, we no longer need to rely on theoretical studies. We watched the containments at Fukushima Daiichi catastrophically fail.
o  Despite the efforts of the operators at the Fukushima Daiichi nuclear power plant to maintain control, the reactor cores in Units 1-3 overheated, the nuclear fuel melted and the three containment vessels were breached. Hydrogen was released from the reactor pressure vessels, leading to explosions inside the reactor buildings in Units 1, 3 and 4 that damaged structures and equipment and injured personnel. Radionuclides were released from the plant to the atmosphere and were deposited on land and on the ocean. There were also direct releases into the sea. [The Fukushima Daiichi Accident, IAEA Report by the Director General]
o  The common cause failures of multiple safety systems resulted in plant conditions that were not envisaged in the design. Consequently, the means of protection intended to provide the fourth level of defense in depth, that is, Prevention of the progression of severe accidents and mitigation of their consequences, were not available to restore the reactor cooling and to maintain the integrity of the containment. [The Fukushima Daiichi Accident, IAEA Report by the Director General]
o  The confinement [containment] function was lost as a result of the loss of AC and DC power, which rendered the cooling systems unavailable and made it difficult for the operators to use the containment venting system. Venting of the containment was necessary to relieve pressure and prevent its failure. The operators were able to vent Units 1 and 3 to reduce the pressure in the primary containment vessels. However, this resulted in radioactive releases to the environment. Even though the containment vents for Units 1 and 3 were opened, the primary containment vessels for Units 1 and 3 eventually failed. Containment venting for Unit 2 was not successful, and the containment failed, resulting in radioactive releases. [The Fukushima Daiichi Accident, IAEA Report by the Director General]
4
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments
* Now that we know containments have and will fail in credible nuclear accidents, it is important to understand why they may fail again.
1.1.3 Integrity of Facility Design Basis The DBA source term used for dose consequence analyses is a fundamental assumption and the basis for much of the facility design. Additionally, many aspects of an operating reactor facility are derived from the radiological design analyses that incorporated the TID-14844 accident source term.
* Otherwise stated, nuclear power plants were constructed with physical barriers and safety systems that were specifically designed to prevent and mitigate the TID-14844 maximum credible accident source term release example. As we now know that is no fault of TID-14844, its example of the maximum credible accident was far from conservative; it incorrectly assumed that containments would not fail. Unfortunatley, the designers of nuclear power plants did not fulfil their responsibility for carefully examining, in his [each] particular case, every significant facet of the [radiological] hazards and safety problem. Instead, they designed nuclear plants, and their safety systems, assuming containments would never fail. This is, in part, the fault of the AEC and now the NRC.
 
===Background===
An accident source term is intended to represent a major accident involving significant core damage not exceeded by that from any other credible accident. NRC staff experience in reviewing license applications has indicated the need to consider other accident sequences of lesser consequence but higher probability of occurrence. Facility-analyzed DBAs are not intended to be actual event sequences; rather, they are intended to be surrogates to enable deterministic evaluation of the response of engineered safety features (ESFs). These accident analyses are intentionally conservative to compensate for known uncertainties in accident progression, fission product transport, and atmospheric dispersion.
* After studying containment failures for the past 40 years and watching the containments at Fukushima Daiichi catastrophic fail, the NRC required that similar U.S. designed nuclear power plants install hardened containment vents to reduce the risks posed to the public from credible containment failure accidents.
* Now that we know containments will fail, it is important to understand why venting (intentionally releasing) large quantities of highly radioactive material directly to the environment would be required to reduce the risk to the public.
o    Nuclear power plants were designed and built based on the defense in depth philosophy. They were designed with four (4) physical barriers, such that, each barrier must fail before there would be a significant radiological release to the environment. The first barrier is the uranium-oxide fuel pellet. The second barrier is the fuel pellet cladding. The reactor coolant system provides a third barrier to fission product release. The final and ultimate barrier to fission product release is the reactor containment.
