ML21225A565

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NRR E-mail Capture - Info and Action: Transmittal of NRC Comments and Slides Regarding August 3, 2021, Version of Industry'S Technology Inclusive Content of Application Project (Ticap) Guidance Document
ML21225A565
Person / Time
Issue date: 08/13/2021
From: Joseph Sebrosky
NRC/NRR/DANU
To: Afzali A
Southern Company Services
References
Download: ML21225A565 (126)
v • d • e


Text

From: Sebrosky, Joseph Sent: Friday, August 13, 2021 1:18 PM To: Afzali, Amir Cc: NICHOL, Marcus; TSCHILTZ, Michael; Shams, Mohamed; Smith - NRR, Brian; Sanfilippo, Nathan; HOLTZMAN, Benjamin; Christopher P. Chwasz; Jung, Ian;

'Tom King'; Thomas Hicks; Jim C. Kinsey Jr; Steven Nesbit; Chisholm, Brandon Michael; AUSTGEN, Kati; Cyril Draffin; Stutzke, Martin; Travis, Boyce; NRR_DANU_UARP Distribution; Orenak, Michael; Siwy, Alexandra

Subject:

info and action: Transmittal of NRC comments and Slides Regarding August 3, 2021, Version of Industry's TICAP Guidance Document Attachments: 8-17 TICAP NRC slides - final version.pptx; Industry TICAP Guidance Document Rev E - NRC comments 8-13-21.docx Amir Afzali Southern Company Services Licensing and Policy Director - Next Generation Reactors Mr. Afzali, The purpose of this email is to provide you with the attached NRC comments on Industrys August 3, 2021, Technology Inclusive Content of Application Project (TICAP) guidance document. Also please find attached the slides that the staff intends to use for the August 17, 2021, TICAP public meeting. Please note that the NRC comments found in the attached word file include yellow highlighted material. The yellow highlighted material was identified by the staff as significant comments. The yellow highlighted material is the basis for the material found in the slides. This email will be captured in ADAMS and the email will be made publicly available so that interested stakeholders will have access to the information to support the August 17, 2021, TICAP public meeting.

If you have questions regarding the attached documents please contact me.

Sincerely, Joe Sebrosky Senior Project Manager Advanced Reactor Policy Branch Office of Nuclear Reactor Regulation 301-415-1132

Hearing Identifier: NRR_DRMA Email Number: 1315 Mail Envelope Properties (PH0PR09MB74362535BD1F564D369C5055F8FA9)

Subject:

info and action: Transmittal of NRC comments and Slides Regarding August 3, 2021, Version of Industry's TICAP Guidance Document Sent Date: 8/13/2021 1:17:34 PM Received Date: 8/13/2021 1:17:00 PM From: Sebrosky, Joseph Created By: Joseph.Sebrosky@nrc.gov Recipients:

"NICHOL, Marcus" <mrn@nei.org>

Tracking Status: None "TSCHILTZ, Michael" <mdt@nei.org>

Tracking Status: None "Shams, Mohamed" <Mohamed.Shams@nrc.gov>

Tracking Status: None "Smith - NRR, Brian" <Brian.Smith@nrc.gov>

Tracking Status: None "Sanfilippo, Nathan" <Nathan.Sanfilippo@nrc.gov>

Tracking Status: None "HOLTZMAN, Benjamin" <bah@nei.org>

Tracking Status: None "Christopher P. Chwasz" <Christopher.Chwasz@inl.gov>

Tracking Status: None "Jung, Ian" <Ian.Jung@nrc.gov>

Tracking Status: None

"'Tom King'" <thomasking2993@gmail.com>

Tracking Status: None "Thomas Hicks" <hickste@earthlink.net>

Tracking Status: None "Jim C. Kinsey Jr" <jim.kinsey@inl.gov>

Tracking Status: None "Steven Nesbit" <steve.nesbit@lmnt-consulting.com>

Tracking Status: None "Chisholm, Brandon Michael" <BMCHISHO@SOUTHERNCO.COM>

Tracking Status: None "AUSTGEN, Kati" <kra@nei.org>

Tracking Status: None "Cyril Draffin" <cyril.draffin@usnic.org>

Tracking Status: None "Stutzke, Martin" <Martin.Stutzke@nrc.gov>

Tracking Status: None "Travis, Boyce" <Boyce.Travis@nrc.gov>

Tracking Status: None "Orenak, Michael" <Michael.Orenak@nrc.gov>

Tracking Status: None "Siwy, Alexandra" <alexandra.siwy@nrc.gov>

Tracking Status: None GROUP: "NRR_DANU_UARP Distribution"

<Prosanta.Chowdhury@nrc.gov>

Tracking Status: None

<Arlon.Costa@nrc.gov>

Tracking Status: None

<Amy.Cubbage@nrc.gov>

Tracking Status: None

<Jordan.Hoellman2@nrc.gov>

Tracking Status: None

<Dawnmathews.Kalathiveettil@nrc.gov>

Tracking Status: None

<Maryam.Khan@nrc.gov>

Tracking Status: None

<Margaret.O'Banion@nrc.gov>

Tracking Status: None

<Eric.Oesterle@nrc.gov>

Tracking Status: None

<Stephen.Philpott@nrc.gov>

Tracking Status: None

<William.Reckley@nrc.gov>

Tracking Status: None

<Elizabeth.Reed@nrc.gov>

Tracking Status: None

<Joseph.Sebrosky@nrc.gov>

Tracking Status: None

<John.Segala@nrc.gov>

Tracking Status: None

<Juan.Uribe@nrc.gov>

Tracking Status: None

<Nanette.Valliere@nrc.gov>

Tracking Status: None "Afzali, Amir" <AAFZALI@southernco.com>

Tracking Status: None Post Office: PH0PR09MB7436.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1151 8/13/2021 1:17:00 PM 8-17 TICAP NRC slides - final version.pptx 890778 Industry TICAP Guidance Document Rev E - NRC comments 8-13-21.docx 1636101 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Technology Inclusive Content of Application Project Public Meeting August 17, 2021 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 934 684 217#

Agenda Time Topic* Speaker 9:30 - 9:40 am Opening Remarks NRC/Southern 9:40 - 10:40 am First Issue - principal design criteria NRC/Southern 10:40 - 11: 50 am Other Significant Topics NRC/Southern 11:50 - 12 pm Stakeholder Questions All 12:00 - 1:00 pm Break All 1:00 - 1:10 pm Opening Remarks NRC/Southern 1:10 - 1:40 pm Continuation of Discussion of Other NRC/Southern Significant Topics 1:40 - 1:50 pm Stakeholder Questions All 1:50 - 2:00 pm Next Steps and Closing Remarks NRC/Southern

  • Note that Industry's TICAP guidance document is available at:

https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML21215A577 and industrys white paper on principal design criteria is available at:

https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML21214A008 2

TICAP Public Meeting

  • The purpose of this meeting is to discuss with the nuclear industry issues related to the draft guidance document for safety analysis report (SAR) content for an advanced reactor application based on the licensing modernization project (LMP) described in NEI 18-04
  • Key documents associated with this meeting are referenced in the meeting notice and include:
  • Industry-developed draft TICAP guidance document (ADAMS Accession No. ML21215A577)
  • Industry White Paper on Principal Design Criteria (ADAMS Accession No. ML21214A008)
  • Continuation of public meetings held on May 11, May 19, May 26 and June 23, 2021

ARCAP and TICAP - Nexus

  • Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.

Principal Design Criteria

  • The purpose of the PDCs is described in 10 CFR 50, Appendix A, as establishing the necessary design, fabrication, construction, testing and performance requirements for SSCs
  • GDCs in 10 CFR 50, Appendix A are not requirements for non-LWRs, therefore, non-LWR applicants would not necessarily need to request an exemption from the GDCs in 10 CFR Part 50 when proposing PDCs for a specific design.

Principal Design Criteria

  • Applicant must provide PDCs and supporting information that justifies to the NRC how their proposed PDCs demonstrate adequate assurance of safety and how their design meets their proposed PDCs to demonstrate adequate assurance of safety
  • During the June 23rd public meeting, the NRC raised some concerns and examples for discussion regarding the scope of PDCs that may not be addressed when developed using the LMP process but are included in GDCs and advanced reactor design criteria (ARDCs) which are expected to be used as insights during PDC development, for example:

o Normal operations (LMP is focused on licensing basis events (LBEs))

o Safe, stable end state (i.e., subcritical) o Construction, testing, and inspection

proposed scope of PDCs using LMP would be needed

Principal Design Criteria

  • NRC previously identified that using the LMP process may not address all aspects considered necessary for demonstrating adequate assurance of safety (e.g., normal operations, subcriticality, etc.) and is interested in how these would be proposed to be addressed via the TICAP guidance.

Example:

The LMP process is focused on off-normal events from anticipated operational occurrences (AOOs) to beyond design basis events (BDBEs) and identifies the design features, performance and special treatment needed to keep those events within the frequency-consequence (F-C) curve and cumulative individual risk targets. Dose at the exclusion area boundary (EAB) and cumulative individual risk are the only measures used as acceptance criteria. However, LMP does not address other concerns associated with the normal operation portion of the design basis, prevention of severe accidents, recovery from off-normal events or non-reactor on-site hazards. 7

Principal Design Criteria

  • NRC previously identified that the LMP process assigns special treatments to several design attributes (e.g., quality assurance, protection from external hazards, testability, inspectability, etc.) that are addressed in specific and cross-cutting ARDCs.

Examples:

  • Various ARDCs (ARDC 39 & 40 as examples) in RG 1.232 include criteria that the design of certain SSCs accommodate the capability for their inspection and testing. These kinds of considerations should be included when assessing SSC special treatments as they relate to associated PDCs, where applicable.

8

Principal Design Criteria

  • Industry White Paper on PDCs (ADAMS Accession No. ML21214A008) provided to NRC in early August notes the following regarding RG 1.232 ARDCs associated with normal operations and special treatment:

o Namely that industry believes that the TICAP approach demonstrates that its proposed PDCs focused on design and performance functions, combined with its programmatic requirements, meet the intent and purpose of the safety concepts embodied in the 10 CFR Part 50, Appendix A GDCs and/or RG 1.232 ARDCs

  • The NRC staff is still reviewing the Industry White Paper and the question on whether an exemption request is needed regarding the purpose of PDCs
  • At issue is the industry proposal to equate the Required Function Design Criteria (RFDCs) in NEI 18-04 to PDCs. As stated in the Industry White Paper, the RFDCs address the "How" portion of the design, whereas historical PDC address both "How" and "How Well."

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Principal Design Criteria

  • The NRC staff is considering the following options to move forward on TICAP guidance while the PDC issue is under review:

o The NRC staffs draft TICAP Regulatory Guide white paper may be updated after considering the following options:

The LMP-based approach provides an acceptable approach for identifying PDCs associated with off-normal conditions Review the applicants proposed treatment of areas associated with normal operations and areas associated with proposed special treatments (e.g., quality assurance, protection from external hazards, testability, inspectability) and identify if PDCs, as well as proposed complimentary design criteria (CDCs), are appropriate in this area.

The LMP-based approach cannot be reconciled with the current PDC framework under 10 CFR Parts 50 and 52, and an exemption may be required.

10

Principal Design Criteria Figure adapted from previous Proposed Industry Southern presentation Principal Design Required Criteria When? Required Safety Functions (RSFs)

Functional Design Criteria SR SSC Design Criteria (SRDC)

How?

(RFDC)

Functions Safety SR SSC SR SSC Special Provided in the Related Performance Treatment Design (SR) SSCs Targets Requirements LBEs from LMP Design Basis Design Basis PRA Safety (AOOs, DBEs, Accidents External Hazard Functions (PSFs) and BDBEs) (DBAs) Levels (DBEHLs)

Input to Design and Frequency- Other Risk Fundamental Consequence Significant Safety Functions and Cumulative Safety Functions Non-SR NSRST SSC (FSFs) NSRST SSC Risk Targets with ST Special Performance (NSRST) Treatment Content of Application Targets SSCs Requirements Other Safety Functions for Adequate DID What? Non-SR Other Safety With No ST Proposed Industry Functions SSCs (NST)

How Well?

Complimentary Design Criteria 11

Other Comments

  • Staff Feedback on Industry TICAP Guidance document found in ADAMS at MLxxxxxxx
  • Feedback includes both editorial comments and comments that are more significant
  • The more significant comments are highlighted in yellow
  • Significant comments include the following:
  • The NRC staff does not believe that the TICAP and ARCAP guidance are alone sufficient to support issuance of an operating license (Chapter 1 -

page 4)

  • TICAP alone does not provide reasonable assurance of adequate protection
  • The TICAP and ARCAP guidance support development of safety evaluation
  • Other considerations for issuance of operating license (e.g.,

environmental regulations) outside scope of TICAP/ARCAP support issuance of license 12

Other Comments

  • The expectation regarding programs is not clear (Chapter 1 - page 10)
  • Programs descriptions supporting the safety case should generally be a part of the licensing basis, submitted on the docket, and incorporated by reference into the application
  • Regarding discussion in TICAP Chapter 1 (page 10)
  • Industry proposes that this chapter be considered outside the licensing basis
  • Proposal not consistent with requirements such as 10 CFR 50.34 that includes a requirement that applications include a description of intended use of the reactor
  • General Comment that much of the proposed industry italicized text should be part of the guidance
  • What are the criteria?

13

Other Comments

  • The NRC staff is considering whether guidance should include an expectation that the SAR include a summary of the peer review scope, approach, and results
  • Confirm that SAR description of end state would include whether specific radioactive barriers fail
  • Propose inclusion of other information in the SAR relating to calculation of releases from AOOs
  • Design Basis Events (DBEs) (Section 3.4.1 - page 35)
  • Include additional information related to releases (e.g., timing, dispersion analysis)
  • Provide listing of settings for protection system functions used in DBE evaluations 14

Other Comments

  • Beyond Design Basis Events (BDBEs) (Section 3.5.1 - page 37)
  • Include additional information related to releases (e.g., timing dispersion analysis)
  • Design Basis Accidents (DBAs) (Section 3.6.1 - page 38)
  • Include additional information related to releases (e.g., timing, dispersion analysis)
  • Integrated Evaluations (Section 4.1.1 & 4.1.3 - page 40)
  • Describe not only results and margins but assumed plant and site parameters (where not addressed in Chapter 2)
  • Add guidelines to ensure consistency in the individual risk calculations.
  • Some aspects of the calculations will be plant-specific (e.g.,

meteorology) but some should be plant and site independent (e.g.,

source of exposure [cloud shine], risk coefficients, inhalation rate, etc.).

15

Other Comments

  • Plant Capability DID and Programmatic DID Summaries (Section 4.2

- pages 42 and 46)

  • Proposed guidance appears to minimize SAR discussion related to DID.
  • Section 4.2 guidance appears to contradict the statement in Section 1.3.3 that states DID is a key element of the LMP-based affirmative safety case and the demonstration of reasonable assurance of adequate protection of public health and safety.
  • Layers of Defense Evaluation (Section 4.2.1.2 - page 44)
  • The guidance should include an expectation that the applicant state the acceptance criteria for the affirmative safety case determination and explain how they are met, not just provide a statement that they are met.

16

Other Comments

  • Integrated DID Evaluation (Section 4.2.3 - page 47)
  • It would seem appropriate for this section to require a summary of how the attributes in NEI 18-04, Section 5.9.3, IDP Actions to Establish DID Adequacy were evaluated and determined adequate
  • NEI 18-04, Section 5.9.3 provides the most comprehensive list of what determines DID adequacy. Why it is not discussed in the TICAP guidance document is not clear
  • See beginning presentation on PDCs
  • General statement regarding language to the effect of "this section should affirmatively state..." (ex: Section 4.2.1 - page 46; 14.2, p. 49)
  • This language is used throughout the document . In some cases, a statement supported by background documentation could be appropriate; in others, it isn't clear to the staff how an affirmative statement adequately captures in the licensing basis how the affirmative statement is supported.

17

Other Comments

  • Safety-Related SSCs (Section 5.4 - page 50)
  • The NRC staff does not agree with the industry guidance that appears to suggest that not all of the DID features are in the design basis (i.e.,

not in the SAR).

  • Complimentary Design Criteria (CDCs) (Section 5.6 - page 54)
  • Why arent the CDCs that are necessary to achieve the success criteria established in the PRA considered to be a part of the PDCs, and included in that category for the NRC staff review as a part of the safety assessment process?
  • Design Basis External Hazard Levels (DBEHLs) (Section 6.1 - page 57)
  • Beyond listing a hazard level value in a table, the guidance should specify that the applicant describe how each DBEHL is used as an input parameter to the design analysis of safety-related SSCs.
  • DBEHLs are defined in TICAP guidance to include internal plant hazards (e.g., internal fire), contrary to NEI 18-04.

18

Other Comments

  • Reliability and Capability Targets (Section 6.2 - page 60)
  • Expectation regarding documentation of SSC-level targets is not clear.
  • SR and NSRST SSC reliability and capability targets should be described in the application.
  • Special Treatment Requirements (Section 6.3 - page 62)
  • Testing and verification of advanced reactor design features need to comply with 10 CFR 50.43(e) and 10 CFR Part 50 Appendix B, Section III Design Control, that require analysis, appropriate test programs, experience, or a combination thereof, or a prototype facility, before a design certification (DC), combined license (COL), manufacturing license (ML), operating license (OL), or standard design approval (SDA) lapproval.
  • The TICAP guidance focuses only on test programs and should be revised to reflect these other requirements.
  • Plant Programs (Chapter 8 - page 69)
  • There does not appear to be sufficient guidance regarding what information about the program needs to be provided in the SAR if an approved template is not utilized.

19

Timeline for Technology Inclusive Content of Application Project (TICAP) Guidance and Advanced Reactor Content of Application Project (ARCAP) Guidance (rev 7/13/2021)

NRC Comments based on TICAP Workshops 6/10/2021 NRC/Industry update ACRS Subcommittee on status of ARCAP/TICAP guidance documents 7/21/2021 TICAP Tabletop Exercises NEI Revision 1 of TICAP Guidance 2/1/2021 4/2/2021 Southern Revision E of TICAP Document Guidance Document 1/19/2022 8/3/2021 NRC/Industry brief ACRS Subcommittee NEI Revision 0 of TICAP Guidance on final ARCAP/ TICAP guidance Legend TICAP Workshops Document 2/9/2022 5/26/2021 8/27/2021 Industry Action 5/2/2021 NRC/Industry brief ACRS Full NRC TICAP Regulatory Guide (Draft) Committee on final TICAP 9/10/2021 guidance NRC Staff Action 3/3/2022 NRC/Industry brief ACRS Subcommittee on Industry/NRC Southern Revision B of TICAP Guidance ARCAP/TICAP guidance documents (NEI, NRC TICAP Regulatory Joint Action Document Rev0 and Staff Draft RG) Guide 4/15/2021 10/12/2021 3/25/2022 Jan Mar May Jul Sep Nov 2022 Mar 2022 1/30/2021 9/10/2021 Draft ARCAP Roadmap ISG, ARCAP ISG for ARCAP Application Outline Updated to be "Site Information," and ARCAP Chapters 9, Consistent with TICAP outline 10, 11, 12, and Technical Specifications issued 20

Next Steps - Future Milestones TICAP Near-Term Milestones Target Date NEI Revision 0 of TICAP Guidance Late August 2021 Document Update of NRC Draft Guidance Documents September 2021 ACRS Future Plant Subcommittee Meeting October 2021 on ARCAP/TICAP Guidance Documents 21

Technology Inclusive Content of Application Project For Non-Light Water Reactors Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Draft Report Revision E Issued for Nuclear Regulatory Commission Review and Comment Document Number SC-16166-104 Rev E Battelle Energy Alliance, LLC Contract No. 221666 SOW-16166 August 3, 2021 Prepared for:

U.S. Department of Energy (DOE)

Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Disclaimer This report was prepared as an account of work sponsored by an agency of the United States (U.S.) Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, nor Southern Company, Inc., nor any of its employees, nor any of its subcontractors, nor any of its sponsors or co-funders, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

ii

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Abstract This guidance document describes one acceptable means of developing portions of the Safety Analysis Report content for some advanced reactors. Another Nuclear Energy Institute publication, NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, describes the Licensing Modernization Project methodology for selection of Licensing Basis Events; safety classification of structures, systems, and components and associated risk-informed special treatments; and determination of Defense-in-Depth adequacy for non-light water reactors. The NEI 18-04 guidance was endorsed in Nuclear Regulatory Commission Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors. The guidance in this report focuses on the portions of the Safety Analysis Report that are generated by the application of the NEI 18-04 methodology. The goal of the standardized content structure and formulation is to facilitate efficient preparation by the applicant, review by the regulator, and maintenance by the licensee.

iii

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Table of Contents Disclaimer......................................................................................................................................... ii Abstract ........................................................................................................................................... iii List of Abbreviations ........................................................................................................................ v A. Introduction ............................................................................................................................... 1 Purpose ............................................................................................................................... 1 Background ......................................................................................................................... 2 Scope ................................................................................................................................... 3 Organization of this Report ................................................................................................. 7 B. Development of Guidance ......................................................................................................... 8 Overview ............................................................................................................................. 8 SAR Outline ......................................................................................................................... 8 General Instructions for Use of the Guidance .................................................................. 10 Alternative Licensing Paths ............................................................................................... 10 Two-Step Licensing ............................................................................................... 11 Design Certification ............................................................................................... 12 C. SAR Content Guidance ............................................................................................................. 14 D. Summary .................................................................................................................................. 71 Appendix AGlossary of Terms .................................................................................................. A-1 Appendix BSample LBE Descriptions ........................................................................................ B-1 iv

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case List of Abbreviations AOO Anticipated Operational NSRST Non-Safety-Related with Special Occurrence Treatment ARCAP Advanced Reactor Content of NST Non-Safety-Related with No Application Project Special Treatment BDBE Beyond Design Basis Event OL Operating license BOP Balance-of-plant PDC Principal Design Criteria CDC Complementary Design Criteria PRA Probabilistic Risk Assessment COL Combined construction permit PRISM Power Reactor Inherently Safe and operating license Module CP Construction permit PSAR Preliminary Safety Analysis Report CP/OL Construction permit followed by QHO Quantitative health objective operating license RCCS Reactor Cavity Cooling System DBA Design Basis Accident Ref General References DBE Design Basis Event RFDC Required Functional Design DBEHL Design Basis External Hazard Level Criteria DID Defense-in-Depth RG Regulatory Guide DOE Department of Energy RIPB Risk-informed and performance-EM Electromagnetic based FSAR Final Safety Analysis Report RPS Reactor protection system FSF Fundamental Safety Function RSF Required Safety Function IBR Incorporation by Reference RVACS Reactor vessel auxiliary cooling IDP Integrated Decision-Making system Process SAR Safety Analysis Report LBE Licensing Basis Event SCS Shutdown Cooling System LMP Licensing Modernization Project SGACS Steam generator auxiliary cooling LWR Light water reactor system MHTGR Modular high temperature gas- SR Safety-Related cooled reactor SRDC Safety-Related Design Criteria NEI Nuclear Energy Institute SSCs Structures, Systems, and non- Non-light water reactor Components LWR TEDE Total effective dose equivalent NRC Nuclear Regulatory Commission TICAP Technology Inclusive Content of Application Project v

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case A. INTRODUCTION Non-light water reactor (non-LWR) technologies are expected to play a key role in meeting the worlds future clean energy needs and are building on the foundation established by the current light water reactor (LWR) nuclear energy fleet. The Technology Inclusive Content of Application Project (TICAP) is an important part of the nuclear industry efforts to support the Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) initiatives to establish an efficient and cost-effective licensing framework for non-LWR reactors. This DOE cost-shared, owner/operator-led initiative produced this guidance document that describes one acceptable means of developing content for portions of the NRC license application Safety Analysis Report (SAR) for non-LWR designs implementing the Licensing Modernization Project (LMP)-based affirmative safety case (see below and Section 1.3).

This guidance is applicable to applicants that utilize the Licensing Modernization Project (LMP) methodology documented in the Nuclear Energy Institute (NEI) publication NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development (ML19241A336). The NEI 18-04 guidance was endorsed in NRC Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors (ML20091L698).

Purpose The guidance in this report is focused on the portions of the SAR containing material produced by implementing NEI 18-04. The intent of the guidance is to help ensure completeness of information submitted to NRC while avoiding unnecessary burden on the applicant and rightsizing the content of application commensurate with the complexity of the design being reviewed.

This guidance provides a standardized content development process and application format designed to facilitate efficient preparation by the applicant, review by the regulator, and maintenance by the licensee. The content formulation should optimize the type and level of detail of information to be provided in the SAR, based on the complexity of the designs safety case and the nexus between elements of the design and public health and safety.

The goal of TICAP was to develop license application content guidance with the following attributes:

  • Generically applicable to all non-LWR designs (i.e., technology-inclusive)
  • Utilizing a risk-informed and performance-based (RIPB) approach in order to:

o Ensure that the application content facilitates an NRC review that focuses on information that directly supports the safety case of nuclear power plants.

o Provide a consistent and logical approach for establishing portions of the SAR scope and guidelines for level of detail for advanced technologies and designs.

1

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case o Encourage efficiency and effectiveness in the review by focusing on the final results in the SAR as opposed to process details and ancillary information. Commented [A1]: This was brought up in the first review, but this statement is not consistent with NRC views This proposed, technology inclusive RIPB license application content should advance the on performance-based regulation and sends the wrong message to vendors and external stakeholders.

following:

  • The goal of having a safety-focused review that minimizes the burden on developers and owner-operators of generating, supplying, and maintaining information beyond that needed to make a safety determination and licensing decision
  • The NRC and industry objective of reaching agreement on information needed to demonstrate reasonable assurance of adequate protection for non-LWRs
  • NRCs stated objective and policy statement regarding the use of risk-informed decision-making to remove unnecessary regulatory burden This guidance document is intended to provide one acceptable approach for the development of those portions of the Safety Analysis Report required for a combined construction and operating license, a reactor construction permit followed by an operating license, or design certification that employs the LMP methodology endorsed by Regulatory Guide 1.233.

Background

Existing LWRs are the countrys largest source of emission-free, dispatchable electricity, and they are expected to remain the backbone of nuclear energy generation for years to come.

However, as the energy and environmental landscape has evolved, governmental and commercial interest has grown in advanced nuclear energy technologies that promise improved efficiency, greater fissile-fuel utilization, reduced high-level waste generation, better economics, and increased margins of safety. These technologies can expand upon the traditional use of nuclear energy for electricity generation by providing an alternative to fossil fuels for industrial process heat production and other applications.

