ML24205A022

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Master Slide - Advanced Reactor Stakeholder Public Meeting - 2024-07-24
ML24205A022
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Issue date: 07/23/2024
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Advanced Reactor Stakeholder Public Meeting July 24, 2024 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 590 049 732#

Time Agenda Speaker 10:00am Opening Remarks Steven Lynch (NRR) 10:15am Insights from Public Workshop on Risk Metrics to Support Risk-Informed Programs for Advanced Reactors Gerardo Martinez-Guridi (RES),

Marty Stutzke (NRR) 10:30am Guidelines for the Seismic Isolation of Advanced Reactors: Topical Report Benjamin Carmichael (Southern Company),

Brandon Chisholm (Southern Company),

Jason Redd (Southern Company),

Andrew Whittaker (University at Buffalo) 11:30am Observations on the Southern Company Draft Base Isolation Guidelines Report Thomas Weaver (RES),

Weijun Wang (RES),

Jose Pires (RES) 12:00pm Lunch Break 01:00pm Key Terms in NEIMA: Performance-based, Technology-inclusive and Risk-informed Rani Franovich (Nuclear ROSE Consulting, LLC) 01:15pm DG-1290, Revision 1 - Proposed Revision 3 to Regulatory Guide 1.59 Jenise Thompson (NRR) 01:30pm Licensing and Deployment Considerations for Nth-of-a-Kind Micro-Reactors Duke Kennedy (NRR),

Jackie Harvey (NRR),

Peyton Doub (NMSS) 02:30pm Adjourn 2

Opening Remarks

Advanced Reactor Program Recent Highlights

Highlights

  • TerraPower, LLC -- Kemmerer Power Station Unit 1 Construction Permit Application
  • 07/15/2024 - NRC issued General Audit Plan to USO (TerraPower)

(https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML24187A117)

  • The SE was completed 4-months ahead of the 14-month review schedule.
  • The final Environmental Assessment (EA) is on track to be issued in August 2024.
  • Total resources are anticipated to be less than 60% of the original estimate for the review and issuance of the SE and EA.

The staff leveraged the recently completed Hermes 1 CP application review to identify efficiencies in the review.

5

Highlights (continued)

Alternative Physical Security Requirements for Advanced Reactors (RIN 3150-AK19)

(https://www.nrc.gov/docs/ML2417/ML24170A753.html) 6

Highlights (continued)

  • September 25, 2024 -2024 NRC Standards Forum
  • This will be a hybrid meeting:
  • In-person at the TWFN Auditorium
  • Online via MS Teams
  • The meeting notice and registration page are available:

Insights from Public Workshop on Risk Metrics and Reliability Data to Support Risk-Informed Programs for Advanced Reactors NRC Working Group on Technology Inclusive Risk Metrics Office of Nuclear Regulatory Research (RES)

  • Gerardo Martinez-Guridi
  • Matt Humberstone
  • Marty Stutzke
  • Hanh Phan Periodic Advanced Reactor Stakeholder Meeting July 24, 2024

NRC Considering Needs for Non-LWR Risk Metrics and Reliability Data In 1990, the Commission established three risk metrics for new reactors and associated quantitative goals:

Core Damage Frequency (CDF) < 1x10-4/year - A measure of overall safety performance in prevention of severe accidents Large Release Frequency (LRF) < 1x10-6/year - A measure of prevention of significant offsite consequences Conditional Containment Failure Probability (CCFP) < 0.1 - A measure of the capability of design to mitigate a severe accident Traditional risk metrics, e.g., CDF, have been used effectively in NRCs risk-informed decision-making processes May not be applicable to all advanced reactor designs SRM SECY-23-0021 provides direction on applicant proposed risk metrics The staff should revise draft 10 C.F.R. 53.220 to specify that applicants must propose a comprehensive plant risk metric (or set of metrics)

Need to consider alternative risk metrics that:

Are applicable to Non-light-water reactor (NLWR) designs Support NRC licensing and regulatory processes 9

Public Workshop on July 18, 2024

  • SRM-SECY-23-0021 provided motivation and direction to staff-related NLWR risk metrics, but this workshop was not part of the Part 53 rulemaking
  • Presented NRC staffs ongoing efforts on developing the following items related to risk metrics for NLWRs:
  • Interim staff guidance (ISG) on the review of applicant-proposed risk metrics
  • Potential NRC risk metrics
  • Reliability data
  • 166 online attendees and 19 in-person attendees 10

External Presenters (sorted by last name)

Name and Organization Presentation Title Cyril Draffin, U.S. Nuclear Industry Council USNIC Perspectives on Risk Dave Grabaskas, Argonne National Lab.