5
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments (1) The first barrier proved to be ineffective when the NRC discovered that high burnup fuel pellets could fragment, relocate axially and possibly disperse outside of the fuel rod during postulated design-basis accidents including, but not limited to, LOCA. This means that lesser (clad failure) accidents may have larger radiological release source terms than more severe (core melt) accidents.
[SECY-15-0148, Evaluation of Fuel Fragmentation, Relocation and Dispersal Under Loss-Of-Coolant Accident (LOCA) Conditions Relative to The Draft Final Rule on Emergency Core Cooling System Performance During A LOCA (November 30, 2015)]
(2) The second barrier, the fuel cladding, is probably the costliest engineering mistake. This design error, not the tsunami, created the high concentrations of hydrogen that exploded, causing the catastrophic containment failures at Fukushima Dai-ichi.
During accident conditions when the core materials are inadequately cooled, the fuel cladding (zirconium alloy) can overheat which promotes and accelerates a corrosion reaction commonly referred to as the Zirc-Water Reaction. When the core is no longer submerged in water, cladding surface temperature heats up with the uranium fuel being the heat source. At cladding surface temperatures in excess of 2000 0F the reaction rate is significant. At cladding surface temperatures approaching 2500 0F the resultant heat is enough to maintain a high reaction rate (exothermic) regardless of fuel temperature.
When needed the most, the fuel clad barrier will essentially self-destruct, adding heat to an overheated core while creating large concentrations of explosive gas.
The second barrier will contribute to, or directly, cause the failures of barriers three and four (reactor coolant system and containment).
It is also important to recognize that the fuel cladding is the only physical barrier between the spent fuel and the environment. Spent fuel pool Zirc-fires are also credible accidents that have may have the largest source terms.
(3) During core melt accidents, the third barrier (reactor coolant system) will fail from mechanical damage, core-melt creep rupture or high pressure melt ejection.
(4) According to NRC studies (referenced below), the fourth barrier, containment, is expected to fail as a result of direct attack by molten core debris. Drywell
[containment] rupture due to pedestal failure or rapid over pressurization is also an important contributor to early containment failure. Late failure of containment is also most likely to occur in the drywell [containment] but in the form of prolonged leakage past the drywell head. And, early containment failure in station blackout is dominated by hydrogen deflagrations.
Because of the relatively high probabilities that those [BWR Mark I and II]
containments would fail should an accident progress to melting the core, the NRC 6
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments issued Order EA-13-109, requiring hardened containment vents to be installed at specified plants. This NRC order, essentially, requires nuclear plant operators to intentionally vent (release) large amounts of highly radiological material to the environment, because the containment barriers were not designed to survive credible accidents, which include hydrogen explosions.
Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 13) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 14).
The NRC staff did not identify any IAEA Safety Requirements or Guides with enough detailed information relevant for use by Part 50 and 52 licensees and applicants as related to the topic of this RG.
4.1 Offsite Dose Consequences The licensee should use the following assumptions in determining the TEDE for persons located at or beyond the EAB:
: b. The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in International Commission on Radiological Protection (ICRP) Publication 30, Limits for Intakes of Radionuclides by Workers, issued in 1979 (Ref. 28).
* Since 1978, the ICRP has twice updated their recommendations. Nevertheless, the NRC continues to resist the adoption of more recent international standards that are based on the latest available scientific information of the biology and physics of radiation exposure. The NRC evaluated ICRP-103 in 2017:
o    The reason for performing this evaluation is that for certain plants, the margin for meeting acceptable dose limits in the control room for certain DBAs is very small. The concern is that the implementation of the new ICRP 103 values could cause these plants to exceed the regulatory dose limits.
o    In design basis radiological consequence analyses [in which containments do not fail], iodine isotopes are a major contributor to the estimated thyroid dose.