Most of the currently operating nuclear power reactors were initially licensed in the 1970s and 1980s. The regulatory framework for those plants was developed over decades and tailored specifically for thermal neutron spectrum LWRs using light water coolant and moderator, zirconium-clad uranium oxide fuel, and the Rankine power cycle. Many advanced non-LWRs are in development, with each reactor design differing significantly from the current generation of LWRs. For example, advanced reactors might employ liquid metal, gas, or molten salt as a coolant, enabling them to operate at lower pressures but higher temperatures than LWRs. Some may operate with a fast or epithermal neutron spectrum rather than a thermal neutron spectrum.

A range of fuel types is under consideration, including fuel dissolved in molten salt and circulated throughout the reactor coolant system. In general, advanced reactors emphasize passive safety features that do not require operator action or rapid automatic action from powered systems to prevent or mitigate radionuclide releases. Materials may be different, particularly for the high-temperature reactors. Advanced reactors may produce energy for applications other than electricity generation, and they may be coupled to energy storage 2

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case systems. Given these technical differences, applying the current regulatory framework to advanced reactor designs would be difficult and inefficient. Changes to the current regulatory framework are needed to allow for a risk-informed safety evaluation and timely, efficient deployment of advanced reactor designs.

The DOE cost-shared TICAP, a utility-led project, was initiated to interact with NRC in support of the objective of modernizing the regulatory framework to improve the effectiveness and efficiency of NRC reviews. The project team included reactor owner-operators, reactor designers, and consultants, as well as a senior advisory group consisting of several former NRC commissioners. This guidance document reflects feedback received from stakeholders as part of several reviews and interactions. In addition, the team worked with four reactor designers to perform tabletop exercises that applied portions of preliminary TICAP guidance to develop notional SAR content. The final guidance in this document benefits from the lessons learned from those tabletop exercises.

TICAP built on the foundation that was successfully established in NEI 18-04. That document presented a technology inclusive, RIPB process for selection of Licensing Basis Events (LBEs);

safety classification of Structures, Systems, and Components (SSCs); and evaluation of Defense-in-Depth (DID) adequacy for non-LWRs based on a systematic evaluation of the safety case.

NRC endorsed the NEI 18-04 guidance with the publication of Regulatory Guide 1.233.

Although NRC staff expectations were delineated, this regulatory guide took no exceptions to the Commented [A2]: I would be curious as to why the TICAP LMP methodology as described in NEI 18-04. The TICAP guidance contained herein focuses on team felt it necessary to include this phrase. What does it imply? Perhaps we could ask about this during public the portion of the application related to the outcomes of applying the LMP approach and the meeting?

documentation of the applicants safety case. Ultimately, the information included in an application must demonstrate that the safety case of a particular design provides reasonable assurance of adequate protection of public health and safety.

Scope This guidance addresses only the portion of an advanced reactor SAR related directly to the implementation of the NEI 18-04 methodology. Concurrently with the development of this document, NRC is developing guidance for the remaining parts of an advanced reactor license application (including part of the SAR) in its Advanced Reactor Content of Application Project (ARCAP).1 With respect to the SAR portion of an advanced reactor application, this TICAP guidance pertains to most of the content in Chapters 1 through 8, while ARCAP provides guidance for Chapters 9 through 12 as well as portions of some of the earlier chapters. This SAR organization is significantly different from the approach that has evolved for large light water reactors. ARCAP also provides guidance for non-SAR parts of an application. The relationship between this TICAP guidance and ARCAP guidance is shown pictorially in Figure 1.

This document provides guidance on the following:

1 Slides from the February 25, 2021, NRC Advanced Reactor Stakeholder Meeting provide information on the ARCAP project and its relationship with the TICAP project. See ML21055A541 pp.91-105.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case

  • Scope of content to be included in an application (specifically, portions of the SAR)
  • Level of detail for the content
  • Structure to be used for providing the content The guidance on the SAR content scope and level of detail prescribes an appropriate level of design-specific information that should be provided to demonstrate that the designs safety case provides reasonable assurance of adequate protection of public health and safety. To establish an Commented [A3]: Is this appropriate to be in an industry effective and efficient technology inclusive SAR content guidance, this guidance is formulated to document? NRC would determine the level of detail necessary to demonstrate adequate protection, which describe an LMP-based affirmative safety case, defined as follows: would also need to include content from the ARCAP guidance and programmatic details reviewed through audit.

An affirmative safety case is a collection of technical and programmatic evidence The LMP safety case should be a piece of the adequate protection puzzle.

which documents the basis that the performance objectives of the technology inclusive Fundamental Safety Functions (FSFs) are met by a design during design-specific Anticipated Operational Occurrences (AOOs), Design Basis Events (DBEs),

Beyond Design Basis Events (BDBEs), and Design Basis Accidents (DBAs). This is accomplished by the following:

  • Identifying design-specific safety functions that are adequately performed by design-specific SSCs
  • Establishing design-specific features (programmatic, e.g., inspections, or physical, e.g., diversity) to provide reasonable assurance that credited SSC functions are reliably performed and to demonstrate DID adequacy The term safety case is a collection of statements that, when confirmed to be true by supporting technical information, establishes reasonable assurance of adequate protection from LBEs during operation of the nuclear power plant described in the application. An affirmative safety case is a holistic approach that focuses on demonstrating that a set of fundamental safety functions will be accomplished. It may be contrasted with a traditional compliance-oriented safety case that demonstrates the satisfaction of pre-established requirements using a prescribed set of processes or equipment. The TICAP affirmative safety case, in conjunction with other application Commented [A4]: This statement could imply that by information addressed by ARCAP, should be sufficient to warrant the development issuance of using this approach, its then not necessary to comply with the associated and applicable regulations, which is not the a safety evaluation in support of an operating license (OL), construction permit (CP), or design case.

certification rule.

Commented [A5]: Was inclusion of combined license (COL) an oversight in this sentence? The guidance is based The use of the LMP-based affirmative safety case to formulate the application content is on COL application as the base case.

intended to optimize the following: Commented [A6]: TH INL: There are other NRC decisions that need to be made such as in the area of the

  • The scope of information to be included based on relevance to the design-specific safety environmental regulations case
  • The level of detail formulation based on the importance of the functions and SSCs to the LMP-based affirmative safety case (RIPB details) and the relevance to the safety determination 4

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case The content structure facilitates efficient (i) preparation by an applicant, (ii) review by the regulator, (iii) maintenance by the licensee, and (iv) ease of use by stakeholders, including the public.

The baseline guidance presented in this document assumes an applicant is applying for a combined license (COL) under Subpart C of 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The guidance further assumes that the applicant is not referencing an existing design certification. Supplemental guidance also addresses two additional licensing approaches:

  • Two-step license under 10 CFR Part 50 herein referred to as CP/OL (a CP followed by an OL)

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Figure 1. Relationship of TICAP to an Advanced Reactor License Application 6

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Organization of this Report Section A of this report provides information on the purpose, background, and scope, as well as a road map for the content of this guidance document.

Section B provides information on the development of the guidance and general instructions for its use.

Section C is the chapter-by-chapter detailed guidance for the development of content at the appropriate level of detail in the sections of a SAR relating to the implementation of the NEI 18-04 methodology. The baseline guidance assumes the license applicant is requesting a COL for an advanced reactor under 10 CFR Part 52 and is not referencing a design certification.

Section C also contains adjustments to the baseline guidance if the applicant is following one of two different licensing approaches instead of the COL:

  • Design certification under 10 CFR Part 52 The adjustments are provided after the pertinent chapter or section of the baseline guidance.

Section D summarizes the results of the project.

Appendix A provides a glossary of terms.

Appendix B provides example LBE descriptions per Chapter 3 of the guidance.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case B. DEVELOPMENT OF GUIDANCE Overview This document describes the necessary information provided in portions of an applicants SAR to describe and support the LMP-based affirmative safety case for the reactor design, i.e., how the characteristics of the plant and its operation provide reasonable assurance of adequate protection of public health and safety from a radiological consequence perspective. The document presents an organization of the affirmative safety case material. It is important to recognize that this organizational approach is not the only way to present a safety case, so it should not be construed as a requirement for an advanced reactor applicant. However, for an applicant employing the LMP methodology, the following guidance provides a structure in which key technical information is provided in a clear and logical manner.

This content structure for the SAR should enable the following:

  • Efficient preparation by an applicant
  • Efficient review by the regulator
  • Efficient maintenance by the licensee
  • Ease of use by all stakeholders, including the public The information provided in the SAR should be relevant to the design-specific affirmative safety case. The level of detail of the information should be based on the importance of the safety functions, the SSCs, and the programs to the safety case.

SAR Outline Figure 2 provides a high-level outline of the portions of the SAR addressed by this guidance, and the following sections describe the content that applicants should provide therein. The outline is intended to present the overall safety case first and then provide the specific supporting design and operating details in subsequent chapters.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Chapter 1 General Plant and Site Description and Overview of the Safety Case Chapter 2 Methodologies and Analyses Chapter 3 Licensing Basis Events Chapter 5 Chapter 4 Safety Functions, Design Criteria, and Integrated Evaluations SSC Safety Classification Chapter 6 Safety-Related SSC Criteria and Capabilities Chapter 7 NSRST SSC Criteria and Capabilities Chapter 8 Plant Programs Figure 2. SAR Outline 9

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case General Instructions for Use of the Guidance Major divisions are referred to as chapters (e.g., Chapter 2 - Methodologies and Analyses).

Subdivisions of chapters at any level are referred to as sections (e.g., Section 2.1 - Probabilistic Risk Assessment).

Regular text provides instructions for the applicant under each chapter and section.

Italicized text provides additional information for context and perspective. It is intended to provide readily accessible supporting information, but it does not require direct action on the part of the applicant.

In addition to providing information in the SAR itself, applicants can provide material through Incorporation by Reference (IBR) or through General References (Ref).

  • IBR addresses material in design-specific topical reports, application-specific program control documents, industry standards, etc. When material is IBR, the scope must be Commented [A7]: The expectation regarding programs is clear, i.e., if only parts of the reference are applicable to the application, the SAR should not consistent. Programs supporting the safety case should generally be a part of the licensing basis, via IBR. However, state which parts are applicable. In addition, the applicant should identify any deviations the following text from Section 8 indicates that programs from the IBR material. Applicable portions of IBR material are considered part of the that support the safety case may or may not be IBR.

application and the licensing basis.2 The expectation is that each program supporting the LMP-

  • Ref identifies internal design or program documents or other sources of information that based affirmative safety case will have an associated contain additional detail. Unlike IBR, citing material as Ref in the SAR does not make it controlling document maintained by the design authority or licensee. This document could be a topical report that has part of the licensing basis. been reviewed and approved by NRC or an internal document.

NEI 98-03 (Rev. 1), Guidelines for Updating Final Safety Analysis Reports, Appendix A, Nuclear Energy Institute, June 1999 (ML003779028),3 provides additional discussion of IBR and Ref documents.

Alternative Licensing Paths NRC regulations provide applicants with multiple approaches for obtaining an OL for a nuclear power reactor. The baseline guidance presented in this section of the document assumes an Commented [A8]: May be better to characterize this applicant is applying for a COL under Subpart C of 10 CFR Part 52, Licenses, Certifications, and more generally as permits, licenses, and certifications rather than limit to OLs.

Approvals for Nuclear Power Plants. The guidance further assumes that the applicant is not referencing an existing design certification. In this scenario, the applicant would need to provide the maximum amount of information compared to other approaches that allow for more incremental provision of equivalent information.

2 The use of the term licensing basis in a general manner here and throughout this guidance document is intended to be consistent with, and limited to, the description of the Licensing Basis Process Development in Chapter 2 of Nuclear Energy Institute (NEI) publication NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development (ML19241A336) and is recognized as the LMP relevant portion of the overall licensing basis.

3 NEI 98-03 Rev. 1 was endorsed by Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance With 10 CFR 50.71(e), Nuclear Regulatory Commission, September 1999 (ML992930009).

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Advanced reactor applicants may choose alternative licensing pathways. This section also provides guidance for two alternative pathways deemed to be reasonably likely. Those pathways are:

  • Two-step licensingThe applicant first applies for and obtains a CP under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and subsequently applies for and obtains an OL also under 10 CFR Part 50. This pathway is herein referred to as CP/OL.
  • Design certificationThe applicant is a reactor vendor that applies for a standard design certification under 10 CFR Part 52, Subpart B. It does not contain site-specific information. A future applicant would reference the design certification along with site-specific and owner-specific information as part of a COL application.

Note that this guidance is not comprehensivethere are other potential licensing pathways involving Part 50 and Part 52 besides the ones addressed in this guidance document. At the time of this writing, these two alternative approaches plus the COL baseline approach were considered the most likely to be employed. Each is discussed in summary fashion below. With respect to the guidance, where there are adjustments to the baseline COL guidance, those adjustments are provided at the end of the applicable section or chapter in Section C. If there are not adjustments, the baseline guidance is assumed to apply to the alternative approaches.

Two-Step Licensing (CP/OL)

With this alternative, the applicant obtains a CP, constructs the plant, and obtains an OL under 10 CFR Part 50. Issuing a CP does not constitute approval to operate a facility, and it provides no finality for the design unless specifically requested by the applicant and approved by NRC.

Accordingly, NRC expects the information in a CP application to be supplemented, updated, and finalized in the OL application. CP applicants provide less information initially than in the OL application (e.g., operational programs are typically not described in the CP application). The SAR submitted as part of a CP application is referred to as a Preliminary Safety Analysis Report (PSAR).

The application content for all licensing paths will be impacted by the overall licensing strategy.

This impact is particularly pronounced for the CP licensing path because the degree of information which is needed in an application is highly dependent on the finality of the decision requested from NRC at the CP stage. Therefore, to optimize the applicability of the CP guidance provided in this document, it is assumed that the applicant will seek the minimum possible level of decision finality when applying for the CP.

With respect to the second part of the two-step licensing process, the scope and level of detail of a SAR for an OL application submitted under Part 50 are expected to be commensurate with the SAR for a COL submitted under Part 52 (the baseline process). One major difference between the approach taken in this guidance document and the large LWR Part 50 and 52 processes is 11

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case that the approach used herein, building on the LMP-based safety case, shifts from the compliance-based set of licensing requirements to a performance-based affirmative safety case. Commented [A9]: For the NRC's RG, we should consider including a clarification here that states something The chapter and section designations for the PSAR are the same as those used in Section C of like..."Notwithstanding, an applicant using the LMP approach must still comply with the regulations applicable this report. Adjustments to the COL guidance for the CP stage of two-step licensing are to their chosen licensing path or request exemptions, as provided at the end of the applicable section or chapter. The SAR content for the OL stage of appropriate."

two-step licensing is assumed to be the same as the content for the baseline COL guidance, so no adjustments to the guidance for the OL stage are provided. The OL stage SAR is often referred to as a Final Safety Analysis Report or FSAR, but this guidance uses the more general SAR term except when it is desirable to distinguish an OL SAR from a PSAR.

Design Certification With this alternative, the applicant submits a SAR as part of a 10 CFR Part 52 Subpart B design certification application.

There is significant similarity between SAR requirements for a COL application and SAR requirements for a design certification application. One exception is that a design certification does not address a specific site or site-specific owner-controlled programs. The other principal exception is that the design certification does not provide information on operational programs because they are the responsibility of the COL applicant that references the design certification.

Thus, the content of application discussions in the COL guidance are directly applicable to a design certification application with the exception of any site- or owner-specific information.

Note that a design certification will require the incorporation of a site parameter envelope to permit the evaluation of the proposed design certification to meet established regulatory criteria.

Such site parameter envelope information may be placed in Chapter 2. Commented [A10]: Should it be more specific regarding ARCAP Chapter 2 ISG?

The site parameter envelope specifies appropriately bounding parameters for a site that might be chosen by an applicant. However, in implementing the Probabilistic Risk Assessment (PRA) standard ASME/ANS RA-S-1.4-2021 referenced in the COL guidance, the concept of a bounding site is introduced for external hazard assessments. The bounding site is a hypothetical site defined as having a set of site characteristics that will be used to establish values for evaluation of the performance capabilities of the design. The site characteristics may be selected from site parameters from actual sites and may reflect hazards from different sites for different external hazards. For the bounding site, site-related parameters are defined using a set of external hazard conditions that are chosen to provide appropriately high external hazard design parameter values and the most adverse meteorological conditions and population data for assessing off-site radiological impact. These considerations are reflected in the selection of the Design Basis External Hazard Levels (DBEHLs) for the standard plant design. It should be noted that the bounding site is consistent with the site parameter envelope used in the design certification application. Ultimately, the COL applicant that references a design certification must confirm that the site characteristics are within the site parameter envelope or, in the COL, address any instances of nonconformance with the envelope.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case The chapter and section designations for the design certification SAR are the same as those used in Section C of this report. Adjustments to the COL guidance for a design certification are provided at the end of the applicable section or chapter. The SAR for a design certification application is often referred to as an FSAR, similar to a Part 50 OL application and a Part 52 COL application.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case C. SAR CONTENT GUIDANCE Section C of this report contains detailed guidance for the entire TICAP portion of the SAR, Chapters 1 through 8. A table of contents and a list of tables for the SAR Content Guidance are provided below.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case Table of Contents for the SAR Content Guidance 1 General Plant and Site Description and Overview of the Safety Case .................................. 18 1.1 Plant Description ............................................................................................................ 19 1.1.1 Reactor Supplier and Model ................................................................................... 19 1.1.2 Intended Use of the Reactor ................................................................................... 19 1.1.3 Overall Configuration .............................................................................................. 19 1.1.4 Description of Plant Structures, Systems, and Components .................................. 19 1.2 Site Description .............................................................................................................. 21 1.3 Safety Case ..................................................................................................................... 22 1.3.1 Safety Case Methodology ....................................................................................... 22 1.3.2 Fundamental Safety Functions ............................................................................... 22 1.3.3 Defense-in-Depth .................................................................................................... 23 1.4 Summary of Reference or Source Materials ................. Error! Bookmark not defined.23 1.4.1 Reference Designs, Licenses, or Certifications ...... Error! Bookmark not defined.24 1.4.2 Topical Reports ...................................................... Error! Bookmark not defined.24 1.4.3 Other Technical Reports (e.g., Environmental Report, Submitted Technical Reports, and Test Data Reports) ........................... Error! Bookmark not defined.24 1.4.4 Industry Codes, Standards, Guidance (e.g., ASME, ANS, ACI, and NEI) .............Error!

Bookmark not defined.24 2 Methodologies and Analyses ............................................................................................ 2425 2.1 Probabilistic Risk Assessment .................................................................................... 2425 2.1.1 Overview of PRA ................................................................................................. 2526 2.1.2 Summary of Key PRA Results .............................................................................. 2627 2.2 Source Term ............................................................................................................... 2728 2.3 DBA Analytical Methods ............................................................................................ 2829 2.4 Other Methodologies and Analyses .............................................................................. 29 3 Licensing Basis Events ....................................................................................................... 3031 3.1 Licensing Basis Event Selection Methodology ........................................................... 3031 3.2 LBE Summary ............................................................................................................. 3132 3.2.1 Summary Evaluation of AOOs, DBEs, and BDBEs ............................................... 3132 3.2.2 Summary Evaluation of DBAs ............................................................................. 3233 3.3 Anticipated Operational Occurrences ....................................................................... 3435 3.3.1 AOO-1 .................................................................................................................. 3435 3.4 Design Basis Events .................................................................................................... 3536 3.4.1 DBE-1 ................................................................................................................... 3536 3.5 Beyond Design Basis Events ....................................................................................... 3637 3.5.1 BDBE-1................................................................................................................. 3637 15

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case 3.6 Design Basis Accidents ............................................................................................... 3738 3.6.1 DBA-1 ...................................................................................................................... 38 4 Integrated Evaluations .......................................................................................................... 40 4.1 Overall Plant Risk Performance Summary ..................................................................... 40 4.1.1 Exclusion Area Boundary Dose ............................................................................... 40 4.1.2 EAB Boundary Early Fatality Risk ............................................................................ 40 4.1.3 Latent Cancer Risk ................................................................................................... 40 4.2 Defense-in-Depth ........................................................................................................... 41 4.2.1 Plant Capability Summary ....................................................................................... 42 4.2.2 Programmatic DID Summary .................................................................................. 45 4.2.3 Integrated DID Evaluation ....................................................................................... 47 5 Safety Functions, Design Criteria, and SSC Safety Classification ........................................... 48 5.1 Safety Classification of SSCs ........................................................................................... 48 5.2 Required Safety Functions ............................................................................................. 48 5.3 Required Functional Design Criteria and Principal Design Criteria ............................... 49 5.4 Safety-Related SSCs ....................................................................................................... 51 5.5 Non-Safety-Related with Special Treatments SSCs ....................................................... 52 5.5.1 Non-Safety-Related SSCs Performing Risk-Significant Functions ........................... 52 5.5.2 Non-Safety-Related SSCs Performing Safety Functions Necessary for Adequate DID ......................................................................................................... 53 5.6 Complementary Design Criteria ..................................................................................... 54 6 Safety-Related SSC Criteria and Capabilities ......................................................................... 56 6.1 Design Requirements for Safety-Related SSCs .............................................................. 56 6.1.1 Design Basis External Hazard Levels ....................................................................... 56 6.1.2 Summary of SRDC ................................................................................................... 58 6.1.3 Summary of DBEHL-Related Requirements for Non-Safety-Related SSCs ............. 59 6.2 Reliability and Capability Targets for SR SSCs ................................................................ 60 6.3 Special Treatment Requirements for SR SSCs ............................................................... 61 6.4 Descriptions for SR SSCs................................................................................................. 62 6.4.1 Description for SR SSC-1 ......................................................................................... 62 7 NSRST SSC Criteria and Capabilities ...................................................................................... 64 7.1 Reliability and Capability Targets for NSRST SSCs ......................................................... 64 7.2 Special Treatment Requirements for NSRST SSCs ..................................................... 6564 7.3 System Descriptions for NSRST SSCs.......................................................................... 6665 7.3.1 Description for NSRST SSC-1 ................................................................................... 66 8 Plant Programs .................................................................................................................. 6867 8.1 Maintenance Program ............................................................................................... 7069 8.2 Other Programs.......................................................................................................... 7170 16

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case 17

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case List of Tables for the SAR Content Guidance Table 1-1. Example Table of Design, License, and Certification References . Error! Bookmark not defined.24 Table 1-2. Example Table of Topical Report References ............ Error! Bookmark not defined.24 Table 1-3. Example Table of Other Technical Report ReferencesError! Bookmark not defined.24 Table 1-4. Example Table of Industry Codes, Standards, and Guidance References .............Error!

Bookmark not defined.24 Table 3-1. Example Summary Table of AOOs, DBEs, and BDBEs ............................................. 3233 Table 4-1. Example Table of LBE Risk Margins ............................................................................. 43 Table 4-2. Example Table of AOOs ............................................................................................... 44 Table 4-3. Example Table of DBEs ................................................................................................ 44 Table 4-4. Example Table of BDBEs .............................................................................................. 44 Table 5-1. MHTGR Required Safety Functions and Associated Required Functional Design Criteria......................................................................................................................... 50 Table 5-2. Example Table of Evaluation of SSCs for Core Heat Removal RSF .............................. 51 Table 5-3. Example Table of Evaluation of SSCs for PRA Safety Functions .................................. 52 Table 5-4. Example Table of SSCs Risk-Significant Due to F-C Curve ........................................... 53 Table 5-5. Example Table of SSCs Risk-Significant Due to Cumulative Risk ................................. 53 Table 5-6. Example Table of SSCs Risk-Significant Due to DID .................................................... 54 Table 5-7. Example Table of NSRST SSCs with Corresponding CDC ............................................. 55 Table 6-1. Example Table of Design Basis External Hazard Levels............................................... 57 Table 6-2. Example Table of Safety-Related Design Criteria for SSCs ......................................... 59 Table 6-3. Example Table of Non-Safety-Related SSCs Protecting SR SSCs from DBEHLs ........... 60 Table 6-4. Example Table of SR SSC Special Treatments ......................................................... 6261 Table 7-1. Example Table of NSRST SSC Special Treatments ................................................... 6665 Table 8-1. Example Table of Special Treatment Programs ...................................................... 6968 18

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case 1 GENERAL PLANT AND SITE DESCRIPTION AND OVERVIEW OF THE SAFETY CASE Commented [A11]: The IBR tables were removed from this version. Where does TICAP propose these tables should be located in the application?

The primary audience for the information in the SAR is the NRC. However, it is recognized that other stakeholders will also use the report. In particular, Chapter 1 is expected to be a resource for members of the public who want to understand key features of the plant and its operations and how it will provide adequate protection of public health and safety.

The applicant should provide general descriptive information about the plant and the site to provide context for the NRC safety review.

The descriptive information is divided into four sections as follows:

  • Section 1.1 provides descriptive information about the reactor and supporting systems that provides a framework and context for the information in subsequent chapters.
  • Section 1.2 provides descriptive information about the site that provides a framework and context for the information in subsequent chapters.
  • Section 1.3 provides a high-level overview of the LMP-based affirmative safety case in terms that can be understood by non-subject matter experts. This section is intended to summarize and provide context to the information in Chapters 3 through 7. Commented [A12]: Chapter 2 through 8? Chapter 2
  • Section 1.4 provides a listing of reference information and a road map to the chapters contains vital information for the LMP-based affirmative safety case.

in which the information is explicitly used to support the licensing basis.

While important for providing an understanding of the plant and the safety case, to establish review context, Sections 1.1 through 1.4 are not intended to be bases for regulatory conclusions by NRC and are therefore outside the licensing basis that is provided in Chapters 2 through 8. It Commented [A13]: Is this outside the licensing basis is understood that as part of the SAR, Chapter 1 will be maintained and updated as changes to approach consistent with associated regulatory requirements? For instance, 50.34 requires applications to Chapters 2-8 occur. include a description of intended use.

Applicants who use this guidance are expected to employ a variety of technologies differing in See comment below re: Section 1.1.2.

numerous aspects, including size, physical characteristics, materials, reactor power level, fuel Commented [A14]: We had discussions on this earlier type, coolant type, and power conversion system. Rather than prescribe a specific organization and I thought the industry wanted to reconsider. Are we ok with Chapter 1 of SAR not being licensing basis? Any for the information, this guidance specifies elements that should be included in an adequate delta/vagueness between Chapter 1 and Chapter 2-8 can description. cause confusion during the staff review and for the change process during operation.

With respect to the level of detail, the information should accomplish the following goals: Commented [A15R14]: I am OK with what they've proposed. I think they clarified the Chapter 1 is part of

1. Provide a stakeholder who is not an expert on nuclear technology with sufficient licensing basis but shouldn't be used as part of change process since it is summary info.

information to understand the purposes of the plant and the general means by which each purpose is accomplished

2. Provide a high-level summary of specific site information with a focus on the information that is relevant to the LMP-based affirmative safety case (Detailed site information Commented [A16]: There is no mention of site supporting the development of design basis external hazards should be provided in information in Chapter 2. Shouldnt there be at least an outline of the site information expected?