NLWR Data Insights and Experience Kyle Hope, Westinghouse Electric Company Challenges and Lessons Learned in Applying NEI 18-04 During Active Design: The eVinci' Microreactor Ed Lyman, Union of Concerned Scientists UCS Views on Advanced Reactor Risk Metrics Jessica Maddocks, X-Energy Hazard Level Selection for LMP Diego Mandelli, Idaho National Laboratory Intertwining of Data, Decisions, and Reliability Adam Stein, The Breakthrough Institute Breakthrough Institute Perspectives on Risk Metrics Eric Thornsbury, Electric Power Res. Inst.

EPRIs Risk Metric Work Patrick White, Nuclear Innovation Alliance NIA Perspectives on Comprehensive Risk Metrics Sai Zhang, Idaho National Laboratory Advanced Reactor Operating Experience Data Analysis to Support Risk Estimation - Challenges, Current Practice, and Needs 11

Development of ISG to Review Applicant-Proposed Risk Metrics

  • The staff presented its tentative approach to developing interim staff guidance (ISG) for review of applicant-proposed risk metrics:
  • ISG applicability
  • Terminology related to risk metrics
  • Flowchart for conducting the review
  • Change provisions
  • Intellectual property 12

Summary of Initial Approach to Risk Metrics for NLWRs

  • For accident prevention:
  • Use CDF whenever core damage is applicable
  • Use new metrics when core damage is not applicable (e.g., frequency of failure of initial confinement of radioactive material)
  • Consequence metrics are technology inclusive, but there are challenges associated with them
  • Desirable attributes for risk metrics and for using the metrics were proposed 13

Planned Data Activities Workshop (July 18, 2024)

Examine available existing databases of operating experience of advanced reactors Establish database templates, reporting criteria, and data methods/procedures to support risk modeling and regulatory oversight Populate the new database with operational data from prominent advanced reactor designs 14

Next Steps

  • Received many constructive and thoughtful suggestions during the workshop
  • The staff looks forward to continued stakeholder engagement
  • Develop ISG on the review of applicant-proposed risk metrics
  • Tentative plan to issue the ISG when proposed Part 53 is finalized
  • Develop NRC white paper on potential risk metrics
  • Tentative middle of calendar year 2025
  • Establish a task on reliability data
  • By the end of calendar year 2024
  • Contact if interested in providing additional feedback

Acronyms CCFP Conditional Containment Failure Probability CDF Core Damage Frequency CFR Code of Federal Regulations ISG Interim Staff Guidance EPRI Electric Power Research Institute LERF Large Early Release Frequency LMP Licensing Modernization Project LRF Large Release Frequency NEI Nuclear Energy Institute NIA Nuclear Innovation Alliance NRC Nuclear Regulatory Commission NRR (NRC) Office of Nuclear Reactor Regulation RES (NRC) Office of Nuclear Regulatory Research UCS Union of Concerned Scientists USNIC U.S. Nuclear Industry Council 16

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 1

7 ML24134A173 Guidelines for the seismic isolation of advanced reactors: topical report Ben Carmichael, Southern Company Brandon Chisholm, Southern Company Jason Redd, Southern Company Andrew Whittaker, University at Buffalo

1 8

Project background

  • Effort selected for cost-shared funding in 2020 by DOE through Funding Opportunity DE-FOA-0001817, US Industry Opportunities for Advanced Nuclear Technology Development
  • Builds upon decades of non-nuclear seismic isolation experience and efforts to develop standards for use in nuclear industry
  • Modeled an archetype advanced reactor building, equipment, and site on which to demonstrate a pathway for performance-based design of a seismic isolation system
  • Requesting NRC endorsement of the pathway via a Topical Report
  • Regular engagement with NRC and industry to validate direction and incorporate feedback to promote/maximize usefulness of pathway NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 1

9 Some terminology

Seismic isolation

  • 2D horizontal or 3D
  • Nearly all 10,000+ applications 2D horizontal NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 2

0

Seismic isolation Arup, 2020

  • Isolated LLWRs: Cruas and Koeberg
  • Synthetic rubber bearings
  • 2D horizontal isolation NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 2

1

Technical basis, USNRC-and DOE-funded USNRC (2008-2017): Seismic isolation of large light water reactors DOE (2014-2016): Seismic isolation of components in advanced nuclear reactors DOE (2016-2018): Evaluation of the potential effect of seismic risk at DOE facilities DOE (2017-2019): Seismic isolation of advanced reactors with considerations of fluid structure interaction DOE (2018-2020): Seismic isolation of major advanced reactor systems for economic improvement and safety assurance EPRI (2018-2019): Cost basis for utilizing seismic isolation for nuclear power plant design ARPA-E (2018-2021): Reducing the overnight capital cost of advanced reactors using equipment-based seismic protective systems DOE via Southern Company (2021-2024): Topical report on seismic isolation of advanced reactors DOE ARDP via MIT (2021-2024): Horizontally configured high-temperature gas reactor DOE NEUP (2022-2025): Gamma irradiation effects on the mechanical properties of seismic protective devices NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 2