ICRP Publication 26 assigns a thyroid tissue weighting fact or of 0.03. ICRP Publication 103 assigns a thyroid tissue weighting factor of 0.04. It is seen therefore that the estimated fatal cancer for the thyroid is about 33 percent higher in the ICRP Publication 103 recommendations compared to ICRP Publication 26 recommendations. The evaluation of the application of the ICRP 103 iodine DCFs as compared to the ICRP 26 DCFs results in an increase in control room TEDE of approximately 23 to 25 percent.
7
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments o  The ICRP-103 evaluation confirmed the NRCs overriding concern that the implementation of the new ICRP 103 values could cause these plants to exceed the regulatory dose limits. Therefore, the NRC concluded: The NRC staffs decision to discontinuing the rulemaking activities associated with potential changes to the radiation protection and reactor effluents regulations was based on the knowledge that the current NRC regulatory framework continues to provide adequate protection of the health and safety of workers, the public, and the environment.
The NRCs reason for rejecting ICRP-103 is my overring concern with RG 1.183 (Revision 0) and DG-1389.
Sincerely, Brian D. Magnuson magnuson28@msn.com 1020 Station Blvd. #212 Aurora, IL 60504 Lead Emergency Management SpecialistConstellation (formerly Exelon Generation)
Former NRC Licensed Senior Reactor Operator/Operations Shift Manager Acting expressly as a member of the public.
 
==References:==
 
1988 - NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents 3.2 Barriers to Fission Product Release For the fission products generated within the core to reach the environment, they must pass through four fission product barriers. These barriers are Illustrated in Figure 3.1 for a pressurized-water reactor (PWR). The first barrier is the fuel pellet often referred to as the fuel matrix. The second barrier is the fuel pellet cladding. The reactor coolant system provides a third barrier to fission product release. The final and ultimate barrier to fission product release is the reactor containment.
(1962) TID-14844 It is the intent that this document to provide reference information and guidance on procedures and basic assumptions whereby certain factors pertinent to reactor siting as set forth in Title 10 Code of Federal Regulations Part 100 (1) (10 CFR 100) can be used to calculate distance requirements for reactor sites which are generally consistent with current siting practices.
8
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments For any proposed reactor: the performance experience accumulated elsewhere; the engineering safeguards; the inherent stability and safety features; and the quality of design, materials, construction, management and operation are all important factors that must be included in the evaluation of the suitability of a site.
For a particular site; size, topography, meteorology, hydrology, ease of warning and removing people in times of emergency, and thoroughness of plans and arrangements for minimizing injuries and interference with offsite activities, all enter an evaluation.
Consideration of these as well as other aspects of hazards evaluation involves so many different situations and such complex technological problems that it would be quite impossible to anticipate and answer all questions that will arise.
This technical document sets forth one method of computing distances and exposures, for one general class of reactors. In developing this example conservative assumptions have been intentionally selected.
Designers of reactors are expected to examine all significant aspects of the hazards and safety problem they believe are appropriate to the particular situation with which they are dealing. In any case, the designer and/or applicant bears the responsibility for justifying all the assumptions and methods of calculation used in a hazards evaluation. The fact that aspects of the problem are not considered in the example set forth here, does not in any way relieve the designer and/or applicant of the responsibility for carefully examining, in his particular case, every significant facet of the hazards and safety problem.
(1978) NUREG-0396 (EPA 520/1-78-016) PLANNING BASIS FOR THE DEVELOPMENT OF STATE AND LOCAL GOVERNMENT RADIOLOGICAL EMERGENCY RESPONSE PLANS IN SUPPORT OF LIGHT WATER NUCLEAR POWER PLANTS This report introduces the concept of generic Emergency Planning Zones as a basis for the planning of response actions which would result in dose savings in the environs of nuclear facilities in the event of a serious power reactor accident.
Design Basis Accidents Under NRC Regulations, the site/reactor design combination must be such that the consequences of design basis accidents are below the plume exposure guidelines of 10 CFR Part 100.
Class 9 Accidents Class 9 accidents cover a full spectrum of releases which range from those accidents which are of the same order as the DBA-LOCA type of releases; i.e., doses on the order of PAGs within 10 miles; to those accidents which release significant fractions of the available radioactive materials in the reactor to the atmosphere, thus having potential for life-threatening doses. The lower range of the spectrum would include accidents in which a core "melt-through" of the containment would occur.