Chapter 2.)

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case

3. Provide a summary of the LMP-based affirmative safety case that is demonstrated in Chapters 2 through 8 in a concise and understandable manner for reviewers 1.1 Plant Description The intent of this section is to describe at an overview level the plant and the plant systems. The focus of this section is on those systems that are relevant to the LMP-based affirmative safety case; however, this section is expected to be a reasonably complete plant description that should enable the reader to understand the fundamental concepts of the plant and how it operates.

Elements of the plant description are listed below.

1.1.1 Reactor Supplier and Model Describe the reactor supplier and the model of the reactor. This should be a very brief description that allows the reviewer and stakeholder to identify and obtain background information (pre-application engagement information, publicly available vendor information, etc.) on the vendor and the design.

1.1.2 Intended Use of the Reactor Commented [A17]: Is a change to this outside of the licensing basis?

Describe the intended purpose of the reactor and the end uses. This could include descriptions of electricity generation, heat generation and use, industrial facilities served, micro-grids served, etc. This description is intended to inform reviewers and stakeholders but is not intended to provide justification for licensing the reactor. Commented [A18]: This is a regulatory requirement 50.34(a)(1)(ii)(a), and therefore probably needs to be in the 1.1.3 Overall Configuration design basis This section provides information on the overall layout of the plant and summarizes any features orof the plant layout that are significant from the perspective of the LMP-based affirmative safety case. This is intended to be a summary discussion of major plant attributes (e.g., thermal and electric power levels) accompanied by illustrative drawings, site plans, etc., that support the discussion. A system-level plant block diagram should be considered for this section, possibly color-coded to identify systems that are relevant to fundamental safety functions. The block diagram should align with the construct of Sections 1.1.4.1 through 1.1.4.4.

If the plant includes more than one reactor, the relationship of the reactors should be described, including major dependencies such as shared systems and structures.

1.1.4 Description of Plant Structures, Systems, and Components This section provides an overview of the plant SSCs. Given that this guidance is technology inclusive, the systems will vary among designs and technologies. The balance of this section provides examples of the information that should be provided. A detailed description of the plant SSCs is not expected in this section but rather a brief description of the SSCs such that the discussion of FSFs in Section 1.3.2 can be put into context with the overall plant design and 20

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case SSCs. This section should reference the location of more detailed SSC-specific information to the extent it is provided elsewhere in the SAR (i.e., subsections of Chapters 6 or 7).

This section should provide a high-level summary description, including figures and diagrams when the text description is not sufficient for general understanding. The NEI 18-04 methodology categorizes SSCs as Safety-Related (SR), Non-Safety-Related with Special Treatment (NSRST), and No Special Treatment (NST). SR and NSRST SSCs will be addressed Commented [A19]: For clarity, this is defined in NEI 18-04 in greater detail in Chapters 6 and 7, respectively. as Non-Safety-Related with No Special Treatment (NST).

Note: Light water reactor SARs contain detailed descriptions of some SSCs (e.g., reactivity control and control rod drive mechanisms, fuel, emergency cooling, etc.) in LWR SAR Chapters 4 and 6. Chapter 1 should not be the primary source of detailed SSC information.

Care must be taken to limit this section consistent with the objective of minimizing any redundancy with subsequent sections. Commented [A20]: Clarity: The previous sentence is about Chapter 1 in relation to the rest of the chapters.

Note: Sections 1.1.4.1 through 1.1.4.4 provide examples of how the information could be Should the last word be Chapters?

organized, recognizing that different designs and technologies will likely be organized differently, based on the systems in the design and relative importance of the SSCs. This section is organized in a traditional systems-centric manner in order to facilitate an overall understanding of the plant by a broad group of stakeholders. It could be organized differently for a given technology as long as the overall plant functionality can be understood.

1.1.4.1 Reactor Systems and Components

1. Nuclear design (e.g., neutron spectrum, reactor control, multi-unit reactor control)
2. Fuel
3. Reactor cooling
4. Reactivity control 1.1.4.2 Secondary Systems and Components
1. Heat transfer and cooling system
2. Power conversion system
3. Power transmission (e.g., switchyard) 1.1.4.3 Significant Support Systems and Components Commented [A21]: Suggest adding additional SSCs. It appears to be very short.
1. Fuel handling e.g., instrumentation and control, remote shutdown, EOF, TSC
2. Fuel management, including spent fuel storage
3. Control room
4. Electrical power
5. Radioactive waste 21

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case 1.1.4.4 Major Structures

1. Reactor building
2. Auxiliary, secondary, and support buildings
3. Cooling towers/systems
4. Co-located facilities (e.g., cogeneration, fuel processing, and buildings)

Two-Step Licensing (CP Content) Commented [A22]: This paragraph should require a description of the R&D planned (if any) in support of the For a CP application, Section 1.1 should follow the COL guidance, but the content will reflect design.

the preliminary nature of the design information as appropriate. The PSAR content would provide the general description of the plant and plant systems. The discussion of plant systems would be preliminary but sufficient to permit the reader to understand fundamental concepts of the plant and how it operates. The descriptions of the overall configuration in Section 1.1.3 also would be preliminary but with sufficient detail to support reader understanding of the design and how the LMP-based affirmative safety case will be developed. Discussion of systems and components in Sections 1.1.4.1 through 1.1.4.4 will be preliminary but sufficiently clear for the reader to understand the initial plant functionality.

Design Certification For a design certification application, the SAR would describe all but the site-specific details, which are deferred to the COL application.

1.2 Site Description This section provides a high-level overview of the site and the general vicinity of the licensed activities. Specific site attributes directly relevant to the affirmative safety case are included in Chapter 2 and are only briefly summarized in this section. This section should include a site Commented [A23]: Suggest this text state that site layout and maps of the general vicinity showing the site exclusion area, low population zone attributes relevant to the affirmative safety case be summarized in this section...to be consistent with DC text boundaries, nearby industrial facilities, and population centers sufficient to provide the reader an below.

overview understanding of the plant and the site. Discussion of site features (flood plains, access roads, etc.) can be included here to the extent that it facilitates an overall understanding of the safety case overview in Section 1.3.

Design Certification This section is not applicable because a design certification is not associated with a specific site.

However, the siting parameters assumed in the design certification application should be summarized here. Note that information on the assumed site parameter envelope should be provided in Section 6.1.1, Design Basis External Hazard Levels.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 1.3 Safety Case Commented [A24]: The words in 1.3 are not consistent with the words in the Scope section of the Introduction. In the Scope section (page 3) the definition of an affirmative This section provides a high-level overview of the safety case methodology and the outcome of safety case includes (a) Identifying design-specific safety executing the methodology. It focuses on the fundamental safety functions and how they are functions that are adequately performed by design-specific accomplished by the plant design described in Section 1.1. SSCs and (b) Establishing design-specific features (programmatic, e.g., inspections, or physical, e.g., diversity) to provide reasonable assurance that credited SSC functions 1.3.1 Safety Case Methodology are reliably performed and to demonstrate DID adequacy.

These correspond to the How and How Well aspects of This section should refer to NEI 18-04 and RG 1.233. If the applicant conforms to the guidance LMP which is also consistent with the safety case description contained in the July 2021 Southern Co. White in its entirety, then a brief statement of conformance, as demonstrated in Chapters 2 through 8, is Paper on non-LWR PDCs. However, Section 1.3 only adequate. An example statement is provided below. addresses the How portion of the safety case. Why is the How well portion of the safety case not described?

The selection of Licensing Basis Events (LBEs); safety classification of structures, systems, and components (SSCs) and associated risk-informed special treatments; and determination of Defense-in-Depth (DID) adequacy were done in accordance with the methodology of Nuclear Energy Institute report NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Report Revision 1 (August 2019), as endorsed by Nuclear Regulatory Commission Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors, Revision 0 (June 2020). This is demonstrated in Chapters 2 through 8. There were no deviations from the endorsed methodology. Commented [A25]: Clarity: In RG 1.233, staff positions exist while endorsing NEI 18-04 Report Revision 1. Should it If the applicant deviated from the methodology, either by not using parts of it or by using an be also discussed here in terms of being consistent with staff positions?

alternate method, a brief statement to that effect should be included here, and a discussion and justification should be provided in the relevant chapters of the SAR to support and clarify the licensing basis.

1.3.2 Fundamental Safety Functions The section should begin by establishing the overall performance objectivesthe regulatory dose criteria and quantitative health objectives (ref. NEI 18-04 Figure 3-1). The discussion should go through each FSF and summarize how it is satisfied. This section is not intended to be complete and exhaustive but is a high-level summary for general consumption. A systematic and thorough discussion of the safety case is provided in subsequent chapters. LBEs and event sequences relevant to each FSF are discussed in detail in Chapter 3, integrated evaluations and overall risk are discussed in Chapter 4, safety functions, design criteria, SSC safety classification, SSC criteria and capabilities are discussed in detail in Chapters 5 through 7, and plant programs supporting reliability and availability of SSCs are discussed in Chapter 8.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Note: The concept of FSFs goes back to International Atomic Energy Agency TECDOC-15704.

NEI 18-04 Section 3.3.4 discusses the FSFs as used in the LMP methodology. RG 1.233 presents the FSFs in a slightly different form from NEI 18-04, but the differences are more stylistic than substantive. The applicant may choose to express its FSFs consistent with either NEI 18-04 or RG 1.233. If the applicant adds one or more FSFs or modifies its FSFs substantially from those documented in NEI 18-04 or RG 1.233, the applicant should discuss the basis for the selection of its FSFs. The form of the FSFs provided below is taken from NEI 18-04.

1.3.2.1 Retaining Radionuclides This section should provide an overview of how the plant design accomplishes the FSF. This should include a high-level discussion of location and types of radiological inventory and the various SSCs that are available to prevent or mitigate releases through various modes of operation, including response to off-normal events or accidents.

1.3.2.2 Controlling Heat Generation This section should provide an overview of how the plant design accomplishes the FSF. This should include a high-level discussion of SSCs that are utilized to control heat generation through various modes of operation, including response to off-normal events or accidents.

1.3.2.3 Controlling Heat Removal This section should provide an overview of how the plant design accomplishes the FSF. This should include a high-level discussion of passive and active heat removal SSCs and their roles through various modes of operation, including response to off-normal events or accidents.

1.3.3 Defense-in-Depth This section should provide an overview of the DID aspects of the design. In this overview, the applicant should cite key examples of design features and programmatic elements included in the DID baseline described in detail in Chapter 4. DID is a key element of the LMP-based affirmative safety case and the demonstration of reasonable assurance of adequate protection of public health and safety.

4 Proposal for a Technology-Neutral Safety Approach for New Reactor Designs, International Atomic Energy Agency, Technical Report IAEA-TECDOC-1570, 2007.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 2 METHODOLOGIES AND ANALYSES Certain analyses and analytical tools (methodologies) are used in the identification of licensing Commented [A26]: General comment.: Much of the italic basis events, the evaluation of the consequences of such events, or assessing the performance of text should be part of the guidance (I.e., non-italic text). See specific comments below.

SR and NSRST SSCs. This chapter of the SAR presents information on some of those analyses and analytical tools. It is intended primarily for cross-cutting information or evaluations that Commented [A27]: Why only some of? Deleting it seems reasonable. Should it be something like "those support multiple LBEs or SSCs. Providing that information or evaluation upfront in one place is analyses necessary to support the safety case and intended to make the documentation that follows in subsequent chapters more efficient and regulatory requirements" concise.

The amount of information provided in this chapter will depend in part on the degree of pre-licensing engagement activities and associated NRC technical reviews and approvals. If limited Commented [A28]: Not sure if the "amount of pre-licensing engagement activities occurred, the level of detail provided in this chapter would information" is dependent on pre-application engagement, rather the form that the information is presented may need to be of sufficient detail that NRC can perform its review. If pre-application submittals change, i.e., topical report IBR or direct SAR text.

(e.g., topical reports) were made and NRC approvals were obtained for some or all of these topics, then the detailed review of these topics will have been documented through a separate and limited licensing processaction. If the review is occurring through separate documents, then only a high-level summary of the topic is required with appropriate references to the separate licensing documents, including references to the pertinent NRC safety evaluations. Commented [A29]: This text should make clear that if TR are used as part of the SAR analysis then those TR or the Note that the extent to which the applicant has utilized the pre-application engagement process relevant portions should be IBR'd into the relevant SAR sections.

with the NRC does not influence the scope and level of detail of information that must be provided by the applicant to demonstrate reasonable assurance of adequate protection.

However, it will likely influence the timing of the NRC reviews of the information that is in the SAR (or IBR).

The number and scope of these methodologies and analyses will vary depending on the technology and the safety case. Several of the methodologies and analyses are expected to be common to all applications and are set forth below. Others may be included in this chapter, depending on the specific details of the application.

Beyond the methodologies and analyses discussed in this guidance document, it is expected that Chapter 2 will also serve as the location for other SAR material addressed by ARCAP guidance.

This would include summaries of the site-related information and analyses used to develop the DBEHLs documented in Section 6.1.1.

2.1 Probabilistic Risk Assessment The PRA is the plant model that provides an integrated assessment of risk to the public from the nuclear power plant. A technically acceptable PRA is essential for implementing the NEI 18-04 methodology. The purpose of this section is to summarize elements of the PRA that are essential to the NEI 18-04 affirmative safety case without duplicating PRA products described in other chapters. The PRA information included in the SAR should be at a summary level only as described below. It should address the requirement in 10 CFR Part 52 that the SAR includes a 25

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case description of the design-specific PRA and its results. It is included near the beginning of the SAR because of the PRAs prominent role in exercising the NEI 18-04 methodology.

Key products of the PRA are reflected in other parts of the application. Chapter 3 presents the LBEs supported by the PRA (Anticipated Operational Occurrences [AOOs], Design Basis Events

[DBEs], and Beyond Design Basis Events [BDBEs] and includes the LBE descriptions, frequencies and uncertainties, consequences and uncertainties, and evaluation of risk significance against the LMP Frequency-Consequence Target. Chapter 4 shows the integrated risks across all the LBEs and compares them to NEI 18-04 cumulative risk metrics.

Uncertainties in the PRA results are considered as part of the DID evaluation described in Chapter 4. Chapter 5 reflects PRA safety functions and PRA success criteria. The purpose of this section is to summarize results and insights from the PRA that are essential to the LMP-based affirmative safety case without duplicating PRA products described in other chapters. The applicant maintains complete PRA documentation in its plant records.

2.1.1 Overview of PRA This section summarizes the scope, methodology, and pedigree of the PRA. The pedigree is intended to be (i) a statement of conformance (with any deviations) with the advanced non-LWR PRA standard,5 ASME/ANS RA-S-1.4-2021, the manner in which the standard was applied, and PRA peer review findings, or (ii) an alternative means of demonstrating PRA technical adequacy that may be proposed by the applicant.

The discussion should include the following items:

  • A statement that describes how the applicant used the non-LWR PRA Standard ASME/ANS-RA-S-1.4-2021 to establish the technical adequacy of the PRA, including the scope of technical requirements that were addressed.
  • A statement that a peer review was completed following the non-LWR PRA Standard Commented [A30]: The staff is considering whether and the guidance in NEI 20-09, Rev. 0, Performance of PRA Peer Reviews Using the guidance should include an expectation that the SAR include a summary of the peer review scope, approach, and results.

ASME/ANS Advanced Non-LWR PRA Standard6

  • Discussion of how the NRC regulatory guide that endorses the non-LWR PRA standard was implemented (pending finalization of the regulatory guide)
  • Identification of the sources of radionuclides addressed and the sources of radionuclides that were screened out
  • Discussion of how multi-reactor scenarios were addressed, if applicable
  • Identification of the internal and external hazards (and supporting methods, data, site information and relevant guides and standards) that were included in the PRA and the ones that were screened out (Other external hazards not supported by a probabilistic 5 ANSI/ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants, American Society of Mechanical Engineers and American Nuclear Society, approved January 28, 2021.

6 NEI 20-09, "Performance of PRA Peer Reviews Using the ASME/ANS Advanced Non-LWR PRA Standard," Rev 0, Nuclear Energy Institute, August 2020.

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Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case hazard analysis will be covered by DBEHLs that are determined using traditional deterministic methods as described in ARCAP guidance related to site information.)

  • Identification of the plant operating states that were included and those that were screened out Commented [A31]: The NRC staff may want to consider
  • Identification of the software and analytical tools that were used to perform the event whether certain PRA assumptions should be included in the sequence modeling and quantification, determine the mechanistic source terms, and SAR. For example, those assumptions that if changed or perform radiological consequence evaluations, including the cumulative dose calculations found to be inappropriate could result in transitions from one LBE category to a higher frequency category or change that form a part of the DID evaluation in Chapter 4 (with appropriate references to in SSC classification or selection of special treatments.

technical and/or topical reports provided as applicable) Could this be included as a clarification or exception in the NRCs RG or handled as part of the licensing basis change process?

The assumptions made in performing the PRA will not be included in the SAR but will be available for NRC audit. Additional considerations include that although the non-LWR PRA Std address assumptions extensively, some of the Note: This guidance document does not address SAR content for a PRA that has not been peer key assumptions can be important regarding the staffs technical acceptability evaluation and the licensing basis. It reviewed using the non-LWR PRA standard. In such an instance, the information to be provided is a topic of high interest to the staff during the review. In on the PRA, either in the SAR or other documentation, may be more extensive than the guidance the Std, the expression Key Assumption is also used and provided herein. defined as Key assumption: an assumption made in response to a key source of uncertainty in the knowledge that an alternative assumption would produce different Two-Step Licensing (CP Content) results, or an assumption that results in an approximation made for modeling convenience in the knowledge that a At the CP stage, neither the plant design nor the PRA is expected to have the level of maturity more detailed model would produce different results. The that will be necessary to support an OL application. At the CP application stage, the applicant term different results refers to a change in the plant risk profile (e.g., RCF, the set of initiating events, and event should describe its ultimate intended approach for qualifying the PRA. If conformance to sequences that contribute most to the RCF) and the ASME/ANS RA-S-1.4-2021 is planned, a simple statement to that effect should be sufficient. If associated changes in insights derived from the changes in risk profile.

the applicant intends to use another PRA methodology, that planned approach for establishing PRA technical adequacy should be described. In either case, the applicant should address the It may be possible to address via changing the sentence to last five items in the Section 2.1.1 list, consistent with the state of the plant design and the PRA something like: The assumptions made in performing the PRA will not be included in the SAR beyond that information at the time of CP application. To be clear, no PRA peer review should be required at the CP that becomes key to other parts of the SAR (e.g., Chapters application stage. 3-7). However, the PRA and related assumptions will be available for NRC audit.

Design Certification Commented [A32]: The NRC staff may want to consider providing clarification in its RG on this position. This could Section 2.1.1 should describe adjustments made to the PRA and uncertainty assessments to vary depending on the applicants desire/request for finality address the bounding site characterizations and SSC fragilities based on the DBEHLs described and the maturity of the design.

in Section 6.1.1. The degree to which the use of the bounding site characterizations could affect Note: In addition, even if it is a CP stage, absence of a peer analyses performed in other chapters and sections would be addressed in the descriptions of review, self-assessment, or independent review will place those analyses and results (determination of LBEs, SSC classification, etc.). the burden of PRA acceptability review squarely with the NRC staff. Less work for the applicant but more work for NRC staff.

2.1.2 Summary of Key PRA Results The staff also notes that RG 1.247, which will endorse the Because NEI 18-04 is a risk-informed methodology, key PRA results are incorporated in the NLWR PRA standard and industry peer review guidance in NEI 20-09, is expected to state that no PRA peer review is descriptions of the outputs of the methodology provided in the SAR. Those results are not required for DC, SDA, ML, COL, CP, or OL applications. This repeated here, but this section provides pointers to those PRA results. position is consistent with SRP 19.0 (refer to Footnote 3 on Page 19.0-13) and with DC/COL-ISG-028, Page 8. Applicants who not have a peer review should expect a longer staff review relative to applicants who do have a peer review.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case The applicant should provide a statement such as the following, identifying those parts of the SAR that include key PRA results:

Key PRA results are provided in subsequent chapters of the SAR.

  • Chapter 3 presents LBEs that are supported by event sequences in the PRA. It includes a plot of the frequencies, consequences, and uncertainties of these LBEs with a comparison against the Frequency-Consequence Target in NEI 18-04 Figure 3-1.
  • Chapter 4 presents the integrated risks across all of the LBEs and compares them to the NEI 18-04 cumulative risk metrics. It also describes the DID evaluation, which is informed by uncertainties in the PRA results.
  • Chapter 5 presents the PRA safety functions addressed by SR SSCs and NSRST SSCs.
  • Chapters 6 and 7 address reliability and capability targets for SR SSCs and NSRST SSCs. These targets are informed by inputs from the PRA. Commented [A33]: Where would the application address human actions relied upon to fulfill SR and NSRST functions in the PRA?

Two-Step Licensing (CP Content)

With respect to results, the COL guidance is applicable to a CP application, with the understanding that the Chapter 3 and Chapter 4 results will be preliminary relative to those to be presented in support of an OL application. The PSAR should include a discussion of how the PRA will be used during the design and construction of the plant.

2.2 Source Term Source term refers to the type, quantity, and timing of the release of radioactive material from a facility during licensing basis events. The source terms vary with the reactor type, plant design, operating characteristics, and the nature of the events. For an LMP-based affirmative safety case, the expectation is that the designer will use a mechanistic source term, consistent with the advanced non-LWR PRA standard definition (see glossary in Appendix A). To the extent that mechanistic source term information is common to some or all the events considered for the reactor, that information may be provided in this section rather than with each event. This may include references to fuel qualification and performance topical reports and the associated NRC safety evaluations.

The applicant should quantify all relevant radionuclide inventories prior to the beginning of the event sequence. For light water reactors with a defined solid core region, this quantification is typically done with computer codes such as the SCALE package. The applicant should describe the key inputs used and associated bases, such as the quantity of fissile material, core operating history, and core operating characteristics. The applicant should address the applicability of the analytical methodology to the characteristics of the reactor, including a discussion of the underlying experimental or analytical basis. The applicant should assess the uncertainty associated with the calculation and make appropriate allowances for it. For sources other than a 28

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Content for a Licensing Modernization Project-Based Affirmative Safety Case defined solid core region, the applicant should describe the basis for the quantity and activity of the material present.

The applicant should address the transport of the radioactive material from its point of origin to the accessible environment. For light water reactors, this is typically done with computer codes such as LOCADOSE for design basis events or MAAP for beyond design basis events. The applicant should describe the available pathways for attenuation, retention, and transport of radionuclides. This includes the description of physical phenomena or empirical justification for the attenuation, retention, and transport of radionuclides through each barrier between the origin and the accessible environment. The applicant should address the applicability of the analytical methodology to the characteristics of the plant, including a discussion of the underlying experimental or analytical basis. The applicant should assess the uncertainty associated with the calculation and make appropriate allowances for it. Mechanistic source terms employed in the PRA are subject to the technical requirements in the non-LWR PRA Standard ASME/ANS RA-S-1.4-2021.

Two-Step Licensing (CP Content)

For a CP application, Section 2.2 will mirror the discussions above but will reflect the preliminary nature of the design information as appropriate. The PSAR should describe the technical areas that require research and development to confirm the assumptions and methodologies used to present the mechanistic source term.

2.3 DBA Analytical Methods Deterministic calculations of DBA sequences are typically performed using one or more computer codes that constitute an analytical model of the plant response. Examples for this practice in light water reactors include the RETRAN and RELAP computer codes. If the models indicate a release of radionuclides, the mechanistic source term discussed above would also be involved in the calculation of consequences. The applicant may elect to describe the analytical methods associated with multiple DBAs in this section of Chapter 2. Multiple subsections (2.3.1, 2.3.2, etc.) can be used to describe multiple methods.

The applicant should describe the overall analytical methodology and identify and describe the significant computer codes used to model the plant response. The applicant should address the Commented [A34]: The section below highlights that applicability of the analytical methodology to the characteristics of the plant, including a other generic analyses should provide key inputs and assumptions; these are more important for the DBA discussion of the underlying experimental or analytical basis. Typically, this is done through analyses as they act as input to other NRC-reviewed and approved topical reports that are incorporated by reference in the SAR or sections/requirements, and should be called out here.

through technical reports that are summarized in the SAR and available for regulatory audits.

Two-Step Licensing (CP Content)

For a CP application, Section 2.3 should mirror the COL guidance, but the PSAR will reflect the preliminary nature of the design information as appropriate. The applicant should describe the 29

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Content for a Licensing Modernization Project-Based Affirmative Safety Case technical areas that require research and development to confirm the assumptions and methodologies used to present the adequacy of the design.

2.4 Other Methodologies and Analyses Sections 2.4, 2.5, et al.: Descriptions and results of other generic analyses and methodologies may reside in additional sections in this chapter. The efficiency of presenting additional generic analyses and methodologies will be driven by the nature of the facility and the LMP-based affirmative safety case. These sections are design specific and are provided at the discretion of the applicant. They may be subdivided as appropriate for the topic. Potential examples include:

  • Civil and structural analysis
  • Piping analysis
  • Electrical load analysis
  • Stress analysis
  • Criticality analysis
  • Thermal-hydraulic analysis
  • Environmental qualification analysis
  • Dispersion modeling These analyses should be pertinent to the LMP-based affirmative safety case (i.e., to SR SSCs and/or associated special treatments). The applicant should describe the analytical methodology and the key inputs and assumptions used. The applicant should address the applicability of the analytical methodology to the specific analysis, including a discussion of supporting data.

Details of the analyses should be in design records and available for regulatory audits.

In addition, results of testing performed to support an application for a first-of-a-kind plant design, pursuant to the requirements of 10 CFR 50.43(e) to provide data on the performance of Commented [A35]: Where are the other 50.43(e) plant safety features for the development and/or validation of analytical models, should be requirements addressed (combination of test programs, experience, analysis)? Are these other considerations summarized briefly Chapter 2. Guidance on documentation of such a test program is beyond the considered to be integral to the approach, and therefore scope of this document and is included in NRCs ARCAP guidance. dont need to be specifically called out?

Commented [A36]: Need to ensure this is addressed in Two-Step Licensing (CP Content) ARCAP.

Commented [A37]: Question: Where in ARCAP is this Section 2.4 should mirror the COL guidance but will reflect the preliminary nature of the design 50.43(e) testing guidance captured?

information as appropriate. The applicant should describe the technical areas that require Commented [A38]: Why is this beyond the scope of research and development to confirm the assumptions and methodologies used to present the TICAP? Isnt documentation of the program required to adequacy of the design. support conclusions on the adequacy of the analytical models, which the previous sentence says are to be summarized in Chapter 2?