2

Technology readiness: seismic isolation NUREG/CR-7253 Technical Considerations for Seismic Isolation of Nuclear Facilities Office of Nuclear Regulatory Research Seismic Isolation of Nuclear Power Plants Using Sliding Bearings Office of Nuclear Regulatory Research NUREG/CR-7254 NUREG/CR-7255 Seismic Isolation of Nuclear Power Plants using Elastomeric Bearings Office of Nuclear Regulatory Research Guidelines for Implementing Seismic Base Isolation in Advanced Nuclear Reactors Final Report Issued for Submission Document Number SC-SND8932-001 Rev 0 Developed Pursuant to a Federal Award from the Department of Energy Funding Opportunity DE-FOA-0001817 Granted to Southern Nuclear Development, LLC Award No. DE-NE0008932 May 2024 Prepared for Submission to the U.S. Nuclear Regulatory Commission (NRC)

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 2

3

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 2

4 Seismic isolation topical report Develops a reactor-agnostic pathway for applicants to develop, document and qualify a seismic isolation system for an advanced reactor by including Seismic isolation systems: technology, use, guidelines Earthquake shaking definitions, performance expectations Archetype reactor building, equipment, siting Risk-based design of a seismic isolation system Qualification, prototype, and production testing Specifications for supply of isolators and dampers Commercial grade dedication Generating a displacement demand curve Achieving a risk target, including derivation of fragility functions Selecting a target performance goal: how to start?

Isolation-system options: judging different systems Guidelines for Implementing Seismic Base Isolation in Advanced Nuclear Reactors Final Report Issued for Submission Document Number SC-SND8932-001 Rev 0 Developed Pursuant to a Federal Award from the Department of Energy Funding Opportunity DE-FOA-0001817 Granted to Southern Nuclear Development, LLC Award No. DE-NE0008932 May 2024 Prepared for Submission to the U.S. Nuclear Regulatory Commission (NRC)

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 25 Seismic isolation topical report

  • Front matter Cradle to grave guidance Passive isolation systems only No semi-active or active systems, why?

Which one: risk-based, risk-informed, or performance based?

Applicability Microreactors, advanced reactors/SMRs, large light water reactors Seismic isolation checklist Role of isolation in defense-in-depth Additional echelon of defense via reduced seismic demands Superior quality, passive, predictable performance (how?)

Does not change the need to confirm adequate DiD Guidelines for Implementing Seismic Base Isolation in Advanced Nuclear Reactors Final Report Issued for Submission Document Number SC-SND8932-001 Rev 0 Developed Pursuant to a Federal Award from the Department of Energy Funding Opportunity DE-FOA-0001817 Granted to Southern Nuclear Development, LLC Award No. DE-NE0008932 May 2024 Prepared for Submission to the U.S. Nuclear Regulatory Commission (NRC)

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 Seismic isolation checklist Establish a target performance goal (TPG) for the seismic isolation system Generate seismic hazard curves, uniform hazard response spectra, and ground-motion time series Establish an isolation-system-specific seismic displacement demand curve Determine the median displacement capacity of the isolation system required to achieve the TPG Establish DB earthquake shaking for the isolation system Compute displacements, velocities, and forces for prototype and production testing Compute earthquake-induced demands on the superstructure and substructure Prepare specifications for supply of isolators and dampers Prepare specifications for prototype and production testing of isolators and dampers Prepare a plan for commercial grade dedication of isolators and dampers Prepare requirements for maintenance, operation, and testing of isolators and dampers

SSCs above the isolation interface treated per industry practice 26

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 27 Requested NRC review and possible endorsement via a SER Role of seismic isolation solutions in ensuring adequate defense-in-depth Earthquake shaking definitions for isolation-system design Performance expectations for the seismic isolation system General requirements for isolation-system design Requirements for seismic inputs to support performance calculations Methods of dynamic analysis, and numerical models for isolators and dampers Requirements for design of connections for isolators and dampers; isolation-system substructure; structures,

systems, and components above the seismic isolation interface; systems and components crossing the seismic isolation interface; and clearance around the isolated building Performance-based design of a seismic isolation system Minimum standards for qualification, prototype, and production testing of isolators and dampers Specifications for supply of isolators and dampers Plan for commercial grade dedication of isolators and dampers Process to establish a log standard deviation for the fragility function of a seismic isolation system

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 28 Performance expectations for isolators and dampers DB shaking