As in the DBA-LOCA class, the doses from "melt-through" releases (involving thousands of curies) generally would not exceed even the most restrictive PAG beyond about 10 miles from a 9
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments power plant. The upper range of the core-melt accidents is categorized by those in which the containment catastrophically fails and releases large quantities of radioactive materials directly to the atmosphere because of over-pressurization or a steam explosion. These accidents have the potential to release very large quantities (hundreds of millions of curies) of radioactive materials. There is a full spectrum of releases between the lower and upper range with all of these releases involving some combination of atmospheric and melt- through accidents. These very severe accidents have the potential for causing serious injuries and deaths. Therefore, emergency response for these conditions must have as its first priority the reduction of early severe health effects. Studies have been performed which indicate that if emergency actions such as sheltering or evacuation were taken within about 10 miles of a power plant, there would be significant savings of early injuries and deaths from even the most severe atmospheric releases.
As discussed in Appendix Ill, the Task Force has concluded that both the design basis accidents and less severe core-melt accidents should be considered when selecting a basis for planning predetermined protective actions and that certain features of the more severe core-melt accidents should be considered in planning to assure that some capability exists to reduce the consequences of even the most severe accidents.
(May 1980) NUREG-0660 (Vol. 1), NRC Action Plan Developed as a Result of the TMI-2 Accident The actions to improve operational safety described in Chapter I are the more important responses to the accident at Three Mile Island. However, possible weaknesses in siting and plant design revealed by the accident should be evaluated and corrected, where necessary.
The NRC had been critically examining siting policy prior to the accident at Three Mile Island.
This examination was concluded in August 1979 (NUREG-0625). Recommended changes in siting policy from that study and other changes are soon to be issued in an Advanced Noticed of Rulemaking, as recently directed by the Commission. The new rule will be applicable to the siting of newly proposed nuclear plants; however, the treatment of existing nuclear plants, either operating or under construction, also will be considered.
The accident at TMI demonstrated the reality of the risk, previously only theoretically assessed, of accidents that result in substantial degradation and melting of the core. This risk arises from the fact that core-degradation accidents can lead to containment failure and the eventual release of large amounts of radioactivity to the environment.
(1983) NUREG/CR-3332, A Textbook on Environmental Dose Analysis In examining the source terms, therefore, one must assess the probability of escape from the containment under routine use conditions or in any postulated accident situation.
(1986) NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Terms" Conclusion 9. Containment performance (survival, failure, or bypass), which is described by input parameters in all current source term codes, is a major factor affecting source terms.
10
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments
[Containment is used 748 times in NUREG-0956.]
(December 1990) NUREG-1150, Severe Accident Risks: An Assessment for Five U.S.
Nuclear Power Plants The consequences of severe reactor accidents depend greatly on containment safety features and containment performance in retaining radioactive material. The early failure of the containment structures at the Chernobyl power plant contributed to the size of the environmental release of radioactive material in that accident. In contrast, the radiological consequences of the Three Mile Island Unit 2 (TMI-2) accident were minor because overall containment integrity was maintained and bypass was small.
Normally three barriers (the fuel rod cladding, the reactor coolant system pressure boundary, and the containment pressure boundary) protect the public from the release of radioactive material generated in nuclear fuel. In most core meltdown scenarios, the first two barriers would be progressively breached, and the containment boundary represents the final barrier to release of radioactivity to the environment.
Maintaining the integrity of the containment can affect the source term by orders of magnitude.
The NRC's 1986 reassessment of source term issues reaffirmed that containment performance "is a major factor affecting source terms" (Ref. 9.1). [Reassessment of the Technical Bases for Estimating Source Terms," United States Nuclear Regulatory Commission (USNRC) Report NUREG-0956, July 1986]
9.3 A. C. Payne, Jr., et al., "Evaluation of Severe Accident Risks: Peach Bottom Unit 2,"
Sandia National Laboratories, NUREG/CR-4551, Vol. 4, Draft Revision 1, SAND86-1309, to be published.