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 3 LICENSING BASIS EVENTS This chapter documents the selection and evaluation of LBEs that serve as the foundation for the safety case. Because the NEI 18-04 methodology has been endorsed for use by NRC in RG 1.233, the scope and content of the SAR are focused on presenting the results and not presenting the details of the process. Commented [A39]: It may be useful here or elsewhere (Chapter 4 perhaps) to mention if an alternative acceptance The method for identification and evaluation of the LBEs is described in NEI 18-04, Section 3.2 criteria or design goal is being pursued such that the results of the LBEs would justify things such as alternative siting and in the text that accompanies Figure 3-2. The LBEs evolve through design and licensing, as criteria (see SECY-20-0045) or alternative emergency discussed in NEI 18-04, Section 3.2.3. At the time the SAR is submitted to NRC for review as planning zones. Although this was not discussed in detail within NEI 18-04, a specific plant SAR would need to address part of the advanced reactor combined license application, the process will have been this topic. This can likely be done with just a couple completed. The SAR documents the results, not the process. sentences here or elsewhere if there is a more appropriate place.

In addition to conventional single-unit reactor events, the LBEs include event sequence families that may involve multiple reactor units and non-reactor radionuclide sources. The initiating events associated with LBEs include those caused by internal and external hazards reflected in the PRA, as well as those hazards addressed deterministically. As discussed in Section 6.1, the Commented [A40]: Suggest that this text be relocated to LBEs derived from the PRA are augmented by the selection of DBEHLs and the requirement that the non-italic section.

SR SSCs are protected against these DBEHLs.

3.1 Licensing Basis Event Selection Methodology This guidance assumes that the applicant followed the NEI 18-04 methodology for the selection and evaluation of LBEs. If the applicant did not use parts of the methodology or used an alternative method, the following statement should be added:

The following deviations from the approved methodology were taken.

The applicant should go on to list each deviation, describe any alternative method, and provide justification for the approach employed.

The NEI 18-04 methodology affords some flexibility in implementation, so the specific manner in which the methodology was applied should be described as necessary to provide an adequate description of the grouping of event sequence families that are used to define the AOOs, DBEs, and BDBEs. The role of the PRA and resulting risk insights that were used to confirm the completeness and classification of the LBEs should be summarized as needed to gain an understanding of how the LBEs are defined. It is not necessary to repeat aspects of the methodology already covered in NEI 18-04, but rather to point out the specifics of how the methodology was applied within the range of options specified in NEI 18-04. Details of the analyses should be contained in design calculations and retained in the design records.

Safe, stable end states are a key element of the reactor safety case and should be covered in this section. In LWR safety analysis reports, it is generally understood how safe, stable end states are defined in such terms as preventing core damage, maintaining containment integrity, achieving cold shutdown, etc. However, for advanced non-LWRs, the safe, stable end states, including 31

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case success criteria that are needed to achieve them, need to be defined for the specific technology and design. The plant parameters used to define the end states, core reactivity, reactor power, fuel temperatures, etc., should be identified.

3.2 LBE Summary 3.2.1 Summary Evaluation of AOOs, DBEs, and BDBEs In this section, a summary of the evaluation of LBEs derived from the PRA is presented. This summary should include:

  • Identification of the radionuclide sources associated with each of the LBEs
  • A plot of the frequencies, consequences, and uncertainties of these LBEs with comparison to the NEI 18-04 Frequency-Consequence Target in Figure 3-1 of NEI 18-04
  • Identification of all risk significant LBEs as defined in NEI 18-04
  • Identification of any high consequence BDBEs as defined in NEI 18-04, i.e., those BDBEs with exclusion area boundary doses greater than 25 rem
  • Definition of the reactor-specific safe, stable end states, described previously and used to establish the success criteria for the safety functions modeled in the PRA, referred to in NEI 18-04 as PRA Safety Functions and reflected in the LBE descriptions The word descriptions of the LBE should be in sufficient detail to indicate the PRA Safety Functions involved in the prevention and mitigation of the LBEs. These PRA Safety Functions are performed by specific SSCs and are used to determine the SSC safety classifications in Chapter 5. Table 3-1 is an example table that was derived from Table 5-5 of the LMP LBE report, which was produced in support of NEI 18-04.7 7

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors, Selection and Evaluation of Licensing Basis Events, Rev 1, Idaho National Laboratory, March 1, 2020.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Table 3-1. Example Summary Table of AOOs, DBEs, and BDBEs LBE LBE Description Designation Anticipated Operational Occurrences Transient initiating event with successful reactor trip and successful cooling through balance-of-plant AOO-1 (BOP) systems; no fuel damage Transient initiating event with successful reactor trip, failure of BOP cooling systems, but successful AOO-2 cooling with the forced-air steam generator auxiliary cooling system (SGACS); no fuel damage Design Basis Events Transient initiating event with failure of active decay heat removal, but success of passive air-cooling DBE-1 with the reactor vessel auxiliary cooling system (RVACS); no fuel damage Beyond Design Basis Events Spurious control rod withdrawal with successful reactor trip, failures of decay heat removal through BDBE-1 both BOP systems and the SGACS, but successful passive air-cooling with the RVACS; no fuel damage Steam generator tube rupture event with successful reactor trip and suppression of sodium-water BDBE-2 reaction, failure of the SGACS, but successful passive air-cooling with the RVACS; no fuel damage The plots of the LBE frequencies and consequences should be made with points corresponding to the mean estimates of frequency and consequence with uncertainty bars indicating the 5th and 95th percentiles of the quantified uncertainty distributions in both frequency and dose for each LBE. LBEs with no release and hence no dose should be plotted on the Y-axis. To accommodate LBEs involving one or more than one reactor or non-reactor source, LBE frequencies are expressed on a per plant-year basis, where plant refers to the facility being licensed.

Identification of risk significant LBEs is based on the criteria in NEI 18-04. The plot discussed above should identify the risk-significant zone for LBEs.

Note: If any part of the plotted uncertainty bands for frequency or dose falls inside the risk significant zone in Figure 3-4 of NEI 18-04, the LBE is regarded as risk significant.

3.2.2 Summary Evaluation of DBAs In this section, a summary of the evaluation of DBAs is presented. This summary should include a reference to Section 3.6 for the details of the evaluation of each DBA as well as the following:

  • A table that shows the mapping of DBEs into DBAs with brief word descriptions of the DBEs and DBAs
  • A table that shows the dose consequences of the DBAs for comparison against the 25 rem criterion derived from 10 CFR 50.34 (Some DBAs may have no releases and, therefore, no doses.)

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Table 3-2 shows an example combining all of the information in one table. It was derived from Table 5-7 of the LMP LBE report.8 The dose values are illustrative only.

Table 3-2. Example Summary Table of DBEs and DBAs DBA 30-Day EAB DBE DBE Description DBA DBA Description Dose (rem)

Spurious control rod withdrawal with pre-existing rod-stop error and failure of BOP DBE-1 cooling; reactor protection system (RPS) shuts down reactor and active SGACS removes decay heat involving one reactor Spurious control rod withdrawal with pre-Spurious control rod withdrawal existing rod-stop error and failure of BOP with pre-existing rod-stop error cooling and SGACS; RPS shuts down reactor, DBE-2 and failure of BOP cooling and the EM pumps trip, and passive RVACS forced SGACS cooling; RPS shuts removes decay heat, supplemented by passive DBA-1 down the reactor and passive 0.0 mode of SGACS involving one reactor RVACS removes decay heat, Spurious control rod withdrawal with pre-including the extra power existing rod-stop error and failure of balance of generated during the transient plant (BOP) cooling and Steam Generator overpower involving one reactor Auxiliary Cooling System (SGACS); Reactor DBE-3 Protection System (RPS) shuts down reactor, the electromagnetic (EM) pumps trip, and passive Reactor Vessel Auxiliary Cooling System (RVACS) removes decay heat involving one reactor Steam generator tube rupture with failure of sodium-water Steam generator tube rupture is detected and reaction detection and suppressed by sodium-water reaction suppression equipment, which DBE-4 detection equipment, RPS shuts down the DBA-2 disables all cooling modes through 0.1 reactor, and active SGACS removes decay heat the intermediate loop; RPS shuts involving one reactor down the reactor and passive RVACS removes decay heat involving one reactor A general transient with failure of BOP cooling and forced SGACS; RPS shuts down the reactor, DBE-5 the EM pumps trip, and passive RVACS removes decay heat supplemented by passive A general transient with failure of mode of SGACS involving both reactors BOP cooling and forced SGACS; A general transient with failure of BOP cooling RPS shuts down the reactor, the DBA-3 0.0 and all modes of SGACS; RPS shuts down the EM pumps trip and passive RVACS DBA-6 reactor, the EM pumps trip, and passive RVACS removes decay heat involving both removes decay heat involving both reactors reactors A general transient with failure of the DBE-7 intermediate sodium coolant loop; RPS shuts down the reactor, the EM pumps trip, and 8

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors, Selection and Evaluation of Licensing Basis Events, Rev 1, Idaho National Laboratory, March 1, 2020.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case DBA 30-Day EAB DBE DBE Description DBA DBA Description Dose (rem) passive RVACS removes decay heat involving both reactors A plant-centered loss of offsite power with failure of backup power to forced SGACS; RPS DBE-8 shuts down the reactor and passive RVACS removes decay heat involving both reactors A major hurricane causes both a A major hurricane causes both a loss of offsite loss of offsite power and an off-power and an off-normal condition for RVACS; normal condition for RVACS; RPS DBE-9 RPS shuts down the reactor and passive RVACS DBA-4 shuts down the reactor and 0.0 removes decay heat under storm conditions passive RVACS removes decay involving both reactors heat under storm conditions involving both reactors 3.3 Anticipated Operational Occurrences This section identifies and describes the plant AOOs that are informed by the PRA event sequence families. AOOs are anticipated event sequences expected to occur one or more times during the life of a nuclear power plant, which may include one or more reactors. Event sequences with mean frequencies of 1x10-2/plant-year and greater are classified as AOOs.

AOOs take into account the expected response of all SSCs within the plant, regardless of safety classification.

3.3.1 AOO-1 For each AOO, the following information should be provided:

  • Narrative of the LBE, including the definition of the initial plant conditions and plant operating state, radionuclide source (including whether it involves multiple reactors and sources), initiating events covered in the family, characterization of the responses of SSCs and operator actions that perform PRA safety functions, credit taken for the functioning of normally operating plant systems, identification of whether or not there is a release, and definition of the safe, stable end state. The AOO description should also identify whether or not a radioactive material barrier (e.g., fuel clad) was breached during the event. Commented [A41]: I think we want to know whether or not a fuel barrier failed during an AOO even though there may not be a release to the public.

The following information should be provided for any AOO with a release. The applicant may elect to provide some or all of the following information for other AOOs:

  • Identification of the reactors and non-reactor sources involved in the AOO
  • Plots of the responses of key plant parameters needed to characterize the plant response and the mechanistic source term, if there is a release. Discussion of the characteristics of fission product releases from the proposed site to the environment including the rates of 35

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case fission product release and the chemical forms of fission products released to the environment.

  • A description of the dispersion model (if not addressed in Chapter 2) and input parameters (e.g., meteorology, off-site population distribution, EAB size) used in the analysis, and assumptions on location of individual members of the public.
  • Tables to describe the mechanistic source term if there is a release (or a reference to the source term description in Chapter 2)
  • The mean, 5th percentile, and 95th percentile values of the estimated frequency and dose Section B.1 of Appendix B provides an example AOO description.

3.3.2, 3.3.3, et al.: The remainder of the AOOs are addressed.

3.4 Design Basis Events This section identifies and describes the plant DBEs that are informed by the PRA event sequence families. DBEs are infrequent event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactors or non-reactor sources, and by definition, are less likely than AOOs. Event sequences with mean frequencies of 1x10-4/plant-year to 1x10-2/plant-year are classified as DBEs. DBEs take into account the expected response (including successful and unsuccessful performance of the modeled PRA Safety Functions) of all SSCs within the plant regardless of safety classification. Note: If uncertainty bands for an AOO or BDBE LBE frequency failfall inside the DBE frequency range, such LBEs are evaluated using the NEI 18-04 rules for both LBE categories. Commented [A42]: This text should be guidance, not italic font.

3.4.1 DBE-1 For each DBE, the following information should be provided:

  • Narrative of the DBE, including the definition of the initial plant conditions and plant operating state, radionuclide source and whether it involves multiple reactors and sources, initiating events covered in the family, characterization of the responses of SSCs and operator actions that perform PRA safety functions, credit taken for the functioning of normally operating plant systems, identification of whether there as a release, a description of the dispersion model (if not addressed in Chapter 2) and input parameters (e.g., meteorology, off-site population distribution, EAB size) used in the analysis, assumptions on location of individual members of the public, the analysis method used, timing and duration of release, uncertainty/sensitivity analysis performed, margins to F-C curve, and definition of the safe, stable end state
  • The applicant should list the settings of all protection system functions that are used in the DBE evaluation. Typical protection system functions may include reactor trips, isolation valve closures, and emergency cooling system initiation. List the expected 36

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case limiting delay time for each protection system function and describe the acceptable methodology for determining uncertainties (e.g., from combined effect of calibration error, drift, instrumentation error) to be included in the establishment of the trip setpoints and allowable values specified in the plant technical specifications.

For the most limiting DBE that was used to map into each DBA (see Section 3.2.2, Bullet 1), the following information should be provided. This will enable a comparison of the realistic behavior of the plant (DBE) to the conservatively analyzed behavior (corresponding DBA). The applicant may elect to provide some or all of the following information for other DBEs:

  • Identification of the reactors and non-reactor sources involved in the DBE
  • Plots of the responses of key plant parameters needed to characterize the plant response and the mechanistic source term if there is a release
  • Characterization of the response of structures, systems, and components that perform PRA safety functions Commented [A43]: Dont need to include this here given
  • Discussion of relevant phenomena that may impact plant response and mechanistic that it is addressed in the paragraph above.

source terms

  • Identification of common-cause failures between reactors, if applicable, and the reactors and sources impacted
  • Tables to describe the mechanistic source term if there is a release (this may involve a reference to Chapter 2)
  • The mean, 5th, and 95th percentile values of the estimated frequency and dose Section B.2 of Appendix B provides an example DBE description.

3.4.2, 3.4.3, et al.: The remainder of the DBEs are addressed.

3.5 Beyond Design Basis Events This section identifies and describes the plant BDBEs that are informed by the PRA event sequence families. BDBEs are rare event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactors but are less likely than a DBE.

Event sequences with mean frequencies of 5x10-7/plant-year to 1x10-4/plant-year are classified as BDBEs. BDBEs take into account the expected response (including successful and Commented [A44]: While this is true, how are events unsuccessful performance of the modeled PRA Safety Functions) of all SSCs within the plant with a mean frequency below 5E-7 but an uncertainty extending above that value documented?

regardless of safety classification.

3.5.1 BDBE-1 For each BDBE, the following information should be provided:

Narrative of the BDBE including the definition of the initial plant conditions and plant operating state, radionuclide source including whether it involves multiple reactors and sources, initiating events covered in the family, the response of plant systems, including a 37

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case characterization of the responses of SSCs that perform PRA safetyfunctions, identification of whether there is a release,a description of the dispersion model andinput parameters (e.g., meteorology,off-sitepopulationdistribution,EABsize)usedin the analysis, assumptions on location of individual members of the public, the analysis method used, timing and duration of release, uncertainty/sensitivity analysis performed, margins to F-C curve, and definition of the safe, stable end state The information below should be provided for any high consequence BDBEs as well as other BDBEs to bound the risks associated with the collection of BDBEs. High consequence BDBEs are those with consequences that exceed 25 rem at the Exclusion Area Boundary for the 30-day period beginning at the onset of release. The set of BDBEs that bound the risks is that with the highest products of frequency times dose. The applicant may elect to provide some or all of the following information for other BDBEs:

  • Identification of the reactors and non-reactor sources involved in the BDBE
  • Plots of the responses of key plant parameters needed to characterize the plant response and the mechanistic source term, if there is a release
  • Characterization of the response of structures, systems, and components that perform PRA Safety Functions Commented [A45]: Dont need to include this here given
  • Discussion of relevant phenomena that may impact plant response and mechanistic that it is addressed in the paragraph above.

source terms

  • Identification of common-cause failures between reactors, if applicable, and the number of reactors impacted
  • Tables to describe the mechanistic source term if there is a release (This may involve a reference to Chapter 2.)
  • The mean, 5th percentile, and 95th percentile of the estimated frequency and dose Section B.3 of Appendix B provides an example BDBE description.

3.5.2, 3.5.3, et al.: The remainder of the BDBEs are addressed.

3.6 Design Basis Accidents This section identifies and describes the DBAs that are analyzed using conservative deterministic safety analysis. The conservative assumptions used in the DBA analysis are informed by the quantitative uncertainty analysis of consequences that was performed for the corresponding DBEs. In view of the fact that advanced non-LWRs will employ a diverse combination of inherent, passive, and active design features to perform the Required Safety Functions (RSFs) across layers of defense, and, taking into account the fact that the reactor safety design approach will be subjected to an evaluation of DID adequacy, the application of a single failure criterion is not deemed to be necessary.

SR SSCs which mitigate the DBAs are documented in Section 5.4.

This section identifies and describes the plant events that will be included in the licensing basis as DBAs. As established in the NEI 18-04 methodology, DBAs are to be derived from the DBEs 38

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Content for a Licensing Modernization Project-Based Affirmative Safety Case by conservatively assuming that only SR SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose criteria. DBAs are used to set design criteria and performance objectives for the design of SR SSCs. Appropriately conservative assumptions for the mechanistic source term and dispersion characteristics are also to be used.

3.6.1 DBA-1 For each DBA, the following information should be provided:

  • Narrative of the DBA including the definition of the initial plant conditions and plant operating state, radionuclide source and whether it involves multiple reactors and sources, initiating events covered in the family, the response of plant systems, identification of whether there as a release, a description of the dispersion model (if not described in Chapter 2) and input parameters (e.g., meteorology, off-site population distribution, EAB size) used in the analysis, assumptions on location of individual members of the public, the analysis method used, timing and duration of release, uncertainty/sensitivity analysis performed, and definition of the safe, stable end state
  • Plots of the plant response to key plant parameters needed to characterize the plant response and the mechanistic source term, if there is a release
  • Characterization of the response of SR SSCs
  • Evaluation of relevant phenomena that may impact plant response and mechanistic source terms
  • Tables to describe the mechanistic source term if there is a release (This may involve a reference to Chapter 2.)
  • Description of the conservative calculation used to demonstrate that the 25 rem total effective dose equivalent (TEDE) dose limit in 10 CFR 50.34 is met
  • An acceptable approach is to use the 95th percentile dose from the corresponding limiting DBE mapped into the DBA Deterministic calculations of DBA sequences are typically performed using one or more computer codes that constitute an analytical model of the plant response. Examples for light water reactors include the RETRAN and RELAP computer codes. If there is a release of radionuclides, the mechanistic source term discussed in Chapter 2 would also be involved. The analytical methodology may be described in Section 2.3 if the methodology is applicable to multiple DBAs. If not, the applicant should address the analytical methodology in this subsection, following the same guidance provided in Section 2.3.

Section B.4 of Appendix B provides an example DBA description.

9 Regulatory Guide 1.203, Transient and Accident Analysis Methods, U.S. Nuclear Regulatory Commission, December 2005.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 3.6.2, 3.6.3, et al.: The remainder of the DBAs are identified and described.

Two-Step Licensing (CP Content)

For a CP application, Chapter 3 should mirror the COL guidance but will reflect the preliminary nature of the design information. The PSAR should describe the methodology to be used in determining the initial set of LBEs and specifically address the conservative DBA calculation used to demonstrate that the 25 rem TEDE dose limit in 10 CFR 50.34 is met to support the site suitability requirements. The structure of the chapter and sections should follow the structure for the COL guidance. The discussions should be sufficiently robust so the reader can clearly see how the methodology will lead to a final set of LBEs to be used in developing the final design, safety margins, operational program content, and FSAR content. The discussion should clearly describe the role of the PRA in determining the initial set of DBEs. The PRA methodology described in Chapter 2 should be used to determine the preliminary assessments of the Licensing Basis Events, as described in the COL guidance for Sections 3.3 through 3.6 above. (Note that ASME/ANS RA-S-1.4-2021 includes guidance on the performance of PRAs at various design stages.)

This section should describe how the performance of fission product barriers credited to prevent and/or inhibit the release of radionuclides are or will be supported by existing or planned experimental data that cover the needed range of applicability.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 4 INTEGRATED EVALUATIONS This guidance assumes that the applicant followed the NEI 18-04 methodology for the overall plant risk performance summary and the incorporation of Defense-in-Depth. If the applicant did not use parts of the methodology or used an alternative method, the following statement should be added:

The following deviations from the approved methodology were taken.

The applicant is to list each deviation, describe any alternative method, and provide justification for the approach employed.

4.1 Overall Plant Risk Performance Summary Commented [A46]: For Sections 4.1.2 and 4.1.3 guidelines should be added to ensure consistency in the individual risk calculations. Some aspects of the calculations The overall plant risk summary presents results of the PRA that reflect the cumulative risk to the will be plant-specific (e.g., meteorology, source term, EP public. The PRA is discussed in Chapter 2. assumptions, etc.) but some should be plant and site independent (e.g., source of exposure [cloud shine, ground This section describes the integrated plant performance for the three cumulative plant shine, inhalation], risk coefficients, inhalation rate, medical treatment assumptions, etc.). Such guidelines will help performance metrics contained in NEI 18-04 Section 3.2.2, Task 7b for risk to the public from ensure a common approach to the overall risk calculations.

radiation. AOOs, DBEs, and BDBEs are included in the evaluation of overall risk. DBAs are addressed deterministically and not included in the overall integrated risk evaluation. Commented [A47]: For each plant performance metric analysis discussed below, the SAR should not only address 4.1.1 Exclusion Area Boundary Dose results and margins, but also address the following where different from the analysis performed under section 3:

  • The site parameters (e.g. meteorology, off-site This section will address the cumulative risk target that the total frequency of exceeding an population distribution, EAB size) used in the analysis, exclusion area boundary dose of 100 mrem from all LBEs should not exceed one per plant-year.
  • Assumptions on location of individual members of the public, This section should provide the predicted total risk from the entire range of LBEs from higher
  • The analysis method used, frequency, lower consequences to lower frequency, higher consequences. This result should be
  • Key assumptions (e.g., emergency preparedness based on mean values without uncertainties. The margin between the target of 100 mrem and the measures, source terms, timing and duration of release, predicted plant performance should be described. credit for medical treatment, early and latent fatality risk coefficients) used in the analysis,
  • Modes of operation (full power, low power & shutdown, 4.1.2 EAB Boundary Early Fatality Risk refueling) considered in the analysis.
  • How multiple units on the site were considered, This section addresses the cumulative average individual risk target that early fatality risk within Uncertainty/sensitivity analysis performed.

1 mile of the exclusion area boundary should not exceed 5x10-7/plant-year. This section should Commented [A48]: Wouldnt it be more clear if risk is provide the predicted cumulative average individual risk from the entire range of LBEs from replaced with frequency of LBEs with a dose at the EAB greater than 100 mrem? Risk is too loose a term.

higher frequency, lower consequences to lower frequency, higher consequences. This result should be based on mean values without uncertainties. The margin between the target and the Commented [A49]: proposed change predicted plant performance should be described.

4.1.3 Latent Cancer Risk This section addresses the cumulative average individual risk target that the average individual risk of latent cancer fatality within 10 miles of the EAB should not exceed 2x10-6/plant-year.

This section should provide the predicted cumulative average individual risk from the entire 41

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case range of LBEs from higher frequency, lower consequences to lower frequency, higher consequences. This result should be based on mean values without uncertainties. The margin between the target and the predicted plant performance should be described.

Two-Step Licensing (CP Content)

For a CP application, Section 4.1 should provide a preliminary description of the integrated plant performance for the three cumulative plant performance metrics contained in NEI 18-04 Section 3.2.2, Task 7b for risk to the public from radiation. The PRA methodology described in Chapter 2 should be used in the dose and risk estimates addressed in Section 4.1. The applicant should identify limitations in the scope and level of detail of the CP-stage PRA that are to be addressed in the OL application.

4.2 Defense-in-Depth Commented [A50]: Why is all of the text in this section italic? This text should be part of the guidance. If it says This section should describe then it should be guidance.

This section should describe the baseline level of DID provided by the facility. This baseline represents the result of recurring evaluation of plant capability and programmatic capability associated with design and PRA update cycles leading to the application content. It identifies risk-significant vulnerabilities where compensatory actions made a practical, significant improvement to the LBE risk profiles or risk-significant reductions in the level of uncertainty in characterizing the LBE frequencies and consequences. The baseline DID adequacy evaluation results are to be documented in sufficient detail to assure that future changes to physical, functional, operational, or programmatic features of the facility can be effectively evaluated for their potential for reduction of DID before proceeding.

The following sections provide a summary of results of Plant Capability DID, Programmatic DID, and the Integrated Assessment of DID, respectively. They reflect the results for the topics listed in NEI 18-04 Table 5-1, Risk-Informed Evaluation of DID Adequacy, including:

  • Evaluation of design attributes for DID
  • Input to the identification of safety-significant SSCs
  • Input to the selection of SR SSCs
  • Evaluation of roles of SSCs in the prevention and mitigation of LBEs
  • Evaluation of the LBEs to assure adequate functional independence of each layer of defense
  • Evaluation of single features that have a high level of risk importance to assure no overdependence on that feature and appropriate special treatment to provide greater assurance of performance
  • Evaluation of risk-significant uncertainties
  • Input to SSC performance requirements for reliability and capability of risk-significant prevention and mitigation functions
  • Input to SSC performance and special treatment requirements
  • Integrated evaluation of the plant capability DID
  • Integrated evaluation of programmatic measures for DID 42

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Note that the above information is provided for background, and there is no requirement to address each topic in the SAR material in this chapter. Portions of the DID adequacy evaluation Commented [A51]: This statement appears to over-ride may be provided in Chapters 3, 5, 6, 7, and 8 and are not repeated in this chapter. Section 4.2 the text above. Suggest deleting this sentence.

focuses primarily on DID attributes that are cross-cutting in nature, i.e., examine the integrated risk profile, common features in multiple LBEs, and the adequacy of the collective layers of defense to address all risk-significant LBEs. The completeness and sufficiency of the DID adequacy evaluation are summarized in the Integrated DID Summary (Section 4.2.3).