@TPG Use Production testing of isolators and dampers Prototype testing of isolators and dampers Isolator and damper displacement Mean maximum from fragility analysis Damper velocity Mean maximum Mean maximum from TPG shaking Acceptance criteria Production testing of each isolator for mean maximum horizontal displacement and corresponding axial force Prototype testing of three isolators of each type and size for displacement and TPG axial force Production testing of each damper for mean maximum displacement and corresponding maximum velocity Prototype testing of three dampers of each type and size for displacement and TPG velocity No damage to isolators Isolator damage is acceptable but load-carrying capacity for gravity loading is maintained No damage to dampers No damage to dampers 50 D

50 D

50 D

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 29 Archetype reactor building, equipment, siting, SDC4

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 30 Performance-based design of a seismic isolation system CD 0

1 2

3 4

5 6

Spectral acceleration (g) 10-6 10-5 10-4 10-3 10-2 10-1 Mean annual frequency of exceedance PGA 0.1 s 2 s BC BC CD

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 31 Performance-based design of a seismic isolation system F

u Kiso,h F

v v

Fd =sign(v)*Cd*lvl F

u NUREG/CR-7255 Seismic Isolation of Nuclear Power Plants using Elastomeric Bearings Office of Nuclear Regulatory Research Seismic Isolation of Nuclear Power Plants Using Sliding Bearings Office of Nuclear Regulatory Research NUREG/CR-7254 Makris and Constantinou, 1992 Parsi et al., 2024

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 32 Generating a displacement demand curve BC F

u Kiso,h F

u

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 33 Deriving an isolation system fragility function

  • Isolation-system fragility function Defined by a median capacity (D50) and a log standard deviation, beta Median capacity underestimated by testing, why?

Beta addresses uncertainties in displacement demand and capacity Parameters per EPRI (2018)

Ground motion, modeling, damping, mode contribution, input time series, foundation-structure interaction, strength, inelastic energy absorption Monte Carlo analysis of Huang et al. (2009) for linear and bilinear isolation systems for beta for ground motion and modeling Beta for model fidelity uncertainty per EPRI (2018) = 0.05 Composite beta = 0.13 for linear systems, = 0.22 for nonlinear systems

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 34 Performance-based design of a seismic isolation system

  • Isolation-system fragility function Pre-determined log standard deviation per Appendix D (0.15 to 0.25)

Increment D50 to achieve user-specified target performance goal (TPG)

TPG

35 Performance-based design of a seismic isolation system

  • Use of displacement F50
  • Prototype testing of isolators and dampers
  • Clearance to adjacent construction
  • Forces at F50 along each horizontal axis +

TPG shaking demands Design and detailing of substructure (pedestals, foundation)

Design of connections to substructure and superstructure Analysis and design of SSCs Isolated structure Adjacent construction D50 D50

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 36 Qualification, prototype, and production testing Requirements based on 35 years of US industry practice ASCE 7, ASCE 41, AASHTO, applications to mission-critical structures Nuclear applications no less onerous requirements than non-nuclear Qualification testing Project independent, to be executed before being considered for a nuclear energy application Dynamic testing of devices and shake-table testing of systems of devices Verified and validated numerical models Demonstrate that change in mechanical properties less than (+20%, -20%) over design life Obliges prior experience and testing of technology, accelerated aging tests not permitted Prior deployment in non-nuclear, mission-critical applications in the US, moderate to high seismic hazard Prototype testing Project specific, requirements for all isolators and dampers addressed in the topical, full size Test three of each type and size of isolator and damper (rationalize number of types)

Demonstrate that isolators and dampers can sustain demands consistent with the TPG Dynamic per loading environment Production testing Project specific, requirements for all isolators and dampers addressed in the topical Test 100% of isolators and dampers prior to shipment to construction site Demonstrate that isolators and dampers can sustain demands consistent with DBE shaking Dynamic or slow speed, depending on device; dampers tested dynamically DB shaking

@TPG Use Production testing of isolators and dampers Prototype testing of isolators and dampers Isolator and damper displacement Mean maximum from fragility analysis Damper velocity Mean maximum Mean maximum from TPG shaking Acceptance criteria Production testing of each isolator for mean maximum horizontal displacement and corresponding axial force Prototype testing of three isolators of each type and size for displacement and TPG axial force Production testing of each damper for mean maximum displacement and corresponding maximum velocity Prototype testing of three dampers of each type and size for displacement and TPG velocity No damage to isolators Isolator damage is acceptable but load-carrying capacity for gravity loading is maintained No damage to dampers No damage to dampers 50 D

50 D

50 D

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 37 Specifications

  • Provides a pathway for both applicants and regulator
  • Based on 35 years of US industry practice Low-damping natural rubber bearings Lead-rubber bearings, natural rubber as the base elastomer Spherical sliding bearings (or Friction PendulumTM)

Single concave and triple concave 1D fluid viscous dampers

  • Generated by best-in-class consultants Deep experience in seismic isolators and dampers Analysis, design, testing, procurement, inspection
  • Reviewed by the US suppliers of isolators and dampers

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 38 Commercial grade dedication Why? The alternative to NQA-1 CGD undertaken to provide reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety function and, in this respect, is deemed equivalent to an item designed and manufactured under a 10 CFR Part 50, Appendix B Assurance is achieved by identifying the critical characteristics of the item and verifying their acceptability by inspections, tests, or analyses by the purchaser or third-party dedicating entity.