Shortly after the accident at Three Mile Island, the NRC initiated a program to review the adequacy of the methods available for predicting the magnitude of source terms for severe reactor accidents. After considerable effort and extensive peer review, the NRC published a report entitled "Reassessment of the Technical Bases for Estimating Source Terms," NUREG-0956 (Ref. 10.1).
As expected, the magnitude of the source term varies between different accident progression bins depending on whether or not containment fails, when it fails, and the effectiveness of engineered safety features (e.g., BWR suppression pool) in mitigating the release. However, within an accident progression bin, which represents a specific set of accident progression events, the uncertainty in predicting severe accident phenomena is great.
With regard to the lack of treatment of some severe accident phenomena and containment failure modes, it would not be appropriate to attempt to identify or analyze all containment failure modes or scenario pathways. It is the intent of the latest sequence selection, reported in NUREG/CR-4624, to analyze at least the most risk significant of these pathways.
: 11. Unfortunately, the necessary calculations cannot be made based on data and scientifically accepted principles and methodologies. Because of major inadequacies in the data base, because of the vast complexity of nuclear plants, because a tremendous 11
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments number of assumptions must be made, because all contributors to risk cannot be quantified, and because core meltdown phenomena are poorly understood, no one calculation of risk yields a remotely meaningful value of risk (Ref. C.5, p. 38109).
(1989, 1992) NUREG/CR-5247, RASCAL Version 2.1 Users Guide The Radiological Assessment System for Consequence Analysis (RASCAL) (Athey et al. 1989, 1992) is a set of personal computer-based tools. RASCAL Version 2.1 contains tools to estimate source term, atmospheric transport, and dose from a radiological accident (ST-DOSE),
to estimate dose from field measurements of radionuclide concentrations (FM-DOSE), and to compute decay of radionuclides (DECAY). RASCAL was developed for use by U.S. Nuclear Regulatory Commission (NRC) personnel who report to the site of a nuclear accident to conduct an independent assessment of dose projections.
Predicting [nuclear accident] doses or consequences . . . requires several steps: (1) predicting the quantity and timing of the release from the plant (source term), (2) predicting the movement of the plume (transport), and (3) predicting the dose from the plume and predicting the health effects from the dose. Each of these steps requires collection of appropriate data, and data collection and the subsequent computations are subject to uncertainties.
The largest single component of uncertainty is expected in the estimate of the source term.
Unanticipated catastrophic containment failure is a case in which the source term could be underestimated by a factor of 1,000,000 if monitor readings are used to estimate the source term. For lesser (non-core damage) accidents in which the total release is through a monitored pathway and consists mostly of noble gases, the source term uncertainty can be reduced.
However, the transport and dose uncertainties would remain unchanged.
(1989) NRC Generic Letter 89-16, Installation of a Hardened Wetwell Vent (2001) NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants (2007) The 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103. Ann. ICRP 37 (2-4).
Abstract - These revised Recommendations for a System of Radiological Protection formally replace the Commissions previous, 1990, Recommendations; and update, consolidate, and develop the additional guidance on the control of exposure from radiation sources issued since 1990.
Thus, the present Recommendations update the radiation and tissue weighting factors in the quantities equivalent and effective dose and update the radiation detriment, based on the latest available scientific information of the biology and physics of radiation exposure. They maintain the Commissions three fundamental principles of radiological protection, namely justification, optimisation, and the application of dose limits, clarifying how they apply to radiation sources delivering exposure and to individuals receiving exposure.