Because the NEI 18-04 methodology has been endorsed for use by NRC in RG 1.233, the scope and content of the SAR are focused on presenting the results and not presenting the details of the process. The summary focus is on safety-significant topics, LBEs, SSCs, and operator actions Commented [A52]: The process for identifying DID is not that receive special treatments as described in NEI 18-04. The summary need not address DID so formulaic that a reader could derive relevant methodological inputs from a provide set of results.

evaluations that did not identify further provisions for DID. The content of the DID Summary Relevant details from the process should be identified if provides the foundation for the DID adequacy evaluation baseline as described in NEI 18-04 they were key to augmenting or excluding aspects of the Section 5.9.5. Evidence of the complete DID evaluation should be retained in design records. plant response.

4.2.1 Plant Capability Summary The purpose of this section is to provide a description of the plant capability and the layers of defense that address cross-cutting topics in the overall achievement of an acceptable level of DID. This section of the SAR need not describe how the design meets all the guidelines for plant Commented [A53]: Why shouldnt the SAR describe how capability attributes provided in NEI 18-04 Table 5-2, Guidelines for Establishing the Adequacy the design meets all of the guidelines in NEI 18-04, Table 5-2? What this says is that DID is not important enough to of Overall Plant Capability Defense-in-Depth. The applicable information for individual LBEs document its basis. This contradicts the statement in is located in Chapter 3 - Licensing Bases Events, Chapter 5 - Safety Functions, Design Criteria, Section 1.3.3 that states DID is a key element of the LMP-and SSC Safety Classification, Chapter 6 - SR SSC Criteria and Capabilities, and Chapter 7 - based affirmative safety case and the demonstration of reasonable assurance of adequate protection of public NSRST SSC Criteria and Capabilities. health and safety. If DID is a key element of the safety case, why shouldnt it be thoroughly documented in the SAR?

Plant capability DID attributes are listed in NEI 18-04 Table 5-3: initiating event and event Commented [A54]: Earlier NRC/INL comments stated sequence completeness, layers of defense, functional reliability, and prevention and mitigation that the SAR should describe how the design meets the balance. As outlined in NEI 18-04 Table 5-9, the qualitative evaluation should address the guidelines in Table 5-2.

evaluation of margin adequacy, multiple protective measures, and prevention and mitigation balance across layers of defense and the physical categories of functional reliability and over-reliance on any single feature. Commented [A55]: This text should be part of the guidance and not italic font.

The application should state affirmatively that the guidelines for plant capability attributes provided in NEI 18-04 Table 5-2 have been evaluated and confirmed. For SSCs that are relied upon to perform DID prevention and mitigation functions for risk-significant LBEs, describe the set of requirements related to the performance, reliability, and availability of the SSC functions that are relied upon to ensure the accomplishment of their tasks, as defined by the PRA or deterministic analysis. This description should include how that capability is ensured through testing, maintenance, inspection and performance monitoring. If this information is provided in other sections it need not be repeated here. Separate discussions of plant capabilities added as a result of plant capability attribute evaluations should be provided in this section.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case During the DID adequacy evaluation process, safety-significant SSC functions may have been deemed necessary for DID adequacy. Where so, this information should be documented in tabular form in a manner that is traceable to the LBEs in Chapter 3. The information should include the rationale for the selection of LBE SSCs for NSRST classification and guide the selection of NSRST SSC performance criteria and special treatments as documented in Chapter 7.

4.2.1.1 LBE Margin This section provides the baseline margins established between the frequencies and consequences of individual risk-significant LBEs and the F-C Target. These margins are established for the risk-significant LBEs within each of the three LBE categories: AOOs, DBEs, and BDBEs. A tabular format example for mean values is shown in Table 4-1, based on Section 2.9.1 of the LMP DID report.10 Both mean and 95th percentile values should be provided.

Table 4-1. Example Table of LBE Risk Margins Mean F-C Target Freq. Mean F-C Target Mean LBE Mean Freq./

Name Dose at LBE Frequency Dose at LBE Dose Category plant-yr (rem) Dose/plant-yr[a] Margin[b] Freq. (rem) [c] Margin[d]

AOO AOO-5 4.00E-02 2.50E-04 4.00E+02 1.0E+04 1.00E+00 4.0E+03 DBE DBE-10 1.00E-02 2.00E-03 6.00E+01 6.0E+03 1.00E+00 5.0E+02 BDBE BDBE-2 3.00E-06 4.00E-03 2.50E+01 8.3E+06 2.50E+02 6.2E+04 Notes:

[a] Frequency value measured at the LBE mean dose level from the F-C Target

[b] Ratio of the frequency in Note [a] to the LBE mean frequency (Mean Frequency Margin)

[c] Dose value measured at the LBE mean frequency from the F-C Target

[d] Ratio of the dose in Note [c] to the LBE mean dose (Mean Dose Margin) 4.2.1.2 Layers of Defense Evaluation The DID evaluation in this section includes confirmation that plant capabilities for DID are sufficient to prevent and mitigate each risk-significant LBE across the available layers of defense; confirmation that a balance between event prevention and mitigation is reflected across the layers of defense for risk-significant LBEs; and confirmation of sufficient independence between layers of defense for risk-significant LBEs. For multi-reactor plants, layers of defense supporting more than one reactor should be included in the DID adequacy evaluation. Commented [A56]: This text should be guidance and not italic font.

For each LBE, each qualitative guideline in NEI 18-04 Table 5-2 should be addressed and any deviations from the stated criteria addressed. The applicant should provide a summary identification of the layers of defense for each risk-significant LBE and describe the extent of 10 Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Risk-Informed and Performance-Based Evaluation of Defense-in-Depth Adequacy, Rev 0, Idaho National Laboratory, August 2019. (It should be noted that there is a subsequent revision to this report; however, that version is not readily accessible on the internet.)

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Content for a Licensing Modernization Project-Based Affirmative Safety Case independence between different LBE layers of defense. The LBE sequences include the contributions from common-cause failures in accordance with the PRA standard. Any risk-significant common-cause failures that were not eliminated by design should be identified along with identification of the layers of defense impacted (i.e., layer independence impacts or reactor independence impacts) for risk-significant LBEs and the programmatic special treatments applied. The applicant should include an affirmative statement that there is no overreliance on a Commented [A57]: The applicant should be required to single layer of defense for any risk-significant LBE. state the acceptance criteria for this determination and explain how they are met, not just provide a statement.

Would the basis for the affirmative or conclusion statement An example presentation of the layers of defense evaluation is provided below. on no overreliance on a single layer be demonstrated by the applicant addressing Table 5-2 in NEI 18-04 Layer 1Prevent off-normal operation and AOOs Summarize non-safety-related provisions for overall plant reliability and availability that achieve a frequency of plant transients consistent with owner-operator performance objectives.

Layer 2Control abnormal operation, detect failures, and prevent DBEs Table 4-2. Example Table of AOOs Functions to Maintain Frequency of all Functions to Minimize Frequency of AOO DBEs < 10-2/plant-year Challenges to SR SSCs AOO-1 AOO-2 AOO-N Layer 3Control DBEs within the analyzed design basis conditions and prevent risk-significant BDBEs Table 4-3. Example Table of DBEs Functions to Maintain Frequency of all Any single design or operational DBE BDBEs < 10-4/plant-year feature relied upon?

DBE-1 DBE-2 DBE-N Layers 4 and 5Control severe plant conditions, mitigate consequences of BDBEs, deploy adequate offsite protective actions and prevent adverse impact on public health and safety Table 4-4. Example Table of BDBEs Functions to Maintain Individual Risks Any single barrier or plant feature BDBE

< QHOs* relied upon?

BDBE-1 BDBE-2 BDBE-N

  • QHO = Quantitative health objective 45

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 4.2.1.3 Single Feature Reliance This section should affirmatively state that the evaluation of the dependence on a single feature across the layers of defense was performed for all risk-significant LBEs and DBAs. A brief discussion of the method and acceptance criteria used to evaluate this DID topic should be Commented [A58]: proposed change provided. The summary should identify any special treatments added to provide assurance against over-reliance on any single feature across multiple layers of defense for any risk- Commented [A59]: can this just be reliance? over-significant LBE. The details of this evaluation should be retained in design records. reliance implies that there are scenarios that this would be acceptable 4.2.1.4 Prevention-Mitigation Balance The determination of prevention-mitigation balance across layers of defense is described in NEI 18-04 Section 5.7. Additional discussion can also be found in the LMP supporting report, Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Risk-Informed and Performance-Based Evaluation of Defense in Depth Adequacy, which was previously referenced.

This discussion should affirmatively state that the evaluation of prevention-mitigation balance across the layers of defense was performed for all risk-significant LBEs and DBAs. A brief Commented [A60]: This language is used in a number of discussion of the method and acceptance criteria used to evaluate this DID topic should be cases throughout the document. In some cases, a statement supported by background documentation could be provided. The details of this evaluation should be retained in design records. appropriate; in others, such as this one, it isn't clear to the staff how an affirmative statement adequately captures in 4.2.2 Programmatic DID Summary the licensing basis how the affirmative statement is supported. As an example, in this case, a reference to the preceding tables and identification of which design features Programmatic DID attributes are listed in NEI 18-04 Table 5-5: quality/reliability, are relied on for prevention versus those for mitigation compensation for uncertainties, and offsite response. might be appropriate.

Commented [A61]: proposed change Programmatic DID should be used to provide the basis for defining special treatment requirements to ensure there is reasonable assurance that the predicted performance of SSCs can be achieved throughout the life of the plant. This section should provide any additional special treatments identified from the integrated DID adequacy evaluation in Chapter 4 and complement the discussion of special treatment programs selected for safety-significant SSCs described in Chapters 6 and 7. The application should identify DID considerations in NEI 18-04 Table 5-6 that led to additional special treatments. Special treatments described in NEI 18-04 Table 5-7 should be considered, although the SAR does not need to address items that are not applicable. Commented [A62]: This text should be guidance not italic font The application should state affirmatively that the guidelines for programmatic capability attributes provided in NEI 18-04 Table 5-6 have been evaluated and included in the design development. . The application should describe how the design incorporates the programmatic capability attributes provided in NEI 18-04 Table 5-6 to provide adequate assurance that the risk, reliability, and performance targets will be met and maintained throughout the life of the plant with adequate consideration of sources of significant uncertainties. This description should support the discussion of special treatment programs selected for safety-significant SSCs described in Chapters 6 and 7. Separate discussions of additional programmatic additions or 46

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Content for a Licensing Modernization Project-Based Affirmative Safety Case changes as a result of the DID programmatic attribute evaluations, including identification of the safety-significant LBEs leading to additional DID programmatic actions and resulting safety-significant compensatory actions, should be provided in this section along with their bases. Commented [A63]: proposed change Summary information should be provided for the individual DID evaluation results that led to additional changes to the protective measures required for adequate programmatic DID.

4.2.2.1 Evaluation of Significant Uncertainties PRA quantification of uncertainties as required by the non-LWR PRA standard and other means such as sensitivity analysis are tools that can be used to determine the impact of uncertainties.

The results of these analysis should be documented in design records.

As noted in Chapter 3 Section 3.2.1, if any part of the plotted LBE uncertainty bands for frequency or dose falls inside the risk significant zone in NEI 18-04 Figure 3-4, the LBE is regarded as risk-significant. The purpose of this DID section is to summarize the impact of any specific source of uncertainty identified during the Integrated Decision-Making Process (IDP) that determined specific additional DID protective measures to manage or further mitigate the quantitative or qualitative impact of a source of uncertainty. This is described in NEI 18-04 Section 5.9.4 for SSCs required for DID adequacy. Commented [A64]: This text should be guidance and not italic font The application should state affirmatively that the guidelines for evaluation of significant uncertainties or assumptions in the PRA were included in the IDP. This section should identify significant uncertainties or assumptions that were decided by the IDP to warrant supplemental special treatments associated with risk-significant LBEs (and LBEs that are not-risk significant but with uncertainty bands approaching the risk-significant level).

The action types taken for any LBE uncertainty in this part of the DID evaluation should be categorically described and cross-referenced to specific SSC requirements in Chapters 6 or 7, as appropriate.

4.2.2.2 Programs Required for SR SSC Performance Monitoring Chapter 6 should identify the plant-specific programs used to perform monitoring of SR SSCs and to assure human performance and operational controls for risk-significant functions. The requirements include consideration of DID. This section should summarize additions to or modification of the programmatic controls provided in Chapter 6 to account for and manage risk-significant uncertainties as a result of the DID evaluation. The description should be sufficient to identify the DID objectives requiring those actions.

4.2.2.3 Programs Required for NSRST SSC Performance Monitoring Chapter 7 should identify the plant-specific programs used for performance monitoring of NSRST SSCs and to assure human performance and operational controls for safety-significant functions. The requirements include consideration of DID and uncertainty. This section should 47

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Content for a Licensing Modernization Project-Based Affirmative Safety Case summarize additions to or modification of the programmatic controls provided in Chapter 7 to account for and manage safety-significant uncertainties as a result of the DID evaluation. The description should be sufficient to identify the DID objectives requiring those actions.

4.2.3 Integrated DID Evaluation Commented [A65]: It would seem appropriate for this section to require a summary of how the attributes in NEI The purpose of the integrated DID evaluation action is to look at the sum of the information and 18-04, Section 5.9.3, IDP Actions to Establish DID Adequacy were evaluated and determined adequate.

insights provided in the plant capability and programmatic action activities that address the Section 5.9.3 provides the most comprehensive list of what integrated decision-making attributes, listed in NEI 18-04 Section 5.9.2 (Table 5-8: use of the determines DID adequacy and why it is ignored in the TICAP guidance document is not clear. Although difficult to risk-triplet outside of the PRA, state of knowledge adequacy, uncertainty management, and summarize the deliberations of expert panels such as the action refinements. They are discussed in NEI 18-04 Section 5.9.4. IDPP, how would the guidance support documenting the process.

The application should summarize how the integrated DID process was applied in evaluating the overall adequacy of DID. The description should address how each of the decision guidelines listed in NEI 18-04, Section 5.9.3, was evaluated and the basis for an affirmative response. The criteria used in making the decisions (e.g., risk margins are sufficient, prevention/mitigation balance is sufficient, etc.) should be provided. If quantitative measures were used as part of the criteria, they should be provided. A description of how the results of the integrated DID process are documented and available for future DID decision-making and operations support should also be provided.

In order to support the DID baseline development, the applicant should identify (i) additional actions taken as a result of the integrated DID evaluations, (ii) the LBEs leading to those actions, and (iii) the plant or program features addressed by the actions. The applicant should also provide a brief summary of the rationale for the actions.

4.2.4 Evaluation and Incorporation of Changes to Defense in Depth The change control process should be described addressing how the baseline DID evaluation will be re-evaluated, based on proposed changes, to determine which programmatic or plant capability attributes have been affected for each layer of defense. Changes that impact the definition and evaluation of LBEs, safety classification of SSCs, or risk significance of LBEs or SSCs must be assessed. This section should also describe how any changes to the baseline DID evaluation will be documented and implemented.

The DID CP discussion should be plant capability-centric (Section 4.2.1). While not all of the plant capability DID attributes can be fully addressed at the CP stage, qualitative performance-based objectives for DID may be useful in establishing performance boundaries for FSAR results. It will not be practical to address programmatic DID (Section 4.2.2) and the integrated evaluation of DID adequacy (Section 4.2.3) in the PSAR, and those areas should be reserved for the FSAR developed as part of the OL application unless fundamental to the CP affirmative safety case envelope. Provide necessary commitments to establish DID adequacy per RG 1.233 at the OL stage. Commented [A66]: proposed change 48

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 5 SAFETY FUNCTIONS, DESIGN CRITERIA, AND SSC SAFETY CLASSIFICATION This chapter documents the Required Safety Functions and Required Functional Design Criteria (RFDC), Principal Design Criteria (PDC), Safety Classification of SR and NSRST SSCs, and the Complementary Design Criteria. Because the NEI 18-04 methodology has been endorsed for use by NRC in RG 1.233, the scope and content of the SAR are focused on presenting the results and not presenting the details of the process.

This guidance assumes that the applicant followed the NEI 18-04 methodology for the development of RSFs, RFDC, and safety classification of SR and NSRST SSCs. If the applicant did not use parts of the methodology or used an alternative method, the following statement should be added:

The following deviations from the approved methodology were taken.

The applicant should go on to list each deviation, describe any alternative method, and provide justification for the approach employed.

5.1 Safety Classification of SSCs The NEI 18-04 methodology affords some flexibility, so the specific manner in which the classification approach has been applied should be described as necessary to provide an adequate description of the LMP-Based Affirmative Safety Case. It is not necessary to repeat aspects of the methodology already covered in NEI 18-04, but rather to point out the specifics of how the methodology was applied within the range of options specified in NEI 18-04. Details of the analyses should be present in the design records.

The safety classification approach in NEI 18-04 is based on the PRA Safety Functions that are identified in the definition and selection of the AOOs, DBEs, and BDBEs in Chapter 3. Tables in the following subsections list the SR SSCs and NSRST SSCs and the specific prevention and mitigation functions reflected in the LBEs and responsible for the safety classification.

5.2 Required Safety Functions This section should present the RSFs, which are the product of applying Step 5a in Figure 3-2 of NEI 18-04. The RSFs are the PRA Safety Functions that are responsible for successfully mitigating the consequences of all the DBEs inside the F-C Target and for successfully preventing any high consequence BDBEs (i.e., those with doses exceeding 25 rem) from increasing in frequency beyond the F-C Target. A summary-level justification of how the reactor-specific RSFs adequately support the FSFs should be included. Examples of RSFs from the modular high temperature gas-cooled reactor (MHTGR) and Power Reactor Inherently Safe 49

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Module (PRISM) reactors are found in the LMP LBE report,11 and other examples are found in LMP tabletop reports for the Xe-100,12 Fluoride-Cooled High Temperature,13 eVinci',14 Molten Salt Reactor Experiment,15 and PRISM16 reactors.

Commented [A67]: Identification of RFDCs as the sole PDCs does not meet the Part 50 Appendix A requirements.

5.3 Required Functional Design Criteria and Principal Design Criteria While the GDC in 10 CFR 50 Appendix A are not requirements for non-LWRs, they are considered to be Regulations (e.g., 10 CFR 50.34 and 10 CFR 52.79) require the identification of PDC. Plants generally applicable to non-LWRs and are intended to that use the NEI 18-04 methodology may adopt the Required Functional Design Criteria provide guidance in establishing the principal design criteria. When developing the PDCs, the underlying safety (RFDC), which are derived from the RSFs, as the PDC. The identification of RFDC is described objectives of the GDC still apply. For example, RFDCs do in Task 7 under Figure 4-1 in NEI 18-04. Each RFDC constitutes a PDC. address PDCs for normal operation while GDCs do address requirements for normal operation.

The systematic approach detailed in this guidance for identifying PDC using the RFDC produces Commented [A68]: (a) As stated in Southern Co. July a set of risk-informed PDC that establish the functional requirements of a plant that are required 2021 white paper on PDCs, the design-specific RFDCs comprise the design-specific PDCs for a design. However, to meet the performance objectives of the FSFs. This is an alternative to the traditional the RFDCs only address the How aspect of the design, deterministic approach to identifying PDC used for light water reactors. Notably, the whereas PDCs represent the How and How well aspects information contained within the set of design-specific PDC using the approach identified in this of the design. Therefore, RFDCs can not be considered equivalent to the PDCs required by 10 CFR 50.34(a) and guidance may not be identical to the information contained within the General Design Criteria described in 10 CFR 50, Appendix A. PDCs address the How in 10 CFR Part 50, Appendix A or the Advanced Reactor Design Criteria in Regulatory Guide well aspects of a design by specifying the single failure 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors.17 criterion, redundancy, diversity, etc. as well as performance criteria. The LMP process does not specify these aspects According to Regulatory Guide 1.232, neither the General Design Criteria nor the Advanced (redundancy, etc.) of design but rather specifies reliability in Reactor Design Criteria are regulatory requirements for non-LWR applicants. their place. Therefore, to be a substitute for PDCs, the RFDCs should specify a reliability target along with the design feature. This would also be a good way to document This section should present the PDC in terms of the RFDC for each of the RSFs as described in reliability targets in the SAR.

Task 7 of Figure 4-1 in NEI 18-04. These RFDC may be regarded as a decomposition of the RSFs into sub-functions that are necessary and sufficient to support the RSFs. The key elements (b) Applicants are required to propose PDCs for their designs. It sounds like TICAP is saying the RFDCs are good of the RFDC that should be identified include: enough, but they do not meet the intent of the PDCs as defined in 10 CFR 50, Appendix A.

  • The design criteria that must be satisfied to meet each of the design-specific RSFs
  • A breakdown of each RSF into reactor design-specific sub-functions that are necessary Commented [A69]: There is no specific mention of PDCs and sufficient to ensure successful completion of the RSF for all the DBAs (The RFDC or RFDCs associated with normal operations or ensuring safe stable end states or any discussion of exemption are qualitatively described in a manner that translates the definition of each RSF into requests related to scope of PDCs/RFDCs. This topic may be functional design criteria; they form a bridge between the RSFs and the Safety-Related an area where the NRC staff should consider whether to Design Criteria (SRDC), which are assigned to specific SSCs in performing the RSFs.) include clarification and/or exceptions based on the feedback provided by OGC and ultimate alignment on an NRC position.

Commented [A70]: NEI 18-04 states that NRC 11 Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Selection and Evaluation of Licensing Basis Regulatory Guide 1.232, Developing Principal Design Events, Idaho National Laboratory, Rev. 1, March 1, 2020. Criteria for Non-Light Water Reactors, should be used as 12 High Temperature Gas-Cooled Pebble Bed Reactor Licensing Modernization Project Demonstration, Southern Company, August 2018.

one input by designers to initially establish principal design (ML18228A779) criteria for a facility based on the specifics of its unique 13 Fluoride-Cooled High Temperature Reactor Licensing Modernization Project Demonstration, Southern Company, December 2019.

design. It states further that RFDCs are defined to capture (ML19247C198) 14 Westinghouse eVinciTM Micro-Reactor Licensing Modernization Project Demonstration, Southern Company, August 2019. (ML19227A322) design-specific criteria that may be used to supplement or 15 Molten Salt Reactor Experiment (MSRE) Case Study Using RiskInformed, PerformanceBased Technical Guidance to Inform Future Licensing modify the applicable General Design Criteria or Advanced for Advanced NonLight Water Reactors, Southern Company, September 2019. (ML19249B632) Reactor Design Criteria in the formulation of Principal 16 PRISM Sodium Fast Reactor Licensing Modernization Project Demonstration, Southern Company, December 2018. (ML19036A584) Design Criteria.

17 Regulatory Guide 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, U.S. Nuclear Regulatory Commission, Rev 0, April 2012. Why are these statements from NEI 18-04 not used in this guidance?

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Content for a Licensing Modernization Project-Based Affirmative Safety Case

  • An identification of the design-specific inherent or intrinsic reactor characteristics that must be preserved to support the LMP-based safety case and are credited in the selection of the SR SSCs (Examples of such characteristics include but are not limited to fuel and reactor material properties, geometry, power level, and power density when they enable the satisfaction of the RSFs via passive means.)

It is also important to note that RFDC include intrinsic features of the reactor and plant. Table 5-1 provides selected examples of RFDC that were developed for the MHTGR for two RSFs (see Appendix A, Table A-3 of the LMP SSC report18 for the entire list covering additional RSFs).

This table is not intended for direct inclusion in the SAR but is provided for informational purposes. Commented [A71]: This text and table are all italics. The guidance (non-italic text) should specify that the application Table 5-1. MHTGR Required Safety Functions and Associated Required Functional Design Criteria contain similar content even if the specific table format below is not used.

Required Safety Functions - Required Functional Design Criteria Subfunctions Retain I: The reactor fuel shall be designed, fabricated, and operated in such a manner that minor Radionuclides in radionuclide releases from the fuel to the primary coolant will not exceed acceptable values. Commented [A72]: Clarity: Any, vs minor, radiological Fuel Particles releases?

III: The reactor shall be designed, fabricated, and operated in such a manner that the inherent nuclear feedback characteristics will ensure that the reactor thermal power will not exceed Control Heat acceptable values. Additionally, the reactivity control system(s) shall be designed, fabricated, Generation and operated in such a manner that during insertion of reactivity, the reactor thermal power will not exceed acceptable values.

IIIa: Two independent and diverse sets of movable poison equipment shall be provided in the Control with design. Either set shall be capable of limiting the heat generation of the reactor to acceptable Movable Poisons levels during off-normal conditions.

IIIb: The equipment needed to sense, command, and execute a trip of the control rods, along Shutdown with any necessary electrical power, shall be designed, fabricated, and operated in such a Reactor manner that reactor core shutdown is assured during off-normal conditions.

IIIc: The equipment needed to sense, command, and execute a trip of the reserve shutdown Shutdown control equipment, along with any necessary electrical power, shall be designed, fabricated, Reactor Diversely operated, and maintained in such a manner that the shutdown of the reactor core is assured during off-normal conditions.

IIId: The design, fabrication, operation, and maintenance of the control rod guide tubes, the graphite core and reflectors, the core support structure, the core lateral restraint assemblies, the reactor vessel, and reactor vessel support shall be conducted in such a manner that their integrity Maintain is maintained during off-normal conditions as well as provide the appropriate geometry that Geometry for permits the insertion of the control rods into the outer reflector to effect reactor shutdown.

Insertion of IIIe: The design, fabrication, and operation of the reserve shutdown control equipment guide Movable Poisons tubes, graphite core and reflectors, core support structure, core lateral restraint assemblies, reactor vessel, and reactor vessel support shall be conducted in such a manner that their integrity is maintained during off-normal conditions, as well as provide the appropriate geometry that permits the insertion of reserve shutdown control material to effect reactor shutdown.

Note: The examples above are only a subset of the complete list of MHTGR RFDC.

18 Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Safety Classification and Performance Criteria for Structures, Systems, and Components, Rev 0, Idaho National Laboratory, March 2020.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 5.4 Safety-Related SSCs Table 5-2 is an example means of displaying the combinations of SSCs that are provided in the design to fulfill each RSF and identifying whether each set of SSCs is available or not on each of the DBEs. There is one table per RSF. The provisions in the design for alternative ways to perform each RSF constitute one element of Plant Capability DID. The tables identify which combination of SSCs is selected as SR for each RSF. For this RSF example, the selected SSC combination is shown in red font. Table 5-2 illustrates an important intermediate step in the LMP methodology, but the SSC combinations that are not selected are not design basis Commented [A73]: As stated in this paragraph, the information. The applicant is not required to provide Table 5-2 information in the SAR, but the alternate ways to perform each RSF constitute one element of plant capability DID. If so, why are they not considered information should be available in the design records. design basis information? In effect, this says not all of the DID features are in the design basis (i.e., not in the SAR).

Table 5-2. Example Table of Evaluation of SSCs for Core Heat Removal RSF How can this be justified? Would the combinations not selected be described elsewhere as part of defense in depth SSC Combinations Capable of Providing Available for Available for Available for Selected or NSRST discussions.