Draws on EPRI (2014) Plant engineering: guideline for the acceptance of commercial-grade items In nuclear safety-related applications Chapter 8 plan for CGD Step 6: select acceptance criteria methods and document acceptance criteria Step 7: conduct acceptance activities 100% testing of all production isolators and dampers for DBE demands No other safety-related items in a nuclear power plant are tested so rigorously

NRC Advanced Reactor Stakeholder Public Meeting, July 24, 2024 39 Acknowledgments US Department of Energy, Award No. DE-NE0008932 University at Buffalo: Sharath Parsi, Kaivalya Lal, Ching-Ching Yu, Faizan Ul Haq Mir, Manish Kumar^2, Yin-Nan Huang Idaho National Laboratory: Chandrakanth Bolisetti Kairos Power: Matt Denman, Brian Song SGH: Mohamed Talaat KPFF: Reid Zimmerman Advisory committee: Jacopo Buongiorno and Koroush Shirvan (MIT), John Richards and Hasan Charkas (EPRI), Karl Fleming (KNFC), Troy Morgan (Exponent)

NEI Advanced Reactor Regulatory Task Force: Bob Iotti (ARC Clean Energy),

Mory Diane (Oklo), Jim Kinsey (INL), Amanda Spalding (WEC), Dennis Henneke (GEH), and Ian Gifford (TerraPower)

Observations on the Southern Company Draft Base Isolation Guidelines Report Thomas Weaver, Weijun Wang, and Jose Pires Office of Nuclear Regulatory Research, Nuclear Regulatory Commission July 24, 2024 Advanced Reactors Stakeholders Meeting

Overview

  • Draft Report Purpose and Scope:
  • Document and provide the technical justification for a process to select, analyze, design, and deploy a passive isolation system beneath an advanced reactor building that meets applicable regulatory requirements
  • The guidelines are limited to the isolation system design and do not address design of SSCs supported by base isolators and viscous dampers
  • To be submitted to NRC as topical report for formal review and issuance of a safety evaluation report.
  • Staff observations provided for consideration in developing a final draft.

Draft Report:

Guidelines for Implementing Seismic Base Isolation in Advanced Nuclear Reactors, Southern Company, Document Number SC-SND8932-001 Rev A, May 2024 41

Endorsement Request (Section 1.2)

  • Section 1.5
  • The role of seismic isolation solutions in defense-in-depth
  • Section 3.2 - 3.7
  • Earthquake shaking definitions, performance expectations, and other requirements
  • Section 5
  • Performance-based design of a seismic isolation system
  • Section 6
  • Qualification, prototype, and production testing
  • Section 7
  • Specifications for the supply of isolators and viscous damping devices
  • Section 8
  • Appendix D
  • Achieving a performance target for a seismic isolator 42

Endorsement Request (Section 1.2) and Seismic Isolation Checklist (Section 1.3)

  • Item 7 in Section 1.2 includes requirements for design of structures, systems and components above the seismic isolation interface
  • Confirm that the requirement for the SSCs above the seismic isolation interface is only the one pertaining to their seismic design category (SDC) vis-

-vis the SDC for the isolation system (Section 3.7.4)

  • Section 1.3 states that the seismic isolation checklist does not apply to structures, systems and components above the isolation plane
  • The last paragraph of Section 1.3 provides a guideline for the analysis and design of the SSCs above the isolation plane including referring to standards for the assignment of seismic design categories
  • Confirm that if this is outside the scope of the report guidelines for which endorsement is requested 43

Role of Isolator Solutions in Defense-in-Depth (Section 1.5)

  • Key Attribute of Defense in Depth
  • Multiple independent and redundant layers of defense so that no single layer, no matter how robust, is exclusively relied upon
  • Draft Guidance: seismic isolation supports the objective of defense-in-depth
  • Seismic isolation augments the approach to achieve defense-in-depth
  • The rules for implementing seismic isolation systems satisfy criteria for implementation of defense in depth
  • Needs clarification of what specific endorsement is requested
  • E.g., if the endorsement is for a process to assess the isolation system contribution, what are the specific steps for that assessment?