The Recommendations evolve from the previous process-based protection approach using practices and interventions by moving to an approach based on the exposure situation. They 12
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments recognise planned, emergency, and existing exposure situations, and apply the fundamental principles of justification and optimisation of protection to all of these situations. They maintain the Commissions current individual dose limits for effective dose and equivalent dose from all regulated sources in planned exposure situations. They reinforce the principle of optimisation of protection, which should be applicable in a similar way to all exposure situations, subject to the following restrictions on individual doses and risks; dose and risk constraints for planned exposure situations, and reference levels for emergency and existing exposure situations. The Recommendations also include an approach for developing a framework to demonstrate radiological protection of the environment.
(2013) EA-13-109, NRC Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions The events at the Fukushima Dai-ichi nuclear power plant following the March 2011 earthquake and tsunami highlight the possibility that events such as rare natural phenomena could challenge the traditional defense-in-depth protections related to preventing accidents, mitigating accidents to prevent the release of radioactive materials, and taking actions to protect the public should a release occur. At Fukushima Dai-ichi, limitations in time and unpredictable conditions associated with the accident significantly hindered attempts by the operators to prevent core damage and containment failure. In particular, the operators were unable to successfully operate the containment venting system. These problems, with venting the containments under the challenging conditions following the tsunami, contributed to the progression of the accident from inadequate cooling of the core leading to core damage, to compromising containment functions from overpressure and over-temperature conditions, and to the hydrogen explosions that destroyed the reactor buildings (secondary containments) of three of the Fukushima Dai-ichi units. The loss of the various barriers led to the release of radioactive materials, which further hampered operator efforts to arrest the accidents and ultimately led to the contamination of large areas surrounding the plant.
The events at Fukushima reinforced the importance of reliable operation of hardened containment vents during emergency conditions, particularly, for small containments such as the Mark I and Mark II designs. On March 12, 2012, the NRC issued Order EA-12-0501 requiring the Licensees identified in Attachment 1 to this Order to implement requirements for a reliable hardened containment venting system (HCVS) for Mark I and Mark II containments. Order EA-12-050 required licensees of BWR facilities with Mark I and Mark II containments to install a reliable HCVS to support strategies for controlling containment pressure and preventing core damage following an event that causes a loss of heat removal systems (e.g., an extended loss of electrical power). The NRC determined that the issuance of EA-12-050 and implementation of the requirements of that Order were necessary to provide reasonable assurance of adequate protection of the public health and safety.
While developing the requirements for a reliable HCVS in EA-12-050, the NRC acknowledged that questions remained about maintaining containment integrity and limiting the release of radioactive materials if the venting systems were used during severe accident conditions.
The requirements in this Order, in addition to providing a reliable HCVS to assist in preventing core damage when heat removal capability is lost (the purpose of EA-12-050), will ensure that venting functions are also available during severe accident conditions. Severe accident conditions include the elevated temperatures, pressures, radiation levels, and combustible gas 13
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments concentrations, such as hydrogen and carbon monoxide, associated with accidents involving extensive core damage, including accidents involving a breach of the reactor vessel by molten core debris.
As discussed in SECY-12-0157, the NRCs determination that a venting system should be available during severe accident conditions considered both quantitative assessments of costs and benefits, as well as, various qualitative factors. Among the qualitative factors, one of the more important is enhancing the defense-in-depth characteristics of Mark I and Mark II containments by addressing the relatively high probabilities that those containments would fail should an accident progress to melting the core. Other qualitative factors supporting installation of severe accident capable vents include addressing uncertainties in the understanding of severe accident events, supporting severe accident management and response, improving the control of hydrogen generated during severe accidents, improving readiness for external and multi-unit events, and reducing uncertainties about radiological releases and thereby improving emergency planning and response. The installation of a reliable, severe accident capable containment venting system, in combination with other actions such as ensuring drywell flooding capabilities, reduces the likelihood of containment failures and thereby enhances the defense-in-depth protections for plants with Mark I and Mark II containments.
APRIL 2015 - IAEA REPORT ON ASSESSMENT AND PROGNOSIS IN RESPONSE TO A NUCLEAR OR RADIOLOGICAL EMERGENCY - INTERNATIONAL EXPERTS MEETING VIENNA Harmonization of approaches and international guidance:
IAEA safety standards and guidance documents (for example, IAEA Safety Standards Series No.