Core Heat Removal* DBE-1? DBE-2? DBE-N? as SR?

Commented [A74]: This text and table are all italics. The Reactor guidance (non-italic text) should specify that the application Heat Transport System Yes No No No contain similar content even if the specific table format Energy Conversion Area (ECA) below is not used.

Reactor Shutdown Cooling System No Yes No No Shutdown Cooling Water System (SCWS)

Reactor Reactor Vessel (RV) Yes Yes Yes Yes Reactor Cavity Cooling System (RCCS)

Reactor Reactor Vessel Yes Yes Yes No Commented [A75]: Clarity: Not being selected as SR here Reactor Building (RB) passive heat sinks is not clear. The following from Appendix A of the LMP SSC report is very useful: The option that relied on the passive

  • The entries in this column and the example selection as SR are examples from the MHTGR are found in heat sinks in the reactor building as the ultimate heat sink Appendix A of the LMP SSC report.1819 was rejected, as that approach involved the need to address uncertainties regarding concrete degradation, which are The entries in Table 5-2 are an example developed for the MHTGR for a core heat removal RSF. removed with a robust and reliable RCCS. This is an example of how deterministic defense-in-depth considerations had a Note that the selection of SR SSCs in this example includes SSCs needed to preserve the intrinsic tangible impact on the selection of safety-related SSCs and characteristics of the reactor, such as power level, power density, and shape and selection of selection of DBAs for the MHTGR.

materials that enable the RSF to be fulfilled with the other identified SSCs.

A summary, as shown in Table 5-3, should be presented that lists all the SR SSCs, the AOOs, DBEs, and BDBEs, and the PRA Safety Functions responsible for preventing or mitigating each of these LBEs. Given there are multiple RSFs and that each RSF may require the use of multiple SSCs, there will, in general, be multiple SR SSCs. Operator actions that may be necessary to perform any of these functions should be identified through instrumentation and equipment needed to implement those operator actions.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Table 5-3. Example Table of Evaluation of SSCs for PRA Safety Functions LBE Type PRA Safety SR SSC LBE (AOO, DBE, or BDBE) Function LBE-11  ? PSF-11 LBE-12  ? PSF-12 SR SSC-1 LBE-1n  ? PSF-1n Additional SR SSCs This table is organized by SR SSC in order to identify the capabilities of each SR SSC in preventing or mitigating each applicable LBE. These capabilities are used in Chapter 6 to support selection of special treatments. While SSCs are identified as SR for their role in mitigating DBEs and high consequence BDBEs, those SSCs may be used in the mitigation of other LBEs as well, and the table captures that information. The reliability and capability targets for SSCs address all LBEs, not just DBEs and high consequence BDBEs.

The LBE index numbers in the second column should be keyed to LBE indexes identified in Chapter 3 or alternatively spelled out. For each PRA Safety Function identified in the last column, the spelled-out function should be listed.

5.5 Non-Safety-Related with Special Treatments SSCs This section presents the NSRST SSCs, the technical basis for the selection of NSRST SSCs, and identifies the PRA Safety Functions for the NSRST SSCs reflected in the LBEs in Chapter 3.

Non-safety-related SSCs that are classified as NSRST because they perform risk-significant safety functions are identified in Section 5.5.1. Non-safety-related SSCs that are classified as NSRST because they perform safety functions deemed necessary for adequate DID are identified in Section 5.5.2.

5.5.1 Non-Safety-Related SSCs Performing Risk-Significant Functions This section identifies the non-safety-related SSCs that perform risk-significant functions and meet the risk significance criteria for classification as NSRST. The risk significance classification is based on applying Steps 4B and 5B in Figure 4-1 in NEI 18-04 and the SSC risk significance criteria noted in Section 4.2.2 of NEI 18-04. Supporting documentation for details, calculations, etc., that were used to establish risk-significant SSC functions should be part of the design records. Commented [A76]: The guidance should specify that information similar to the content of Tables 5-1 and 5-2 There are two types of risk significance criteria that come into play in NEI 18-04 for non-safety- should be included in this section.

related SSCs. The first criterion is based on identifying non-safety-related SSCs whose prevention or mitigation function is necessary to prevent one or more LBEs from exceeding the F-C Target. Any SSC functions that are risk risk-significant based on this criterion should be identified in a table such as the example in Table 5-4. The purpose of the table is to identify the risk-significant SSCs, the PRA Safety Functions that are responsible for the classification, and 53

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Content for a Licensing Modernization Project-Based Affirmative Safety Case the LBEs that would exceed the F-C Target if the PRA Safety Functions were not available.

Operator actions that may be necessary to perform any of these functions should be identified in the description of the PRA Safety Functions, as well as the instrumentation and equipment needed to implement those operator actions.

Table 5-4. Example Table of SSCs Risk-Significant Due to F-C Curve LBE Type (AOO, PRA Safety NSRST SSC LBE DBE, or BDBE) Function RS-NSRST SSC-1 LBE-RS-1  ? PSF-RS-1 RS-NSRST SSC-2 LBE-RS-2  ? PSF-RS-2 RS-NSRST SSC-N LBE-RS-N  ? PSF-RS-N The second risk significance criterion is based on whether the cumulative contribution of the LBEs in which an SSC safety function is failed exceeds 1% of the cumulative risk metrics used for evaluating the risk significance of LBEs. In this case, each risk-significant SSC is classified this way based on an accumulation of risk from multiple LBEs. These risk-significant SSCs should be identified in a table such as the following example (Table 5-5). The purpose is to identify the SSC classified as risk significant, the LBEs in which the SSC is failed, and the PRA Safety Function associated with that LBE. Operator actions that may be necessary to perform any of these functions should be identified in the description of the PRA Safety Functions, as well as the instrumentation and equipment needed to implement those operator actions.

Table 5-5. Example Table of SSCs Risk-Significant Due to Cumulative Risk LBE Type PRA Safety NSRST SSC LBE (AOO, DBE, or BDBE) Function LBE-RS-11  ? PSF-RS-11 LBE-RS-12  ? PSF-RS-12 RS-NSRST-SSC-1 LBE-RS-1n  ? PSF-RS-1n Additional RS-NSRST SSCs 5.5.2 Non-Safety-Related SSCs Performing Safety Functions Necessary for Adequate DID This section identifies the non-safety-related SSCs that are classified as NSRST because they perform safety functions deemed necessary for adequate DID. It should be noted that the SR SSCs identified in Section 5.4 are also key elements of the plant capability DID.

As with the risk-significant SSCs, the SSC classification for DID adequacy is normally tied to specific LBEs and should be summarized in a table such as Table 5-6. There may be some NSRST SSCs that were identified via the IDP that were not modeled in the PRA and not reflected explicitly in the LBEs due to limitations in the PRA or items screened out of the PRA.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Table 5-6. Example Table of SSCs Risk-Significant Due to DID LBE Type PRA Safety NSRST SSC LBE (AOO, DBE, or BDBE) Function LBE-DID-11  ? PSF-DID-11 LBE-DID-12  ? PSF-DID-12 DID-NSRST-SSC-1 LBE-DID-1n  ? PSF-DID-1n Additional DID-NSRST SSCs Note: If an SSC is classified as NSRST but is not associated with a specific LBE, specify N/A under LBE and LBE type and identify the SSC function responsible for the NSRST classification under PRA Safety Function.

5.6 Complementary Design Criteria Commented [A77]: It would be useful to give examples of CDCs similar to what is done for RFDCs in Table 5-1.

It is important to understand that Complementary Design Criteria (CDC), as they apply to Since CDCs are not discussed in regulations or NEI 18-04, it NSRST SSCs, are somewhat, but not totally, analogous to PDC and how PDC apply to SR SSCs. is unclear why this section is included in this document.

PDC identified through TICAP are defined at the functional level and are based on the RFDC, as described in Section 5.3 of this guidance. PDC are top-down in that they correspond to RFDC and are independent of the actual SR SSCs that satisfy the RFDC. NEI 18-04 defines no direct counterpart for RFDC that flow down to NSRST SSCs. Section 5.5 describes how SSCs are identified as NSRST because they perform risk-significant functions or are identified as necessary for DID. CDC are not defined in NEI 18-04 but are useful in the description of the Commented [A78]: Are CDCs part of the safety case?

LMP-based safety case. Unlike PDC, CDC are identified as bottom-up in that they relate to What does useful mean?

specific NSRST SSCs, not higher-level functions. Accordingly, while PDC are defined at the Commented [A79]: This sounds like a deviation from the functional level, CDC may also be defined at the functional level (related to the PRA Safety endorsed NEI 18-04 methodology without specifically calling it out. Perhaps the NRC staff should include a clarification in Functions that are satisfied by the NSRST SSCs), or CDC may be expressed more at an SSC its RG on this topic.

level, directly linked to the NSRST SSCs themselves. The designer has latitude in choosing the approach, as long as the success criteria (discussed below) are clearly conveyed. Commented [A80]: A sentence might be added here to convey that an applicant may also choose to use the CDC to The CDC for NSRST SSCs are defined in terms of the success criteria for the PRA Safety address design functions that go beyond achieving the analytical end states associated with LBEs. This would make Functions that are represented in the PRA model to prevent and mitigate the LBEs responsible the CDCs available to address some of the questions related for the safety classification. For example, a PRA Safety Function might be Provide adequate to normal operations and possible longer term functions such as bring a reactor to a subcritical condition.

heat removal from the reactor following initiating event X, and the success criterion might be Provide a coolant flow rate of Y kg/sec within Z minutes and maintain maximum fuel temperature less than ZZ for a specified set of LBEs. SSCs are classified as NSRST either because the LMP risk significance criteria are met as identified in Section 5.5.1, or the criteria for adequate DID established by the IDP are met as identified in Section 5.5.2. The reliabilities and capabilities that are modeled in the PRA for the PRA Safety Functions associated with the SSC trigger the meeting of the risk significance or DID adequacy criteria for NSRST classification. These, in turn, serve to prevent and/or mitigate a specific set of LBEs. Hence the CDC for the NSRST SSCs are directly tied to the success criteria established in the PRA for the Commented [A81]: Why arent the CDCs that are PRA Safety Functions responsible for the SSC classification as NSRST. necessary to achieve these success criteria considered to be a part of the PDCs, and included in that category for NRC staff review as a part of the safety assessment process?

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Content for a Licensing Modernization Project-Based Affirmative Safety Case These should be presented in tabular form by listing the SSC, the PRA Safety Function(s) responsible for its safety classification as NSRST, and the design criteria that are necessary and sufficient to meet the PRA Safety Function (see example in Table 5-7). There may be more than one PRA Safety Function that is associated with the NSRST classification and more than one design criterion for each PRA Safety Function because the SSC may be represented in multiple LBEs.

Table 5-7. Example Table of NSRST SSCs with Corresponding CDC PRA Safety NSRST SSC Complementary Design Criteria Function PSF-11 Design criterion for PSF-11 PSF-12 Design criterion for PSF-12 NSRST SSC-1 PSF-1n Design criterion for PSF-1n Additional NSRST SSCs Two-Step Licensing (CP Content)

For a CP application, Chapter 5 includes preliminary determination of the RSFs and RFDC (which are the PDC when using the LMP methodology), safety classification of SR and NSRST SSCs, and the Complementary Design Criteria leading to specific NSRST SSC design requirements. The LMP methodology for assessing safety functions, design criteria, and SSC safety classification, as described in Chapter 5 of the COL guidance, draws on results from the initial PRA. The PRA methodology described in Chapter 2 should be used in the preliminary determination of RSFs, determination of RFDC (PDC), CDC, and SSC safety classification. The structure and preliminary content of the Chapter 5 sections should follow the structure and content of the COL guidance Chapter 5 content. It should be noted that as the design evolves and construction continues, changes to the PRA modeling and results can be expected. Thus, the results in Chapter 5 are preliminary and can be expected to change as the design is finalized and the FSAR is developed for the OL application. At the CP stage, the use of performance-based bounding conditions for RFDC may be possible depending on the simplicity of the design and corresponding confidence in the PRA outputs.

Thus, the PSAR content for Chapter 5 should include functional decomposition of FSFs to RSFs, a preliminary set of RFDC (PDC) with performance-based criteria, and preliminary SSC classifications based on preliminary PRA results.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 6 SAFETY-RELATED SSC CRITERIA AND CAPABILITIES In Section 5.4, the SR SSCs were identified, and the bases for their classification were provided.

Chapter 6 provides further detail on the criteria and capabilities of all SR SSCs in the LMP-based affirmative safety case, consistent with the NEI 18-04 methodology. This further detail includes SRDC, reliability and capability performance-based targets, and special treatment requirements to provide sufficient confidence that the performance-based targets intended in the design will be achieved in the construction of the plant and maintained throughout the licensed plant life. Chapter 6 also summarizes design requirements for non-safety-related SSCs that provide confidence that the non-safety-related SSCs will not adversely impact the ability of SR SSCs to support RSFs in the event that a hazard occurs at the DBEHL. Note that these non-safety-related SSCs may be classified as NSRST or NST in the SSC safety classification process of LMP, but that classification process is based on their normal PRA Safety Functions, not on the passive function of protecting (or at least not impairing) SR SSCs as discussed in Section 6.1.3.

As discussed in NEI 18-04, an LMP-based safety case is both risk-informed and performance-based. The formulation of the SRDC should be quantitatively framed so that the successful performance of the SSC may be confirmed via calculation or monitoring to facilitate the performance-based attribute of the safety case, as discussed in Section 6.1. Additional performance-based elements are incorporated in the selection of SSC reliability targets (Section 6.2) that are used to establish special treatment requirements (Section 6.3).

This guidance assumes that the applicant followed the NEI 18-04 methodology for the establishment of SR SSC criteria and capabilities. If the applicant did not use parts of the methodology or used an alternative method, the following statement should be added:

The following deviations from the approved methodology were taken.

The applicant should go on to list each deviation, describe any alternative method, and provide justification for the approach employed.

6.1 Design Requirements for Safety-Related SSCs This section describes the outputs of NEI 18-04 Section 4.1, Task 7. Details of the analyses and justifications for the development of SRDC should be in the design records.

6.1.1 Design Basis External Hazard Levels One general category of design requirements flows from the need to protect the SR SSCs in the performance of their RSF from design basis external hazards. Each external hazard is characterized by a DBEHL (e.g., wind speed). This is discussed in NEI 18-04 Section 3.2.2, Task 6, and the following text from the first page of Section 4 in NEI-NEI 18-04.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case It is noted that there will be design requirements to protect all SR SSCs from any adverse impacts of any DBEHLs. This may lead to design requirements to prevent any adverse impacts from the failure of an SSC classified as NST or NSRST that could otherwise prevent an SR SSC from performing its RSFs.

It is important to note that the DBEHLs go beyond environmental hazards originating external to the plant. As defined in NEI 18-04, the scope of the DBEHLs includes external hazards such as seismic events, wind including tornados and wind-generated missiles, external flooding, hazards from external facilities, and internal plant hazards such as internal fires, internal floods, high energy line breaks, and internally generated missiles. These internal plant hazards are frequently described as area events. Guidance on the scope of hazards may be found in Commented [A82]: See below my comments on DBEHLs Chapter 3 of the Standard Review Plan (NUREG-0800). The concept is to ensure that hazards related to internal plant hazards (e.g., internal fire, HELB, etc) with a frequency down to 10-4/plant-year are identified so that design requirements identified in Section 5.4 for the SR SSCs to protect them against any DBEHL can be specified. Each DBEHL It appears that NEI 18-04s discussions and definition of may impact one or more reactors as well as non-reactor radioactive sources. Note that the DBEHLs are limited to external hazards (not internal plant hazards) as written. If this is true, it is a deviation from NEI DBEHLs are one of the inputs to the analysis of hazards in the PRA. 18-04.

Commented [A83]: How are external event frequencies NEI 18-04 identifies two alternative approaches for the selection of DBEHLs. One is to use the lower than 10E-4/yr addressed?

existing criteria in Chapter 3 of the Standard Review Plan, in which case the selection is made deterministically. The second alternative is to derive the DBEHLs based on a probabilistic hazard analysis subject to review by the NRC staff. For each hazard, the method used should be clearly identified in the SAR. For hazard levels derived probabilistically, the technical requirements for probabilistic hazard analysis in the non-LWR PRA standard should be used and referenced. The PRA documentation would then provide the technical basis for the Commented [A84]: where? ARCAP Chapter 2?

derivation of those DBEHLs. Commented [A85]: Beyond listing a hazard level value in a table, the guidance should specify that the applicant The DBEHLs should be summarized in this section. A tabular form such as Table 6-1 is describe how each DBEHL is used as an input parameter to the design analysis of safety-related SSCs. For example:

recommended. The determination of the DBEHLs is documented elsewhere in the SAR.

  • Regarding the seismic design input, where would the application describe the design ground motion response Table 6-1. Example Table of Design Basis External Hazard Levels spectra, floor response spectra, damping values, and time histories? Refer to SRP Section 3.7 for relevant topics to be Hazard Design External Hazard Level addressed.

Specify design basis earthquake parameters such as safe shutdown earthquake peak

  • For design basis wind hazards, where would an ground acceleration, design response spectra, damping values, time histories, etc. application describe the design-basis wind loadings Seismic Events Note that there are no requirements for specifying an operational operating basis including design wind velocity and its recurrence interval earthquake as part of an LMP-based safety case. and the methods used to transform the wind velocity into an effective pressure applied to surfaces of structures?

Tornadoes Wind Speed Specify design basis wind loadings and design parameters Refer to SRP Section 3.3 for relevant topics to be addressed.

External Floods Specify design basis flood levels and design parameters Internal Fires Identify fire areas where SR SSCs are located and fires may occur Commented [A86]:

Clarity: Is identifying areas sufficient? DBEHL is defined as Internal Floods Identify flood areas where SR SSCs are located and flood may occur A design specification of the level of severity or intensity of an external hazard for which the Safety-Related SSCs are High energy line breaks Identify areas where SR SSCs are located and a high energy line break may occur designed to withstand with no adverse impact on their capability to perform their RSFs. In addition to the areas, Other internal and external appropriate hazard severity and design parameters should Identify hazards and affected areas and specify appropriate hazard severity and hazards that SR SSCs are be added. The same comments apply to the next two design parameters.

protected against hazards - internal flood and HELBs.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Note: The development of the DBEHLs is addressed by ARCAP and summarized in SAR Chapter 2. Commented [A87]: Question: In this document, DBEHLs include internal plant hazards (e.g., internal flood). ARCAP Chapter 2 ISG (Site Info) appears address only hazards Two-Step Licensing (CP Content) external to plants. Is there any gap in ARCAP guidance?

The CP application content for Section 6.1.1 addressing DBEHLs should be as complete as possible based on the site characterization required in 10 CFR 50.34(a). The external and internal hazard levels are inputs to the PRA and to SSC design. Thus, including details on this information in the PSAR will support demonstrating the viability of the design and developing the FSAR consistent with Chapter 6 of the COL guidance.

Design Certification Section 6.1.1 should describe the bounding site characterizations considered in the definition of the DBEHLs. To the extent that the hazard characterizations impact internal hazard levels, those impacts should be identified and included in the PRA.

PRA requirements for the selection and documentation of the bounding site characteristics (i.e.,

the site parameter envelope) used in the hazard PRAs are found in the Advanced non-LWR PRA Standard, ASME/ANS RA-S-1.4-2021.

6.1.2 Summary of SRDC In the text for Task 7 of Figure 4-1 in NEI-18-04, it is stated:

The RFDC, SRDC, the reliability and capability targets for SR and NSRST SSCs, and special treatment requirements for SR and NSRST SSCs define safety-significant aspects of the descriptions of SSCs that should be included in safety analysis Commented [A88]: Does this mean that TICAP now reports. agrees to put reliability and capability targets in the SAR?

The RFDC are identified in Section 5.3, and the RSFs that they support are identified in Section 5.2. For each of the RFDC, this section should identify a set of SRDC appropriate to the SR SSCs selected to perform the RSFs. These SRDC exclude Special Treatment Requirements, which are separately covered in Section 6.2. The RFDC, which are expressed in the form of functions and involve collections of SSCs and intrinsic capabilities of the plant, may be viewed as a bridge between the RSFs and the SRDC. The SRDC is more detailed requirements for specific SR SSCs in the performance of the RSF functions in specific DBAs. Examples of SRDC that were developed for the MHTGR are found in Appendix A of the LMP SSC report.1819 For the SRDC, the following information is presented in tabular form, as shown in Table 6-2:

  • The first column contains the SSC names.
  • The second column provides brief SSC functional descriptions.

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  • The third column lists the RFDC that the SR SSCs support. Most likely, there is only one RFDC associated with each SR SSC, but if there is more than one, all should be listed.

Note that the links from the SR SSCs back to the LBEs that define the RSFs are provided in Chapter 5.

  • The fourth column lists the SRDC. There may be more than one SRDC for each SR SSC.

Table 6-2. Example Table of Safety-Related Design Criteria for SSCs SR SSC Functional Description RFDC SRDC SRDC11 SRDC12 SR SSC1 Functional Description of SR SSC1 RFDCx SRDC1n Additional SR SSCs 6.1.3 Summary of DBEHL-Related Requirements for Non-Safety-Related SSCs Chapter 6 also identifies DBEHL-related design requirements for non-safety-related SSCs.

These design requirements are to support the special safety functions that are applied to the non-safety-related SSCs to prevent adverse impacts on the ability of the SR SSCs to perform the RSFs. An example is the requirement for anchorage to prevent a non-safety-related SSC from failing in such a manner that it would impact an SR SSC and cause it to fail to perform its RSF.

It is important to note that the non-safety-related SSCs covered in these requirements are not for the SSC functions that they normally perform but for the special function of preventing any adverse impact on the capability of any SR SSC in the performance of the RSF. The DBEHL includes external hazards such as seismic events as well as internal plant hazards such as internal fires and floods, turbine missiles, and high energy line breaks. When a non-safety-related SSC is required to protect the SR SSCs in their ability to perform their RSFs, such non-safety-related SSCs are not necessarily NSRST. The NSRST classifications are based on the PRA Safety Functions these SSCs perform to prevent or mitigate event sequences and not functions that are focused on protecting the SR SSCs. Commented [A89]: Where are these covered?

For the non-safety-related SSCs that have design requirements to protect the SR SSCs in the performance of the RSFs in response to a DBEHL, the following information in tabular form should be provided, as illustrated in the example below (Table 6-3).

  • The first column identifies the non-safety-related SSCs.
  • The second column lists the RSFs and SR SSCs that are protected.
  • The third column identifies the DBEHLs that are associated with these requirements.
  • The fourth column identifies the specific design requirement for the function to protect the SR SSCs for each of the DBEHLs. Note that this function is different from the PRA Safety Functions for the same non-safety-related SSCs.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Table 6-3. Example Table of Non-Safety-Related SSCs Protecting SR SSCs from DBEHLs Non-Safety-Related Non-Safety-Related SSC Protected RSF and SR SSC DBEHL SSC Design Requirement Non-safety-related DBEHL-1 DC-11 Non-safety-related Non-safety-related DBEHL-2 RSF/SR SSCx DC-12 SSC-1 Non-safety-related DBEHL-n DC-n Additional non-safety-related SSCs 6.2 Reliability and Capability Targets for SR SSCs In the NEI 18-04 methodology, the main purpose of establishing reliability and capability targets is to identify special treatment requirements. In accordance with NEI 18-04 Section 4.1, the targets are formally documented at the SSC level. Whether it is done at the system level or for components of a system is a detail that is left to the applicant.

The term reliability as used informally in NEI 18-04 refers to the reliability performance metrics involved in the estimation of event sequence frequencies. Reliability metrics include unavailability, unreliability, event occurrences, time out-of-service, fraction of time in an operating state, etc. as needed to evaluate safety function failure probabilities in the PRA.

Reliability is not observable but rather is calculated based on observed performance measures and available generic evidence.

The term capability is a performance measure used to establish the successful prevention or mitigation of LBEs. Capability is linked to the success criterion used to quantify the failure probability in the PRA. An SSC involved in mitigating two separate LBEs could have different capability targets for each LBE.

Reliability and capability targets can be established at different levels.

  • Plant level: controlling the frequencies, consequences, and risk significance of the LBEs
  • Functional level: controlling the reliabilities and capabilities of safety functions across multiple LBEs
  • SSC level: controlling the reliabilities and capabilities of individual SSCs (or humans) in the performance of safety function Plant level targets are addressed as part of Chapter 3 and require no additional documentation.

Applicants may choose to document functional level targets in Chapter 5, but that is not a requirement. Commented [A90]: The expectation regarding documentation of SSC-level targets is not clear. The first sentence indicates theyll be included, while the second SSC level targets should be documented in Section 6.2. Except for the simplest of designs, sentence indicates that it would be a challenge to include reliability and capability targets for every component could at the SSC level will be quite them in the SAR because of their high volume.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case voluminous and challenging for incorporation directly in the SAR. Therefore, it is expected the SSC-level reliability and capability targets will instead be maintained in plant records as IBR material. In those records, each SSC should specify the reliability and capability targets as well Commented [A91]: Need to be clear what is the IBR as the LBE(s) that the SSC prevents or mitigates. Describe SR SSC reliability targets and material. If the associated programs are IBR as indicated in the following paragraph, delete as IBR material or replace performance requirements used as input to the PRA for SSCs that were used to develop the with referenced vs IBR. Reference to the actual SSC selection of special treatment requirements (i.e., programmatic actions used to maintain information being IBR would require that it be submitted to performance within the design reliability targets). This description should include: the NRC as a separate document, not just available for audit.

  • numerical targets for SSC reliability and availability, Strictly speaking it appears to be a deviation from NEI 18-04, which states: The RFDC, SRDC, the reliability and capability
  • design margins for performance of the RSFs, and targets for SR and NSRST SSCs, and special treatment requirements for SR and NSRST SSCs define safety-significant aspects of the descriptions of SSCs that should be
  • requirements for monitoring of performance against these targets with appropriate included in safety analysis reports. It is also stated under corrective actions when targets are not fully realized. Section 6.1.2 earlier (Pg 57).

Commented [A92]: Plant records are not IBR material.

In addition to the target information or associated reference, Section 6.2 should specify the plant IBR material must be put into the docket.

program(s) used to maintain the reliability and capability targets. Information on the program(s) should be provided in Chapter 8.

6.3 Special Treatment Requirements for SR SSCs NEI 18-04 adopted the definition of special treatment that is provided in RG 1.201, which was developed for implementing 10 CFR 50.69.

special treatment refers to those requirements that provide increased assurance beyond normal industrial practices that structures, systems, and components (SSCs) perform their design-basis functions.

Anything that is done beyond procuring commercial-grade equipment to provide increased assurance in the capability and reliability of the SSC falls into the category of special treatment.

Hence, all the design requirements provided in Section 6.1 are part of the special treatment.