44

  • Additional technical information and discussion on the following is recommended:
  • Is the use of a scale factor SF = 0.5 to calculate the DB spectral from the TPG spectra a simplification for the examples shown or a guideline? (Section 3.2)
  • Details on how to evaluate the isolators and viscous dampers so that they do not suffer damage under wind loads (Section 3.4.1.4 and 3.4.1.1) including wind-borne missiles (Section 3.4.1.1)
  • Power spectral density (PSD) requirements for spectrally matched input motions (Sections 3.5, 3.7 and 5.2.2 for example)
  • Can the recommended use of 11 sets of ground motion be compared with the use of 30 sets of ground motions used in the example of Section 5? (Section 3.5)
  • Discussion to support comments in the report on the significance of soil-structure-interaction for base-isolated advanced reactor buildings and equipment (e.g., Section 3.6.1; Appendix D 2.2) (e.g., contrasting with ASCE 4-16 provisions in Section 12.3)

Earthquake Shaking Definitions, Performance Expectations, and Other Requirements (Section 3) 45

Earthquake Shaking Definitions, Performance Expectations, and Other Requirements (Section 3)

  • Additional technical information and discussion on the following is recommended:
  • Is D50 established from a fragility analysis alone or fragility analysis in combination with the seismic hazard and a target MAFE? (Section 3.7.2 and Section 3.3)
  • What is the intent of the sentence in Section 3.7.2 stating a dynamic analysis of a 3D finite element model of the isolated building shall be performed for DB and TPG shaking?
  • For the calculation of D50 only the lumped building reactive mass above the basemat is used in Section 5.2.2
  • Is a verification of D50 expected after the building design is more detailed?
  • Is it for the calculation of axial forces on the isolators?
  • Why is ACI 318-19 proposed for the design of the isolators pedestals? (Section 3.7.3)
  • Justification for use of precast concrete pedestals (Section 3.7.5) 46

Performance Expectations (Section 3.3)

Design Basis Shaking At Target Performance Goal Use Production testing of isolators and dampers Prototype testing of isolators and dampers Isolator and Damper Displacement Mean maximum D50 from fragility analysis Damper Velocity Mean maximum Mean maximum from TPG shaking Acceptance Criteria Production testing of each isolator for mean maximum horizontal displacements and corresponding axial force Prototype testing of three isolators of each type and size for D50 displacement and TPG axial force Production testing of each damper for mean maximum displacement and corresponding maximum velocity Prototype testing of three dampers of each type and size for D50 displacement and TPG velocity No damage to isolators Isolator damage is acceptable but load-carrying capacity for gravity loading is maintained No damage to dampers No damage to dampers From Table 3.1 in the Draft Report 47

Performance Expectations (Section 3.3)

  • Consider adding discussion on how and why performance expectations vary from ASCE 4 Table 12-1.
  • Expand on meaning and clarification of terms in table,
  • Mean maximum displacements (DB) and their use
  • TPG axial force (TPG)
  • TPG velocity (TPG)
  • E.g., terms denoting response demands, viscous dampers demands and terms referring to isolators' capacities 48

Performance Based Design (Section 5) 49

Performance-Based Design of a Seismic Isolation System (Section 5, Appendix D)

  • Separation between guidelines and example calculations
  • Listing of guidelines and then their illustration with the examples
  • For example, can the guideline for establishing D50 be written as the median fragility displacement capacity corresponding to a MAFE equal to the TPG MAFE?
  • Define the building reactive mass at the level of the basemat (Section 5.2.2)
  • Comparison of the proposed fragility analysis Section 5.2.3) and displacement hazard approach (Section 5.2.4) with the traditional approach where the original hazard curve is convolved with a fragility curve which is a function of spectral acceleration in the hazard curve.
  • Is the approach in Section 5 one way to determine D50 for the target MAFE and other approaches are acceptable provided prototype testing per Section 6 confirms D50?
  • E.g., traditional fragility analysis approach and convolution of fragility analysis with the original hazard curve
  • How would the logarithmic standard deviations for fragility analysis change with different approaches such as the traditional fragility analysis methodology?
  • How would the median fragility change?

50

Qualification, Prototype, and Production Test (Section 6)

  • Qualification Tests: Demonstrate basic performance characteristics independent of project application for qualification of vendors/isolators/dampers
  • Prototype Tests: Performed on limited number of devices to confirm the isolators D50 capacity and the dampers requirements
  • Production Tests: Performed on all devices at the DB seismic loads 51

Qualification, Prototype, and Production Test (Section 6)

  • Approach for qualification tests is comprehensive with detailed information and uses parameters in the performance expectation table (Table 3.1)
  • Identify and justify how the testing protocols were derived from ASCE 4-16 (Chapter 12),

ASCE 43-19 (Chapter 9) and ASCE 7-22 (Chapters 17 an 18)

  • What was used or not from each code and what has been added and why
  • Justification and sources for the acceptance criteria, especially the deviations from the target values (e.g., force-displacement relations, stiffness)
  • Testing protocols are per isolator and damping types to address different behaviors
  • Feasibility of expressing protocols in terms of behavior and response characteristics of isolators and dampers rather than type
  • Qualification testing: can exemptions for qualification testing be related to specification and performance requirements rather than through a list of specific vendors?
  • Is there an intent to also validate the isolators and dampers mathematical models using the testing results (Section 3.6.3 models)?