GSR Part 7) takes into account the ICRP recommendations (specifically ICRP 103) but are not always identical. International, regional and national level documents (for example, the Nordic Flagbook and EPA-400 Protective Action Guidelines in the United States of America) can also have differences. Different standards or dosimetry techniques used either in different areas or by different agencies for the same areas, could make public safety messaging inconsistent and difficult. This underscores the importance of the IAEAs efforts to coordinate with the Accident State prior to release of information in an emergency. Similarly, neighboring countries may want to have pre-established arrangements to coordinate emergency management messaging.
It was highlighted that public protection strategies implemented in response to an emergency will always be a national level responsibility. However, internationally focused efforts to support the harmonization of response actions would be useful to avoid or explain situations where neighboring countries (or even countries far from the Accident State) are recommending actions that conflict with the response actions of the Accident State.
(2017) TRANSMITTAL OF THE CONTROL ROOM DOSE EVALUATION USING ICRP 103 DOSE CONVERSION FACTORS LETTER REPORT FOR USER NEED REQUEST NRR-2009-002 14
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments In design basis radiological consequence analyses, iodine isotopes are a major contributor to the estimated thyroid dose. ICRP Publication 26 assigns a thyroid tissue weighting fact or of 0.03.
ICRP Publication 103 assigns a thyroid tissue weighting factor of 0.04. It is seen therefore that the estimated fatal cancer for the thyroid is about 33 percent higher in the ICRP Publication 103 recommendations compared to ICRP Publication 26 recommendations. The evaluation of the application of the ICRP 103 iodine DCFs as compared to the ICRP 26 DCFs results in an increase in control room TEDE of approximately 23 to 25 percent.
On January 31, 2016, in SECY-16-0009, Recommendations Resulting from the Integrated Prioritization and Re-Baselining of Agency Activities, (ADAMS Accession No. ML16028A208),
the NRC staff requested Commission approval to implement recommendations on work to be shed, de-prioritized, or performed with fewer resources. Two of the items listed to be shed (i.e.,
discontinued) were the rulemakings that would have amended the radiation protection regulations in 10 CFR Part 20, and the reactor effluents regulations in 10 CFR Part 50, Appendix I.
The NRC staffs decision to discontinuing the rulemaking activities associated with potential changes to the radiation protection and reactor effluents regulations was based on the knowledge that the current NRC regulatory framework continues to provide adequate protection of the health and safety of workers, the public, and the environment. In the SRM for SECY 0009, dated April 13, 2016 (ADAMS Accession No. ML16104A158), the Commission approved discontinuing the two rulemaking activities. However, since all of the technical work for this task had been completed prior to the Commissions decision above, RES and NRR mutually decided to finish the report.
(2017) CONTROL ROOM DOSE EVALUATION USING ICRP 103 DOSE CONVERSION FACTORS (Information Systems Laboratories, Inc. Contract No. NRC-HQ-13-P-04-0099 -
GS23F0060L)
The reason for performing this evaluation is that for certain plants, the margin for meeting acceptable dose limits in the control room for certain DBAs is very small. The concern is that the implementation of the new ICRP 103 values could cause these plants to exceed the regulatory dose limits.
As noted earlier, the Peach Bottom BWR and the Wolf Creek and STP PWR were selected by NRR for this analysis. The DBAs that were studied were the FHA and the LOCA. NRR staff selected these accidents as they are bounding accidents for the control room dose. Note that only the control room TEDE dose is reported in this study.
Application of the ICRP 103 DCFs will result in an increase in the range of 23 to 25% in the TEDE doses for the control room. Note that only iodine radionuclides are included in these results. Including the ICRP 103 DCFs for all nuclides may (or may not) result in a smaller difference in the TEDE results.