This section identifies the additional special treatments that are applied to SR SSCs. Candidate special treatments for consideration are identified in Table 4-1 of NEI 18-04.

As noted in Section 4.4.5 of NEI 18-04, the selection of special treatments for all safety-significant SSCs (SR and NSRST) is informed by a set of targets for the reliability and availability of the SSCs in their prevention functions as well as targets for the capability of the SSCs in the performance of their mitigation functions. These specific targets are addressed in Section 6.2. The focus of this section in the application is to produce the resulting special treatment requirements.

For the selected special treatments, the license application should identify the treatments in the license application with details available for NRC audits in the license application. This section should describe, as applicable, the special treatment requirements from NEI 18-04, Table 4-1, on 62

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Content for a Licensing Modernization Project-Based Affirmative Safety Case a case-by-case basis and in the context of the SSC functions in the prevention and mitigation of applicable LBEs. Describe special treatments including the following, as applicable:

  • Equipment qualification
  • Seismic qualification
  • Materials qualification
  • Pre-service and risk-informed in-service inspections
  • Pre-op and startup testing requirements Surveillance testing requirements including test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met (i.e., demonstrate the ability to perform the safety function).

The special treatments should be summarized in tabular form (see Table 6-4) by listing each SR SSC, providing a brief performance-based functional description, and identifying which special treatment has been selected for each SR SSC.

Some plant designs may involve first-of-a-kind testing for SR SSCs. To the extent those first-of-a-kind tests have not been completed by the time of the application, they may be included as Commented [A93]: This is somewhat unclear - it seems special treatments. to refer to FOAK plant testing, which is confirmatory in nature and supported by the statement. However, it could be read that it refers to FOAK functional demonstration Table 6-4. Example Table of SR SSC Special Treatments testing, which is (generally) needed to make the safety SR SSC Functional Description SR SSC Special Treatments finding and therefore would not support the timeline SR SSC1 Special Treatment No. 1 described. Additional clarity on the nature of the testing described here would be useful.

Short SSC functional SR SSC1 Special Treatment No. 2 SR SSC1 description for SR SSC1 Commented [A94]: Within the context of the baseline for SR SSC1 Special Treatment No. n this guidance being a COL application, the NRC staff should Additional SR SSCs consider whether or not any of these FOAK tests need to be included in ITAAC and addressed in the ARCAP guidance.

Commented [A95]: Testing and verification of advanced 6.4 Descriptions for SR SSCs reactor design features need to comply with 10 CFR 50.43(e) and Part 50 Appendix B,Section III Design Control, This section provides descriptions for SR SSCs. These descriptions include the specific design that require analysis, appropriate test programs, features for SR SSCs that are responsible for meeting the SRDC and fulfilling their RSFs to experience, or a combination thereof, or a prototype facility, before a DC, COL, ML, OL, or SDA license mitigate the DBAs. This description should include features that demonstrate system capability approval. The granting of a license for a design that uses SR and reliability for both prevention and mitigation of LBEs, as applicable. SCCS that do not meet the above requirements would only be possible through NRC approval, and through the issuance of a license condition to assure the necessary analysis, 6.4.1 Description for SR SSC-1 appropriate test programs, experience, or a combination thereof, or a prototype facility are completed prior to This description should include: timelines stipulated in the license. This TICAP text needs to be revised to reflect these requirements.

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  • Narrative design descriptions that address the design aspects relevant to the performance of the RSFs, including:

o The SSC purpose in the context of supporting the RSFs o The specific SSC function in the context of supporting the RSFs o SSC materials and construction o SSC location and environmental conditions during normal operation and in the performance of the RSFs during LBEs o Key design features relevant to the performance of RSFs o Seismic and industry (e.g., ASME and IEEE) code classifications and the design codes applicable to the SR SSC o A description of all modes of SSC operation, including an account of the performance modes of SSC operation relevant to the RSFs o Identification of operator actions needed to implement the RSFs o Controls and displays needed to accomplish RSFs o Logic circuits and interlocks needed to support RSFs o Electric power, support systems, and interface requirements needed to support the RSFs o Equipment to be qualified for harsh environments as needed to meet SR SSC special treatment requirements defined in Section 6.2 o Brief summaries of first-of-a-kind special treatment tests to be performed (if any) o Cross reference to Chapter 8 where applicable programs to implement the special treatments are identified 6.4.2, 6.4.3, et al.: Descriptions of the remainder of the SR SSCs are provided.

Two-Step Licensing (CP Content)

For a CP application, the classification of SSCs from Chapter 5 is preliminary. However, the approach and methodologies to be used in developing the FSAR Chapter 6 content should be clearly described. The description should include any consensus codes and standards used or expected to be used in the design of SR SSCs. The descriptions for SR SSCs should be provided at a functional level and should identify the performance-based requirements needed for individual major components. Any safety-related first-of-a-kind components SSCs should be identified, as should plans for component SSC performance validation and acceptance criteria.

The guidance for other SSC description content in Chapter 6 should be tailored to the information available at the CP stage. The preliminary results from Chapter 5 should be used to frame the development of SRDC and special treatment requirements for SR SSCs. The content of Chapter 6 should use the tabular format provided in Chapter 6 of the COL guidance.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 7 NSRST SSC CRITERIA AND CAPABILITIES In Section 5.5, the NSRST SSCs were identified. Chapter 7 provides further detail on the role of each NSRST SSC in the LMP-based affirmative safety case, consistent with the NEI 18-04 methodology. Complementary design criteria for NSRST SSCs are covered in Section 5.5.4. The remaining criteria and capabilities for NSRST SSCs include reliability and capability performance targets and special treatment requirements to provide sufficient confidence that the performance targets will be maintained throughout the life of the licensed plant. Anything beyond procuring commercial-grade equipment that is done to provide increased assurance in the capability and reliability of the SSC falls into the category of special treatment.

As noted in Section 4.4.5 of NEI 18-04, the selection of special treatments for all safety-significant SSCs (SR and NSRST) are informed by a set of targets for the reliability and availability of the SSCs in their prevention functions as well as targets for the capability of the SSCs in the performance of their mitigation functions. These specific targets are addressed in Section 7.1, and the special treatments themselves are covered in Section 7.2.

This guidance assumes that the applicant followed the NEI 18-04 methodology for the establishment of SR SSC criteria and capabilities. If the applicant did not use parts of the methodology or used an alternative method, the following statement should be added:

The following deviations from the approved methodology were taken.

The applicant should go on to list each deviation, describe any alternative method, and provide justification for the approach employed.

7.1 Reliability and Capability Targets for NSRST SSCs Background information for SR SSCs in Section 6.2 is applicable to NSRST SSCs as well and is not repeated here. The treatment of targets is the same for NSRST and SR SSCs.

SSC level targets should be documented in Section 7.1. Except for the simplest of designs, Commented [A96]: Is this contradicting with the reliability and capability targets at every component the SSC level could will be quite discussion below, which states SSC-level reliability and capability targets will instead be maintained in plant voluminous and challenging for incorporation directly in the SAR. Therefore, it is expected the records SSC-level reliability and capability targets will instead be maintained in plant records as IBR material. In those records, each SSC should specify the reliability and capability targets as well Commented [A97]: See previous comment - that is this as the LBE(s) that the SSC prevents or mitigates. Describe NSRST SSC reliability targets and comment is consistent with the SR SSC comment above performance requirements used as input to the PRA for SSCs that were used to develop the selection of special treatment requirements (i.e., programmatic actions used to maintain performance within the design reliability targets). This description should include:

  • numerical targets for SSC reliability and availability,
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Content for a Licensing Modernization Project-Based Affirmative Safety Case requirements for monitoring of performance against these targets with appropriate corrective actions when targets are not fully realized.

In addition to the target information or associated reference, Section 7.1 should specify the plant program(s) used to maintain the reliability and capability targets. Information on the program(s) should be provided in Chapter 8.

7.2 Special Treatment Requirements for NSRST SSCs This section documents the special treatment requirements for NSRST SSCs.

NEI 18-04 adopted the definition of special treatment provided in RG 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance,19 which was developed for implementing 10 CFR 50.69: Commented [A98]: Editorial: RG 1.201 was quoted under Section 6.3 earlier. The footnote below should be placed special treatment refers to those requirements that provide increased assurance there.

beyond normal industrial practices that structures, systems, and components (SSCs) perform their design-basis functions.

Anything that is done beyond procuring commercial-grade equipment to provide increased assurance in the capability and reliability of SSCs falls into the category of special treatment.

Section 6.1 identified non-safety-related SSC design requirements associated with protecting SR SSCs from design basis external hazards. Hence, if a non-safety-related SSC identified in Section 6.1.3 is also an NSRST SSC, then all the design requirements provided in Section 6.1 are part of the NSRSTs special treatment. This section identifies the additional special treatments that are applied to NSRST SSCs.

For the selected special treatments, the license application should identify the treatments in the SAR with details available in the design records available for NRC audits. This section should describe, as applicable, the special treatment requirements from NEI 18-04, Table 4-1, on a case-by-case basis and in the context of the SSC functions in the prevention and mitigation of applicable LBEs. Describe special treatments including the following, as applicable:

  • Equipment qualification
  • Seismic qualification
  • Materials qualification
  • Pre-service and risk-informed in-service inspections
  • Pre-op and startup testing requirements 19 Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, U.S. Nuclear Regulatory Commission, May 2006. (ML061090627) 66

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  • Equipment testing requirements including test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met (i.e.,

demonstrate the ability to perform the risk-significant function).

The special treatments should be summarized in tabular form (see Table 7-1) by listing each NSRST SSC, providing a brief functional description, and identifying which special treatment has been selected for each NSRST SSC.

Some plant designs may involve first-of-a-kind testing for NSRST SSCs. To the extent those first-of-a-kind tests have not been completed by the time of the application, they may be included Commented [A99]: See comment in 6.3 on SR SSC as special treatments. special treatments Commented [A100]: See previous comment re:

Table 7-1. Example Table of NSRST SSC Special Treatments considering FOAK testing in ITAAC and the previous comment on the same statement for SR SSCs NSRST SSC Functional Description NSRST SSC Special Treatments NSRST SSC1 Special Treatment No. 1 Short SSC functional NSRST SSC1 Special Treatment No. 2 NSRST SSC1 description for NSRST SSC1 NSRST SSC1 Special Treatment No. n Additional NSRST SSCs 7.3 System Descriptions for NSRST SSCs This section provides descriptions for NSRST SSCs. These descriptions include the specific design features for NSRST SSCs that are responsible for meeting their safety-significant functions identified in the LBEs responsible for the classification as NSRST. This description should include features that demonstrate SSC capability and reliability for both prevention and mitigation of LBEs, as applicable. It is expected that these SSC descriptions will be less detailed than those provided for SR SSCs in Section 6.3, even though the list of covered items is similar.

The bulleted list in 7.3.1 is very similar to the list in 6.4.1 for SR SSCs. The NSRST SSC information does not need to be as comprehensive and detailed as the SR SSC information.

7.3.1 Description for NSRST SSC-1 This description should include:

  • A simplified schematic diagram
  • Narrative design descriptions that address the design aspects relevant to the performance of the safety-significant functions including, as applicable:

o The SSC purpose in the context of supporting the safety-significant functions o Significant functional performance-based characteristics in performing safety-significant functions o SSC materials and construction 67

Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

Content for a Licensing Modernization Project-Based Affirmative Safety Case o SSC location and environmental conditions o Key design features relevant to the performance of safety-significant functions o Seismic and industry (ASME, IEEE, etc.) code classifications and the design codes applicable to the NSRST SSC o Description of all modes of SSC operation, including a description of the performance modes of SSC operation relevant to the safety-significant functions o Identification of any operator actions needed to implement safety-significant functions o Controls and displays needed to support safety-significant functions o Logic circuits and interlocks needed to support safety-significant functions o Electric power, support systems, and interface requirements needed to support the safety-significant functions o Equipment to be qualified for harsh environments as needed to meet NSRST SSC special treatment requirements defined in Section 7.2 o Brief summaries of first-of-a-kind special treatment tests to be performed (if any) o Cross reference to Chapter 8 where applicable programs to implement the special treatments are identified 7.3.2, 7.3.3, et al.: Descriptions of the remainder of the NSRST SSCs are provided.

Two-Step Licensing (CP Content)

For a CP application, Chapter 7 content addressing NSRST SSCs should follow the approach used in Chapter 6 for SR SSCs. A preliminary list of NSRST SSCs should be populated to the extent the DID evaluation has been performed and described in Chapter 4 and should use the tabular format provided in Chapter 7 of the COL guidance. Descriptions for NSRST SSCs as described in Chapter 7 should be developed to identify safety-significant functions to be provided by those SSCs.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case 8 PLANT PROGRAMS The guidance in this chapter is not intended to address all plant programs. It focuses on those that play a substantive role in the LMP-based affirmative safety case. For example, nuclear power reactors require a radiological protection program for plant workers in order to ensure compliance with occupational dose regulations in 10 CFR Part 20. However, the radiological protection program for plant workers is unlikely to be a component of the affirmative safety case, which is based on protection of the public from radiological hazards. Hence, the radiological protection program is not described in SAR Chapter 8 but is provided in another part of the application.

The set of programs in Chapter 8 includes those used for special treatments for SR or NSRST SSCs as described in Chapters 6 and 7, respectively, that provide reasonable assurance (i) reliability and performance targets are achieved throughout the plant lifetime and (ii) safety- Commented [A102]: See above significant uncertainties are effectively addressed as part of DID. In addition to programs supporting special treatments, this chapter should also identify and provide an overview of the program or programs that document SSC reliability and capability targets, as described in Sections 6.2 and 7.1. The discussion of plant programs should address the different plant lifetime phases, i.e., design, construction, testing, and operations, as applicable.

Examples of possible program topics are provided below. The actual program topics are determined by the special treatments documented in Chapters 6 and 7. Not all of the examples may be required for a given application, and additional program topics not listed below may be required.

  • Quality assurance programs (design, construction, operations)
  • Startup testing
  • In-service testing
  • In-service inspection
  • Equipment qualification
  • Performance monitoring
  • Reliability assurance The applicant should provide the following information in this chapter of the SAR:

Commented [A103]: Putting the controlling document revision number and date in the SAR means a SAR change

  • Program topic and summary scope description every time the document is changed. Not necessary unless
  • SSC class applicability (SR, NSRST, or NST) the document is specifically approved by the NRC
  • Controlling program document including title, document number, revision number, and Commented [A104]: Are there examples of controlling date. The expectation is that each program supporting the LMP-based affirmative safety documents that should be maintained by the design authority following commencement of operation?

case will have an associated controlling document maintained by the design authority or Ultimately, I think, at some point in time, all controlling licensee. This document could be a topical report that has been reviewed and approved documents should rest with the licensee and be maintained by the licensee. I don't think that there should a point in by NRC or an internal document.

time beyond the commencement of operations, for example, that there should shared possession and maintenance of controlling documents.

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  • Standard industry program references, if any, used as a basis for a program, e.g.,

standards or Nuclear Energy Institute guidance documents (For each reference, it should be noted whether it is incorporated by reference [IBR] or provided for information [Ref]

as described in Section 1.4.) Where a standard industry program reference is not utilized, the SAR program description should address similar level of program detail, including:

  • Scope and purpose
  • Process used for identifying SSCs included in the programs scope
  • Organization controls
  • Controls for procedures and instructions
  • Quality assurance requirements including program audits and corrective actions
  • Elements specified in applicable regulations or regulatory guidance [e.g., for the maintenance program, 10 CFR 50.65, 10 CFR 52.79(a)(15), and Regulatory Guide 1.60, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants]
  • In the event of any significant deviations from an IBR standard industry program, identification of the deviation and associated explanation The applicant has the option of providing the information in this chapter in a tabular format or in text. For the former, an example table with entries for programs is provided in Table 8-1Table 8-1. (Note that the example entries are meant to be illustrative, representative content for a specific program and not the comprehensive program description.) Commented [A105]: The text here doesnt provide sufficient guidance regarding what information about the Table 8-1. Example Table of Special Treatment Programs program needs to be provided I the SAR if an approved template is not utilized. See added text above.

SSC Applicability Program Topics / Objectives Program References and Deviations IBR Ref SR NSRST NST Quality Assurance (QA) Program Insert controlling program document The QA Program comprises the planned and information (e.g., XYZ123, Quality systematic activities implemented to Assurance Topical Report, Rev. 0, July 1, demonstrate that SSCs and activities will 2025).

fulfill their requirements. It provides ASME NQA-1, Quality Assurance assurance that the actual plant configuration Requirements for Nuclear Facility is consistent with the capabilities modeled in Applications, 2019.

the licensing basis event analyses.

Depending on the requirements associated Deviation: The Quality Assurance Program is based with the specific SSC, the full range of QA on NQA-1, but a graded approach is applied as controls may be applied, or a subset thereof. described in the Quality Assurance Topical Report.

Insert controlling program document information (e.g., ABC789, Reliability Assurance Program Rev. 0, July 1, 2023).

Reliability Assurance Program NUREG-0800 Section 17.4, Reliability The plant reliability assurance program Assurance Program Rev 0, 2007Revision controls the reliability and availability 1, 2014.

targets of SR SSCs and NSRST SSCs consistent ASME Section XI-Rules for Inservice with the PRA.

Inspection of Nuclear Power Plant Components, Division 2, Requirements for Reliability and Integrity Management 70

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Content for a Licensing Modernization Project-Based Affirmative Safety Case SSC Applicability Program Topics / Objectives Program References and Deviations IBR Ref SR NSRST NST (RIM) Programs for Nuclear Power Plants, 2019.

Deviation: NUREG-0800 Section 17.4 applies to SR SSCs only as described in the Reliability Assurance Program document.

Insert controlling program document information (e.g., LMN456-0, Plant Maintenance Program, June 9, 2021).

NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Maintenance Program Revision 4F, Nuclear Energy Institute, The plant maintenance controls the April 2018.

performance of maintenance activities in the Regulatory Guide 1.160, Revision 4, plant. It provides assurance that the actual Monitoring the Effectiveness of plant configuration is consistent with the Maintenance at Nuclear Power Plants, capabilities modeled in the licensing basis August 2018.

event analyses. NEI 07-02A Generic FSAR Template Guidance for Maintenance Rule Program Description for Plants Licensed Under 10 CFR Part 52, (Rev 0) March 2008 Deviations: None.

If the table approach is used, the rest of the guidance in Chapter 8 is not applicable. If a text- Commented [A106]: Same comment as above; if a based approach is used instead of a table, the applicant would be expected to use a section for template is not referenced then the table format may not provide sufficient program description each program topic. An example is provided below.

8.1 Maintenance Program The plant maintenance program controls the performance of maintenance activities in the plant.

It provides assurance that the actual plant configuration is consistent with the capabilities modeled in the licensing basis event analyses. The program applies to SR, NSRST, and NST SSCs. The program is administered in accordance with applicant document LMN456-0, Plant Maintenance Program, Rev. 0, June 9, 2021, which is IBR. Commented [A107]: Same comment as above; dont think this is necessary or appropriate.

NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,20 (Ref) documents acceptable practices for implementation of the Maintenance Rule (10 CFR 50.65). NRC Regulatory Guide 1.160, Monitoring the 20 NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Rev 4F, Nuclear Energy Institute, April 2018.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case Effectiveness of Maintenance at Nuclear Power Plants21 (Ref) endorses NUMARC 93-01 and provides additional guidance. NEI 07-02A, Generic FSAR Template Guidance for Maintenance Rule Program Description for Plants Licensed under 10 CFR Part 52,22 (Ref) provides additional guidance. The guidance documents were developed for light water reactors and therefore informed but did not determine the maintenance program for the ______ reactor.

There are no deviations from the referenced documents.

8.2 Other Programs Other programs are described in Sections 8.2, 8.3, etc.

Two-Step Licensing (CP Content)

The plant program descriptions for the PSAR should be largely the same as for the FSAR content in Chapter 8 of the COL guidance. The PSAR should contain complete identification of design, Commented [A108]: Do not think the PSAR needs to manufacturing, and construction programs to be used and any references to standardized industry have the same level of program detail as the FSAR.

Providing commitments to develop the program programs that are used as frameworks or templates for these programs. Similarly, the CP should descriptions at the OL stage should be sufficient.

contain general descriptions of the operational programs that may be incorporated in the FSAR to support the affirmative safety case. The PSAR content for Chapter 8 should use the tabular format provided in Chapter 8 of the COL guidance, where any preliminary information is clearly identified.

21 Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Rev 4, U.S. Nuclear Regulatory Commission, August 2018.

22 NEI 07-02A, Generic FSAR Template Guidance for Maintenance Rule Program Description for Plants Licensed Under 10 CFR Part 52, Rev 0, Nuclear Energy Institute, March 2008.

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Content for a Licensing Modernization Project-Based Affirmative Safety Case D.

SUMMARY

The guidance contained in Section C of this report is one acceptable means of providing content for the portion of an advanced reactor SAR related to the application of the RIPB LMP-based affirmative safety case methodology described in NEI 18-04.

The Section C guidance applies to a COL under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The guidance further assumes that the applicant is not referencing an existing Design Certification. Section C also provides modifications to the baseline COL guidance for two alternative licensing approaches:

  • Two-step licensing (CP/OL)The applicant first applies for and obtains a CP under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and subsequently applies for and obtains an OL also under 10 CFR Part 50.
  • Design certificationThe applicant is a reactor vendor that applies for a standard design certification under 10 CFR Part 52, Subpart B. It does not contain site-specific information. A future applicant would reference the design certification along with site-specific and owner-specific information as part of a COL application.

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Appendix A Glossary of Terms Appendix AGlossary of Terms Term Acronym Definition Source Anticipated event sequences expected to occur one or more times during the life of a Anticipated nuclear power plant, which may include one or more reactors. Event sequences with NEI 18-04 Operational AOO mean23 frequencies of 1x10-2/plant-year and greater are classified as AOOs. AOOs take Occurrence into account the expected response of all SSCs within the plant, regardless of safety Modified for TICAP Commented [A109]: It may be helpful to explain this in this table. It is used multiple times. It appears the only classification.

difference is between reactors vs reactor modules.

Rare event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactors, but are less likely than a DBE. Event sequences NEI 18-04 Beyond Design BDBE with mean231 frequencies of 5x10-7/plant-year to 1x10-4/plant -year are classified as Basis Event Modified for TICAP BDBEs. BDBEs take into account the expected response of all SSCs within the plant regardless of safety classification.

Design criteria for NSRST SSC that are necessary to satisfy the PRA Safety Function(s) associated with the SSC. The CDC may be defined at a functional level, or more specifically Complementary CDC addressed to the NSRST SSC specific function(s). The CDC for the NSRST SSC are directly TICAP Design Criteria tied to the success criteria established in the PRA for the PRA Safety Function(s) responsible for the classification of the SSC as NSRST.

An approach to designing and operating nuclear facilities that prevents and mitigates accidents that release radiation or hazardous materials. The key is creating multiple Defense-in- independent and redundant layers of defense to compensate for potential human and DID NRC Glossary Depth mechanical failures so that no single layer, no matter how robust, is exclusively relied upon. Defense-in-depth includes the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures.

23 The classification of AOOs, DBEs, and BDBEs is based on the mean frequencies of the underlying uncertainty distributions. When the uncertainty band on the frequency defined by the 95th and 5th percentiles of the distribution straddle one of the frequency boundaries, the LBEs are evaluated on each side of the boundary, per NEI 18-04. For example, if a BDBE has a 95th percentile estimate above 1x10-4 per plant year, it is treated as a DBE for the purposes of defining the RSFs and defining the DBAs.

A-1

Appendix A Glossary of Terms Term Acronym Definition Source Postulated accidents that are used to set design criteria and performance objectives for the design of SR SSCs. DBAs are derived from DBEs based on the capabilities and Design Basis DBA reliabilities of SR SSCs needed to mitigate and prevent accidents, respectively. DBAs are NEI 18-04 Accident derived from the DBEs by prescriptively assuming that only SR SSCs classified are available to mitigate postulated accident consequences to within the 10 CFR 50.34 dose limits.

Infrequent event sequences that are not expected to occur in the life of a nuclear power plant, which may include one or more reactors, but are less likely than AOOs. Event Design Basis sequences with mean24 frequencies of 1x10-4/plant-year to 1x10-2/plant-year are classified NEI 18-04 DBE Event as DBEs. DBEs take into account the expected response of all SSCs within the plant Modified for TICAP regardless of safety classification. The objective and scope of DBEs form the safety design basis of the plant.

Design Basis A design specification of the level of severity or intensity of an external hazard for which External Hazard DBEHL the SR SSCs are designed to withstand with no adverse impact on their capability to NEI 18-04 Level perform their RSFs The set of conditions at the end of an event sequence that characterizes the impact of the End State -- sequence on the plant or the environment. In most PRAs, end states typically include ASME/ANS-RA-S-1.4-2021 success states (i.e., those states with negligible impact) and release categories. Commented [A110]: Some font colors were white.

Changed to Auto here and below.

A representation of a scenario in terms of an initiating event defined for a set of initial plant conditions (characterized by a specified plant operating state) followed by a sequence of system, safety function, and operator failures or successes, with sequence Event Sequence ES termination with a specified end state (e.g., prevention of release of radioactive material ASME/ANS-RA-S-1.4-2021 or release in one of the reactor-specific release categories). An event sequence may contain many unique variations of events (minimal cutsets) that are similar in how they impact the performance of safety functions along the event sequence.

24 The AOO and BDBE footnote on mean frequencies (see prior page) applies here as well..

A-2

Appendix A Glossary of Terms Term Acronym Definition Source A grouping of event sequences with similar challenges to the plant safety functions, response of the plant in the performance of each safety function, response of each radionuclide transport barrier, and end state. An event sequence family may involve a single event sequence or several event sequences grouped together. Each release Event Sequence category may include one or more event sequence families. When event sequence

- ASME/ANS-RA-S-1.4-2021 Family models are developed in great detail, identification of families of event sequences with common or similar source, initiating event and plant response facilitates application of the event sequence modeling requirements in this Standard and development of useful risk insights in the identification of risk contributors. Each event sequence family involving a release is associated with one and only one release category.

Frequency- A target line on a frequency-consequence chart that is used to evaluate the risk F-C Consequence significance of LBEs and to evaluate risk margins that contribute to evidence of adequate NEI 18-04 Target Target Defense-in-Depth Fundamental Safety functions common to all reactor technologies and designs; includes control heat FSF IAEA-TECDOC-1570 Safety Function generation, control heat removal and confinement of radioactive material A perturbation to the plant during a plant operating state that challenges plant control and safety systems whose failure could potentially lead to an undesirable end state and/or radioactive material release. An initiating event is defined in terms of the change in plant status that results in a condition requiring a response to mitigate the vent or to limit the Initiating Event IE ASME/ANS-RA-S-1.4-2021 extent of plant damage caused by the initiating event. An initiating event may result from human causes, equipment failure from causes internal to the plant (e.g., hardware faults, flood, or fires) or external to the plant (e.g., earthquakes or high winds), or combinations thereof.