52

Specifications for the Supply of Isolators and Viscous Damping Devices (Section 7)

  • Draft specifications for supply of 2D isolators and 1D fluid viscous dampers.
  • 3D isolation systems are not addressed
  • Gamma and Neutron radiation resistance are not addressed
  • Clarify that the Section 7 specification are not related for technical plant specifications in Appendix E
  • Part II.4 Design Calculations
  • The only seismic demands considered are D50
  • Axial loads at TPG shaking
  • Is this the shaking characterized by the uniform hazard spectra at the TPG MAFE?
  • Are any design calculations at the DB shaking needed?
  • Why is the AISC 360 chosen for the mounting plates as opposed to the AISC N690?
  • Level of detail recommended for the SSCs above the isolation interface to calculate the demands in the load combinations?

53

Commercial Grade Dedication of Seismic Isolators and Dampers (Section 8)

  • Section 8 provides a template for the technical evaluation tasks required by 10CFR21 and 10CFR50 Appendix B and follows the process in Section 4 of EPRI TR-102260 (Plant engineering: guideline for the acceptance of commercial-grade items in nuclear safety-related applications)
  • Recognizes that the entirety of the supplier CGD is out of scope for Section 8
  • Address if CGD of isolators is anticipated for most designs
  • Section 8 would be reviewed by staff specialized in CGD regulations, guidance and processes 54

Other Items

  • Appendix E - Considerations for ITAAC
  • Endorsement of Appendix E is not requested and observations on Appendix E are outside of the scope of these observations including the assertion in Section E.3 that a seismic isolation system and associated components need not be explicitly addressed in a plants Technical Specifications 55

Summary

  • Guidance document is limited to isolation system design and does not address design of structures, systems, and components supported by base isolators and viscous dampers
  • Additional technical justification and discussion is recommended for some sections of the report to facilitate a safety evaluation review 56

Acronyms ACI: American Concrete Institute AISC: American Institute of Steel Construction ASCE: American Society of Civil Engineers CFR: Code of Federal Regulations CGD: Commercial Grade Dedication DB: Design Basis D50: Median displacement from fragility curve ITAAC: Inspections, tests, analyses, and acceptance criteria MAFE: Mean annual frequency of exceedance PSD: Power spectral density RG: Regulatory Guide TPG: Target performance goal, frequency of unacceptable performance SSC: Structures, Systems, or Components SDC: Seismic design category SF: Scale factor applied to the TPG hazard curve to develop the design response spectrum 57

LUNCH BREAK Meeting will resume at 1:00 pm EDT Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 590 049 732#

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Performance-based Regulation:

Where We Are and Where we Need to Be Rani Franovich Principal Consultant and Expert Witness Nuclear ROSE Consulting, LLC La Reine 1998 Alexander Reznichenko, Ukrainian Private Collection

Where We Have Been Commission Direction re Modernization (1999)

Risk-informed (RI) and Performance-based (PB) approaches Terms defined by the Commission in SRM-SECY-98-144 RIPB approach to oversight (Reactor Oversight Process, ROP)

Nuclear Energy Innovation and Modernization Act (2019) 10 CFR Part 53 Focus on RI Reviews using Licensing Modernization Project (LMP 2021)

Conflation of terms during NRC public meeting on Framework B (June 16, 2022)

NRC public meeting to align on terminology (June 30, 2022)

NRC public meeting to discuss NEIMA concepts and confirm alignment on terminology (July 28, 2022) 60

More Recent Developments 10 CFR Part 53 Submitted to Commission (March 2023)

Commission Directed Staff to Revise Rule (March 2024)

Continued Lack of Coherence Around PB Methods NRCs Regulatory Information Conference (RIC), March 2024 Question about whether NRC really is performance-based and risk-informed Responses focused almost exclusively on RI concepts, not PB methods Advanced Reactor Stakeholder Meetings (March 27 and May 23, 2024)

More focus on RI and LMP Stakeholder requests for further discussion of PB approaches Advanced Reactor Construction Oversight Process Workshops (Feb-Jul, 2024)

Focus on enforcement of compliance, not performance Not pursuing performance indicators because data do not exist Passage of the Accelerating Deployment of Versatile Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE Act, July 9, 2024) 61