Other References Appendix A to Part 50General Design Criteria for Nuclear Power Plants 15
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments Introduction Under the provisions of § 50.34, an application for a construction permit must include the principal design criteria for a proposed facility. Under the provisions of 10 CFR 52.47, 52.79, 52.137, and 52.157, an application for a design certification, combined license, design approval, or manufacturing license, respectively, must include the principal design criteria for a proposed facility. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
These General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The General Design Criteria are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units.
The development of these General Design Criteria is not yet complete. For example, some of the definitions need further amplification. Also, some of the specific design requirements for structures, systems, and components important to safety have not as yet been suitably defined.
Their omission does not relieve any applicant from considering these matters in the design of a specific facility and satisfying the necessary safety requirements. These matters include:
(1) Consideration of the need to design against single failures of passive components in fluid systems important to safety. (See Definition of Single Failure.)
(2) Consideration of redundancy and diversity requirements for fluid systems important to safety.
A "system" could consist of a number of subsystems each of which is separately capable of performing the specified system safety function. The minimum acceptable redundancy and diversity of subsystems and components within a subsystem, and the required interconnection and independence of the subsystems have not yet been developed or defined. (See Criteria 34, 35, 38, 41, and 44.)
(3) Consideration of the type, size, and orientation of possible breaks in components of the reactor coolant pressure boundary in determining design requirements to suitably protect against postulated loss-of-coolant accidents. (See Definition of Loss of Coolant Accidents.)
(4) Consideration of the possibility of systematic, nonrandom, concurrent failures of redundant elements in the design of protection systems and reactivity control systems. (See Criteria 22, 24, 26, and 29.)
It is expected that the criteria will be augmented and changed from time to time as important new requirements for these and other features are developed.
There will be some water-cooled nuclear power plants for which the General Design Criteria are not sufficient and for which additional criteria must be identified and satisfied in the interest of public safety. In particular, it is expected that additional or different criteria will be needed to take into account unusual sites and environmental conditions, and for water-cooled nuclear power 16
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments units of advanced design. Also, there may be water-cooled nuclear power units for which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General Design Criteria must be identified and justified.
Criterion 2Design bases for protection against natural phenomena. Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.
Criterion 4Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
Criterion 16Containment design. Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Criterion 50Containment design basis. The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.
This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by § 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, 17
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
Criterion 19Control room. A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.
[36 FR 3256, Feb. 20, 1971, as amended at 36 FR 12733, July 7, 1971; 41 FR 6258, Feb. 12, 1976; 43 FR 50163, Oct. 27, 1978; 51 FR 12505, Apr. 11, 1986; 52 FR 41294, Oct. 27, 1987; 64 FR 72002, Dec. 23, 1999; 72 FR 49505, Aug. 28, 2007]
Page Last Reviewed/Updated Wednesday, March 24, 2021 Appendix B to Part 50Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants III. Design Control Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and 18
 
RG 1.183 (Revision 1) / DG-1389: Magnuson Public Comments that deviations from such standards are controlled. Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.
Measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces.
The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The verifying or checking process shall be performed by individuals or groups other than those who performed the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, it shall include suitable qualifications testing of a prototype unit under the most adverse design conditions. Design control measures shall be applied to items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials; accessibility for inservice inspection, maintenance, and repair; and delineation of acceptance criteria for inspections and tests.
Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.
[35 FR 10499, June 27, 1970, as amended at 36 FR 18301, Sept. 11, 1971; 40 FR 3210D, Jan.
20, 1975; 72 FR 49505, Aug. 28, 2007; 84 FR 63568, Nov. 18, 2019]
Page Last Reviewed/Updated Monday, June 14, 2021 10 CFR 50.2 Definitions Design bases means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted "state of the art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.
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Latest revision as of 07:12, 18 November 2024

Transcript of the Advisory Committee on Reactor Safeguards Radiation Protection and Nuclear Materials Subcommittee Meeting, March 16, 2022, Pages 1-222 (Open)
ML22095A095
Person / Time
Issue date: 03/16/2022
From:
Advisory Committee on Reactor Safeguards
To:
Wang, Weidong, ACRS
References
NRC-1885
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