Integrated Risk-informed and performance-based integrated decision-making (RIPB-DM) process Decision-Making IDP TICAP (based on NEI 18-04) used for establishing special treatments and evaluating the adequacy of DID.

Process A-3

Appendix A Glossary of Terms Term Acronym Definition Source Layers of defense are those plant capabilities and programmatic elements that provide, Layers of collectively, independent means for the prevention and mitigation of adverse events. The

-- NEI 18-04 Defense actual layers and number are dependent on the actual source and hazard posing the threat. See Defense-in-Depth.

The entire collection of event sequences considered in the design and licensing basis of NEI 18-04 Licensing Basis LBE the plant, which may include one or more reactors. LBEs include AOOs, DBEs, BDBEs, and Event Modified for TICAP DBAs.

The characteristics of a radionuclide release at a particular location, including the physical and chemical properties of released material, release magnitude, heat content (or energy) of the carrier fluid, and location relative to local obstacles that would affect transport Mechanistic away from the release point and the temporal variations in these parameters (e.g., time of MST ASME/ANS-RA-S-1.4-2021 Source Term release duration, etc.) that are calculated using models and supporting scientific data that simulate the physical and chemical processes that describe the radionuclide inventories and the time-dependent radionuclide transport mechanisms that are necessary and sufficient to predict the source term.

An SSC function that, if fulfilled, will eliminate or reduce the consequences of an event in Mitigation which the SSC function is challenged. The capability of the SSC in the performance of such

-- NEI 18-04 Function functions serves to eliminate or reduce any adverse consequences that would occur if the function were not fulfilled.

Non-Safety-Related with NSRST Non-safety-related SSCs that perform risk-significant functions or perform functions that NEI 18-04 Special SSCs are necessary for Defense-in-Depth adequacy Treatment SSCs Non-Safety-Related with No All SSCs within a plant that are neither SR SSCs nor Non-Safety-Related with Special NST SSCs NEI 18-04 Special Treatment SSCs Treatment SSCs A-4

Appendix A Glossary of Terms Term Acronym Definition Source An approach to decision-making that focuses on desired objective, calculable or Adapted from NRC Glossary measurable, observable outcomes, rather than prescriptive processes, techniques, or definition of performance-procedures. Performance-based decisions lead to defined results without specific based regulation (page Performance-PB direction regarding how those results are to be obtained. At NRC, performance-based updated March 9, 2021) in Based regulatory actions focus on identifying performance measures that ensure an adequate order to apply to both safety margin and offer incentives and flexibility for licensees to improve safety without design decisions and formal regulatory intervention by the agency. regulatory decision-making The collection of site, buildings, radionuclide sources, and SSCs seeking a single design NEI 18-04 Plant -- certification or one or more OLs under the LMP framework. The plant may include a single reactor unit or multiple reactor units as well as non-reactor radionuclide sources. Modified for TICAP A standard arrangement of the plant during which the plant conditions are relatively constant, are modeled as constant, and are distinct from other configurations in ways that impact risk. Plant operating state is a basic modeling device used for a phased-mission risk assessment that discretizes the plant conditions for specific phases of plant evolution.

Plant Operating POS Examples of such plant conditions include core decay heat level, reactor coolant level, ASME/ANS-RA-S-1.4-2021 State coolant temperature, coolant vent status, reactor building status, and decay heat removal mechanisms. Examples of risk impacts that are dependent on the plant operating state definition include the selection of initiating events, initiating event frequencies, definition of event sequences, success criteria, and event sequence quantification.

Reactor design specific SSC functions modeled in a PRA that serve to prevent and/or PRA Safety mitigate a release of radioactive material or to protect one or more barriers to release. In NEI 18-04, PSF Function ASME/ANS-Ra-S-1.4-2013 these are referred to as safety functions. The modifier PRA is ASME/ANS-RA-S-1.4-2021 used in NEI 18-04 to avoid confusion with safety functions performed by SR SSCs.

An SSC function that, if fulfilled, will preclude the occurrence of an adverse state. The Prevention

-- reliability of the SSC in the performance of such functions serves to reduce the probability NEI 18-04 Function of the adverse state.

A-5

Appendix A Glossary of Terms Term Acronym Definition Source Required Reactor design-specific functional criteria that are necessary and sufficient to meet the Functional RFDC NEI 18-04 RSFs Design Criteria A PRA Safety Function that is required to be fulfilled to maintain the consequence of one Required Safety RSF or more DBEs or the frequency of one or more high-consequence BDBEs inside the F-C NEI 18-04 Function Target Adapted from NRC Glossary definition of risk-informed regulation (page updated An approach to decision-making in which insights from probabilistic risk assessments are Risk-Informed RI March 9, 2021) in order to considered with other sources of insights apply to both design decisions and regulatory decision-making An LBE whose frequency and consequence meet a specified risk significance criterion. In the LMP framework, an AOO, DBE, or BDBE is regarded as risk-significant if the Risk-Significant

-- combination of the upper bound (95th percentile) estimates of the frequency and NEI 18-04 LBE consequence of the LBE are within 1% of the F-C Target AND the upper bound 30-day TEDE dose at the EAB exceeds 2.5 mrem.

An SSC that meets defined risk significance criteria. In the LMP framework, an SSC is regarded as risk-significant if its PRA Safety Function is: a) required to keep one or more LBEs inside the F-C Target based on mean frequencies and consequences; or b) if the total Risk-Significant frequency LBEs that involve failure of the SSC PRA Safety Function contributes at least 1%

-- NEI 18-04 SSC to any of the LMP cumulative risk targets. The LMP cumulative risk targets include: (i) maintaining the frequency of exceeding 100 mrem to less than 1/plant-year; (ii) meeting the NRC safety goal QHO for individual risk of early fatality; and (iii) meeting the NRC safety goal QHO for individual risk of latent cancer fatality.

Safety-Related Design criteria for SR SSCs that are necessary and sufficient to fulfill the RFDC for those SRDC NEI 18-04 Design Criteria SSCs selected to perform the RSFs A-6

Appendix A Glossary of Terms Term Acronym Definition Source Safety-Related SSCs that are credited in the fulfillment of RSFs and are capable to perform their RSFs in SR SSCs NEI 18-04 SSCs response to any Design Basis External Hazard Level Safety- An SSC that performs a function whose performance is necessary to achieve adequate

-- NEI 18-04 Significant SSC Defense-in-Depth or is classified as risk-significant (see Risk-Significant SSC)

Note: Definitions from the non-LWR PRA Standard ASME/ANS-RA-S-1.4-2021 are excluded pending permission from ASME. Commented [A111]: Note that this yellow highlighted material was identified by Southern not the NRC A-7

Appendix B Example LBE Description Appendix B - Example LBE Descriptions This appendix contains example descriptions of four LBEs - an AOO, a DBE, a BDBE, and a DBA. The intent of the examples is to show how the LBEs could be described consistent with the guidance provided in Chapter 3 of this guidance document. While the examples are based on existing material as indicated for each LBE, they are illustrative only and are not intended to be used to obtain regulatory approval for any past or present reactor design. No pedigree is claimed for the technical results that are presented. In some cases, the information in the examples has been altered or embellished for illustrative purposes.

In some cases, the example descriptions refer to other portions of a SAR (e.g., Chapter 2 for source term information). The purpose of the references is to show how the applicant might refer to information elsewhere in the SAR. Those other referenced portions of the SAR are not included herein as part of the examples.

The nomenclatures for sections and figures (e.g., AOO-N, Section 3.3.N, and Figure 3.3.N-1 immediately below) are arbitrary but consistent with the Chapter 3 guidance.

B.1 Example AOO Description This writeup is based on a fuel pump failure in an experimental molten salt reactor, described in the Molten Chloride Reactor Experiment (MCRE) tabletop exercise report.25 This AOO does not produce a radionuclide release, so the example writeup actually goes beyond the guidance in Section 3.3.1 with respect to detail provided. Specifically, the plots and event frequencies are not required by the guidance for this type of AOO, but the applicant may choose to provide such information.

3.3.N Fuel Pump Failure (AOO-N)

Failure of the MCRE fuel pump can occur for a variety of reasons such as loss of electrical power or a mechanical degradation of components. The fuel pump failure occurs while the plant is operating at nominal full power conditions (see Table 2-X).26 Both the Primary Cooling System and Heat Removal System continue operating during the transient. The failure of the fuel pump results in a loss of forced convection and lower fuel salt flow rate in the Primary Cooling System. The core outlet fuel salt temperature increases as natural circulation flow is established at approximately five per cent of nominal fuel salt flow with the fuel pump operating.

The core inlet fuel salt temperature decreases due to continued operation of the Heat Removal System. The initial increase in average core fuel salt temperature causes a decrease in core fuel salt density, providing negative reactivity which reduces the reactor power. In a countervailing effect, the decreased core fuel salt flow rate provides positive reactivity because delayed neutron precursors remain in the core region longer instead of being removed quickly by forced flow (see discussion under Section 3.3.1, Loss of Offsite Power). However, the negative reactivity effect 25 Reference to be added when tabletop report finalized.

26 For the purpose of this illustrative example, it is assumed that Chapter 2 of the SAR contains a section describing common attributes of LBE analyses, including nominal fuel power initial conditions.

B-1

Appendix B Example LBE Description from the density increase is significantly larger than the delayed precursor residence time effect.

As natural circulation flow is established (Figure 3.3.N-1), the average core fuel salt temperature decreases again and core power goes back up toward the nominal level, as shown in Figure 3.3.N-2. Ultimately, the system returns to a safe, stable end state of steady-state power under natural circulation conditions.

Figure 3.3.N-1. Fuel Salt Mass Flow Rate Figure 3.3.N-2. Core Power No reactor trip setpoints (reactor power or fuel salt temperature) are exceeded, so the reactor continues to operate. No operator actions or safety system actuations are required to mitigate the event. Fuel salt temperatures are shown in Figures 3.3.N-3 and 3.3.N-4. The highest temperature reached by the fuel salt is below 920 K, well below the upper temperature limit of around 980 K, which ensures reactor vessel integrity. The lowest temperature of the fuel salt is above 800 K, well above the freezing temperature of 726 K. The event does not result in a release of radionuclides to the accessible environment.

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Appendix B Example LBE Description Figure 3.3.N-3. Fuel Salt Core Temperature Figure 3.3.N-4. Fuel Salt Temperatures at the Outlet and Inlet of the Heat Exchanger The analysis was performed as part of the plant PRA, which is summarized in Section 2.1. The estimated frequency of the event is 5x10-2 per plant year with 5th and 95th percentile uncertainty values of 3x10-2 per plant year and 8x10-2 per plant year, respectively. There are no dose consequences. Note that the frequency values are illustrative only and are not based on an actual evaluation.

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Appendix B Example LBE Description B.2 Example DBE Description This writeup is based on a moisture inleakage event for a modular high-temperature gas reactor, with some information taken from the MHTGR PSID Section 15.8.27 For the purpose of this example, it is assumed that this is the limiting DBE for the associated DBA, so the complete writeup described in Section 3.4.1 is provided.

3.4.N Moisture Inleakage without Shutdown Cooling System Cooling (DBE-N)

The steam generator is the most likely source of potential moisture inleakage into the MHTGR Primary Coolant System. Steam generator tube leakage can occur for multiple reasons including a manufacturing defect exacerbated by corrosion or wear. Moisture ingress into the primary coolant system can result in a reactivity increase, exposed defective fuel particle hydrolysis, and graphite oxidation.

The event initiates from nominal full power conditions (see Table 2-X).28 A double offset steam generator tube rupture is assumed, bounding the flow rate for this event sequence family (initial leak rate of 5.7 kg/sec). The Moisture Detection System detects the leak, initiates a reactor trip, Commented [A112]: Should include settings of all isolates the steam generator from feedwater and steam flow, dumps steam generator water protection system functions that are used in the DBE evaluation?

inventory, shuts down the Heat Transport System, and starts the Shutdown Cooling System (SCS) to provide active core cooling. However, the SCS does not function properly, so core heat Commented [A113]: what mechanism?

removal is accomplished by conduction and thermal radiation from the fuel particles through the core to the reactor vessel walls and radiation and convection from the reactor vessel to the reactor cavity that is cooled by the passive Reactor Cavity Cooling System (RCCS). Commented [A114]: Values?

Core power increases initially due to positive reactivity from the moisture ingress, an effect that is somewhat mitigated by negative temperature feedback in the core. Power decreases rapidly upon reactor trip. From that point onward, the reactor power is equal to the decay heat power level (see Figure 3.4.N-1).

27 Preliminary Safety Information Document for the Standard MHTGR, DOE-HTGR-86-024, U.S. Department of Energy, September 1988.

28 For the purpose of this illustrative example, it is assumed that Chapter 2 of the SAR contains a section describing common attributes of LBE analyses, including nominal fuel power initial conditions.

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Appendix B Example LBE Description Figure 3.4.N-1. Balance Between Decay Heat and RCCS Heat Removal The failure of the SCS results in a gradual heat-up of the core due to decay heat. As the core temperature increases, heat transfer from the reactor vessel to the RCCS becomes more effective. Commented [A115]: Assumptions used in the analysis?

At 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, the decreasing decay heat matches the increasing heat removal from the RCCS.

As shown in Figure 3.4.N-2, from that point onward, there is a slow cooldown to cold shutdown conditions. Core and metallic components reach their maximum temperatures at approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the beginning of the event. The maximum core temperature peaks at 1,274 °C.

The maximum control rod temperature reaches 863 °C, maximum core barrel temperature is 506 °C, and maximum vessel temperature is 392 °C. For the purpose of this analysis, the safe, stable end state is defined to be at approximately 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, at which point the average core Commented [A116]: Remain subcritical?

temperature has returned to its approximate value at the beginning of the event, RCCS heat removal exceeds core decay heat, and the reactor is slowly cooling toward cold shutdown conditions. Commented [A117]: Values? Time?

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Appendix B Example LBE Description Figure 3.4.N-2. Core Temperature Primary coolant pressure increases about 0.2 MPa following the rupture of the steam generator tube but decreases when the circulator trips and moisture ingress stops. As the core and structural material temperatures increase, the coolant pressure also increases to the point at which the pressure relief valve opens (7.1 MPa) at about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Note that the pressure does not reach the nominal setpoint, but the probability of valve lift due to setpoint drift falls within the DBE region. The relief valve reseats at 6.1 MPa and does not open again. Coolant pressure decreases thereafter as the plant is cooled, as shown in Figure 3.4.N-3.

Figure 3.4.N-3. Primary Coolant Pressure B-6

Appendix B Example LBE Description The total water inleakage is 267 kg. Tripping the Heat Transport System at the beginning of the event limits the amount of water inleakage that is available to oxidize graphite in the reactor vessel to 28 percent of the total inleakage, or 75 kg. The average fraction reaction in active core fuel elements is only 5.2x10-4 weight fraction. That value would increase to 1.8x10-3 weight fraction if all inleakage into the primary circuit were assumed to react. Bottom reflector blocks experience an average reaction of about 4x10-4 weight fraction. The reaction in core support materials (blocks and posts) is lower and does not significantly impair the core support capability of the structural materials.

With respect to the fuel, the high purity carbon is much less reactive than the bulk graphite, so there is no significant oxidation of the fuel. No fuel particle coatings should fail due to the steam in the environment. However, the small fraction of particles with defective coatings could be exposed to small amounts of water vapor diffusing through the graphite. 7.9 percent of iodine and noble gases in failed particles are released due to hydrolysis through the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, when hydrolysis is complete.

The inleakage event results in the release of fission products to the primary circuit through four mechanisms: (i) hydrolysis of UCO particles with failed coatings, (ii) liberation of sorbed fission products in the bulk moderator graphite which is oxidized, (iii) diffusion of fission products out of failed particles due to elevated temperatures, and (iv) steam-induced vaporization and recirculation of fission products plated out on metallic surfaces. Based on all mechanisms, releases to the primary coolant are 300 Ci of noble gas, 440 Ci of iodine, and 142 Ci metallics.

This compares to the nominal noble gas activity in the primary coolant of about 23 Ci.

The one-time relief through the primary system relief valve causes a release of 15 percent of the circulating gases and particulates. The total inventory in the primary coolant available for Commented [A118]: Basis?

release to the environment (by mechanism of release) and the total released at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is shown in Table 3.4.N-1.

Table 3.4.N-1. Fission Products in Primary Coolant and Released from Primary Coolant at 10 Hours Curies in Primary Coolant by Mechanism Initially Hydrolysis of Graphite Recirculation Elevated Total Released Nuclide Circulating Failed Fuel Oxidation of Plateout Temperatures (Ci) (Ci)

Kr-85m 2.30 6.5 -- -- 0.1 8.90 1.33 Kr-88 5.16 17.3 -- -- 5.78 28.2 4.24 Rb-88 0.07 -- -- 3.12 -- 3.19 0.48 Sr-89 -- -- 0.220 1.03 -- 1.25 0.19 etc.

Based on that release, mean exclusion area boundary doses are calculated to be only 4x10-5 rem TEDE over a 30-day period, with 5th and 95th percentile values of 1.9x10-5 rem and 2.0x10-4 rem, respectively. These values are orders of magnitude below the F-C Target in the DBE region. Commented [A119]: How was the Chapter 2 dose methodology applied here?

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Appendix B Example LBE Description The mechanistic source term is addressed in Section 2.J, and the dose analysis methodology is covered in Section 2.K.

The event is assumed to occur at any one of the four reactors on the plant site. The analysis was performed as part of the plant PRA, which is summarized in Section 2.1.

The estimated frequency of the event is 4x10-4 per plant year with 5th and 95th percentile uncertainty values of 1x10-4 per plant year and 8x10-4 per plant year, respectively. Note that the frequency values are not based on analyses but are provided for illustrative purposes only.

SSCs performing PRA safety functions to mitigate the event are discussed in the narrative above.

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Appendix B Example LBE Description B.3 Example BDBE Description This writeup is based on a loss of heat sink event without reactor trip for the PRISM reactor.

Information is taken from Appendix E.4 of the PRISM Preliminary Safety Information Document.29 This BDBE has no consequences so it does not bound the risks of a collection of BDBEs, and the example writeup with plots goes beyond the minimum content outlined in Section 3.5.1 of the guidance.

3.5.N Loss of Heat Sink without Reactor Trip (BDBE-N)

The PRISM reactor is assumed to undergo a loss of the Intermediate Heat Transfer System from beginning of cycle hot full power conditions (see Table 2-X)30 with a concurrent failure of the reactor control and protection systems to reduce power or trip the reactor. The event occurs with a beginning of cycle core configuration, which minimizes negative reactivity feedback.

The primary temperatures increase due to the loss of the primary heat sink. The core inlet region rapidly heats up to 980 °F. The resulting expansion of the radial gridplate provides substantial negative reactivity (approximately $0.68). The net reactivity remains negative throughout the event, shutting down the fission reaction as shown in Figure 3.5.N-1. Primary temperatures remain elevated as shown in Figures 3.5.N-2 and 3.5.N-3, with primary system heat input balanced by heat removal from the Reactor Vessel Auxiliary Cooling System and the Auxiliary Cooling System.

Figure 3.5.N-1. Core Power and Primary System Flow 29 GEFR-00793, PRISM Preliminary Safety Information Document, Volume IV, Appendix E, General Electric, December 1987.

30 For the purpose of this illustrative example, it is assumed that Chapter 2 of the SAR contains a section describing common attributes of LBE analyses, including nominal fuel power initial conditions.

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Appendix B Example LBE Description Figure 3.5.N-2. Reactivity Figure 3.5.N-3. Primary Coolant Temperature B-10

Appendix B Example LBE Description Fuel and core temperatures are shown on Figures 3.5.P-4 and 3.5.P-5. BDBE success criteria are documented in Section 2.R.31 For this event the applicable thermal criteria are summarized below. All criteria are met, so there are no dose consequences for the event.

  • Cladding midwall temperature less than 1450 °F (initial period).
  • Peak fuel-cladding interface temperature less than 1290 °F (long-term).
  • Peak fuel-cladding interface temperature less than 1450 °F (short-term).
  • Peak fuel centerline temperature less than 2000 °F.
  • Peak sodium temperature less than 1800 °F.

Figure 3.5.N-4. Peak Fuel-Cladding Midwall and Sodium Core Outlet Temperatures 31 For the purpose of this illustrative example, it is assumed that the BDBE success criteria are documented in Chapter 2.

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Appendix B Example LBE Description Figure 3.5.N-5. Fuel Temperatures The reactor has attained a safe, stable end state by 2000 seconds, with the core shutdown and primary system heat being removed by the Reactor Vessel Auxiliary Cooling System and the Auxiliary Cooling System.

The analysis was performed as part of the plant PRA, which is summarized in Section 2.1. The estimated frequency of the event is 9x10-7 per plant year with 5th and 95th percentile uncertainty values of 2x10-7 per plant year and 5x10-6 per plant year, respectively. Note that the aforementioned frequency values are illustrative and not based on actual PRA calculations for PRISM.

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Appendix B Example LBE Description B.4 Example DBA Description This writeup is based on a moisture inleakage event for a modular high-temperature gas reactor, with information taken from the MHTGR PSID Section 15.13.7.2730 Note that the PSID uses the Formatted: Superscript nomenclature Safety Related Design Conditions instead of DBAs.

3.6.N Moisture Inleakage Without SCS Cooling (DBA-N)

This MHTGR DBA is derived from the corresponding moisture inleakage DBE-N (see Appendix B.2). The initiating event is the same, but for the DBA it is assumed that only safety-related SSCs are available to mitigate the consequences.

The event initiates from nominal full power conditions (see Table 2-X).32 A double offset steam generator tube rupture is assumed with an initial leak rate of 5.7 kg/sec. Following the leak in the steam- generator, the resulting moisture ingress into the primary coolant causes a rapid increase in primary coolant moisture concentration and within a few seconds the moisture produces an increase in core reactivity. Core power rises such that the reactor trip setpoint on high core power-to-flow ratio of 1.5 is achieved at 8 seconds, at which time the outer reflector rods are tripped. Figure 3.6.N-1 shows the reactor power during this event.

Figure 3.6.N-1. Reactor Power 32 For the purpose of this illustrative example, it is assumed that Chapter 2 of the SAR contains a section describing common attributes of LBE analyses, including nominal fuel power initial conditions.

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Appendix B Example LBE Description Normally, the Moisture Detection System would detect the leak, initiate a reactor trip, isolate the steam generator from feedwater and steam flow, dump steam generator water inventory, shut down the Heat Transport System, and start the SCS to provide core cooling. However, for the DBA analysis the non-safety-related Moisture Detection system is assumed to be unavailable.

Also, the neutron flux controller should mitigate the reactor power increase at the beginning of the event by inserting control reactivity to maintain reactor power, but the non-safety-related controller is also assumed to be unavailable. Moreover, credit is taken for successful closure of the safety-related steam generator isolation valves and opening of the safety-related primary system safety valve, but the non-safety-related steam generator dump valves are assumed to fail which increases the quantity of water/stream ingress to the primary system.

As the primary coolant pressure continues to increase due to inleakage, the reserve shutdown control (RSC) equipment is tripped at 326 seconds when the pressure reaches the setpoint of 7.0 MPa. The RSC material is inserted into the core to maintain reactor subcriticality. At the same time, the main loop is also shut down on high pressure. The main circulator is tripped, and the steam generator is isolated. About 1860 kg of steam enters the primary system prior to steam generator isolation. Depending on the location of the leak, a large portion of the steam generator inventory can subsequently enter the primary system, with as much as 2200 kg flashing to steam.

The differential pressure that initially drives the inleakage decreases with time, so that the ingress rate of steam after isolation ramp from 5.7 kg/sec to zero over about 13 minutes. Once pressure equilibrates, water may continue to enter the primary coolant, but in the absence of a heat source it is assumed to remain in the steam generator vessel as liquid. Therefore, the water is unavailable to react with the core. Furthermore, due to the lack of forced circulation between the steam generator vessel and the reactor vessel, only the steam in the reactor and reactor plenums (28 percent of the total steam inleakage) is available to react with the core.

The primary coolant pressure increases until the relief valve opens at 370 seconds. The valve limits pressure to 7.18 MPa and reseats at 393 seconds when the pressure reaches 6.10 MPa.

After this, the relief valve cycles twice and is then assumed to fail open and depressurize the primary system, which maximizes the radiological dose consequences. Typically, a primary coolant pressure plot would be provided, but none is available for this example.

Following the loss of all forced circulation there is a slow heatup of the core, followed by a cooldown. Before the system depressurizes, the natural circulation within the core redistributes heat from the hottest portions of the core to the cooler regions, thus enhancing the conduction and radiation heat transfer from the core by distributing the heat over a larger surface area. The core temperature reaches a maximum of approximately 1540°C (2800°F) at 95 hr. This is below the 1600°C limit associated with the onset of fuel failure. Thereafter the heat removal rate exceeds heat generation and system temperatures begin to decrease as shown on Figure 3.6.N-2.

The thermal transient of the core through 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> is shown in Figure 3.6.N-3.

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Appendix B Example LBE Description Figure 3.6.N-2. Heat Generation and Heat Removal Figure 3.6.N-3. Maximum and Average Core Temperatures B-15

Appendix B Example LBE Description Offsite releases of radioactivity from the event are due to: 1) the release of circulating activity, 2) steam-induced vaporization and recirculation of plated-out activity, 3) release of a small fraction of the fuel activity due to the temperature transient, 4) release of fission products contained in the matrix and structural graphite that becomes oxidized, and 5) release due to the hydrolysis of failed fuel. The recirculation of plateout and the hydrolysis of failed fuel before the depressurization are the major contributors to iodine and noble gas release. The cumulative number of curies released from the vessel as a function of time is shown in Figure 3.6.N-4 for the important nuclides that contribute to dose. Dose to a receptor at the EAB is calculated mechanistically considering the phenomena of plateout and settling in the Reactor Building, radioactive decay, and atmospheric dispersion. The mean exclusion area boundary dose is conservatively calculated to be 0.045 rem total effective dose equivalent (TEDE) over a 30-day period. This value is well below the 10 CFR 50.34 EAB two hour dose limit of 25 rem TEDE.

The mechanistic source term is addressed in Section 2.J, and the dose analysis methodology is covered in Section 2.K.

The reactor has achieved its safe, stable end state by 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> when the analysis is terminated.

The fundamental safety functions of controlling heat generation, controlling heat removal, controlling chemical attack, and retaining radionuclides have been achieved.

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Appendix B Example LBE Description Figure 3.6.N-4. Curies of Selected Radionuclides Released from the Reactor Vessel B-17

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