Where We Are

  • July 18, 2024 - NRC Risk Metrics Workshop Stakeholders (particularly small reactor developers) struggle with implementing LMP DOE Lab contrasted margin-based and probability-based assessment of structure, system and component (SSC) performance Probability-based assessment of SSC monitoring system and decision process based on plant CDF and LERF Margin-based assessment of SSC health based on monitoring performance data
  • Margin-based Performance Monitoring Represents a PB Approach Provides for assessment of SSC health during reactor demonstration Facilitates seamless transition to continuous performance monitoring of data during commercial operation Affords efficient, effective regulatory oversight of RI safety parameters using performance indicators 62

Where We Need to Be More Attention to PB Solutions Observable parameters (quantitative, qualitative, or combination) directly related to outcomes are developed Margin-based approaches to SSC performance (use demonstration data for FOAK and operational data for oversight and performance monitoring)

Objective acceptance criteria for each parameter is developed.

Margin-based approaches to SSC performance (compare actual performance to objective performance criteria)

A decision-making framework is developed for evaluating and assessing physical and temporal margins.

Structured hierarchy with high-level performance objectives Flexibility is afforded to the licensee to show that the margins are being employed to improve outcomes.

Prescription is minimized in regulations, guidance and industry standards Applicants propose means for demonstrating how high-level performance objectives are met 63

Advantages of PB Approaches Achieve high levels of safety performance without undue regulatory burden Key to satisfying NEIMA mandate for technology-inclusive frameworks Offer flexibility for innovation High-level performance objectives are technology-neutral Incorporate risk insights

  • Data-driven vice theoretical Utilize actual reactor demonstration and operational data Offer seamless transition to PB oversight during plant operation Save time and costs to license first-of-a-kind reactors (FOAK)

Support rapid deployment of nth-of-a-kind reactors (NOAK)

Can leverage advances in digital technology for continuous monitoring Facilitate international harmonization 64

Summary, Conclusion & Next Steps Summary:

Most stakeholders agree that regulatory modernization is necessary to keep pace with innovation, incentivize improved safety outcomes, and mitigate enterprise risk to the industry.

==

Conclusion:==

It is not clear how NRC and stakeholders define modernization, particularly as it pertains to PB approaches.

Next Steps:

A workshop is needed to establish a shared understanding of what PB means and how it will be applied to satisfy 1999 Commission direction, NEIMA, and the ADVANCE Act.

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DG-1290, Revision 1 - Proposed Revision 3 to Regulatory Guide 1.59 Appendix K - Consideration for Applying Guidance to Advanced Reactors and Small Modular Reactors Presentation to Advanced Reactor Stakeholders Meeting July 24, 2024 Jenise Thompson NRR - Division of Engineering and External Hazards 66

Overview

  • RG 1.59, Revision 2, issued in 1977
  • Revision 3 initially issued for public comment as DG-1290 in 2022, reissued for public comments July 15, 2024
  • Appendix K adapts methodology from RG 4.26 to use a flexible stepwise approach to leverage PRA and site characteristics in flood evaluations for advanced reactors and SMRs 67

Adapting from RG 4.26

  • Leverage existing site characterization information
  • Screen hazards and consider risk insights
  • Allow for evaluation of SSC performance and mitigating actions in addition to or in place of detailed hazard analysis RG 4.26, Figure 3 68

RG 4.26, Figure 3 DG-1290, Figure K-1 69

Appendix K Flowchart

  • Step 1 - leverage site characterization information
  • Step 2 - determine which, if any, flood-causing mechanisms affect plant performance
  • Step 3 - determine if there is adequate engineering for SSCs to withstand the hazard Figure K-1, DG-1290, Rev. 1 70

Appendix K Flowchart

  • Step 4 - evaluate mitigating actions for adequacy
  • Step 6 - reassess design features and/or consider PRA
  • Step 5 - assessment is complete and results are documented Figure K-1, DG-1290, Rev. 1 71

Appendix K Summary

  • Focus on flood causing mechanisms of importance to the design
  • End the flood evaluation at the earliest possible point in the process.
  • Consider PRA or comparable analysis
  • Iterate between evaluation of SSCs performance and mitigating actions and design reassessment to achieve satisfactory result.

Figure K-1, DG-1290, Rev. 1 72

Next Steps

  • DG-1290, Revision 1, public comment period open.
  • Disposition of first set of public comments - ML23320A026

Licensing and Deployment Considerations for Nth-of-a-Kind Micro-Reactors Advanced Reactor Stakeholders Meeting July 24, 2024 William Kennedy, Advanced Reactor Policy Branch Jackie Harvey, Advanced Reactor Policy Branch Peyton Doub, Environmental Tech Review Branch 1 U.S. Nuclear Regulatory Commission https://www.nrc.gov/reactors/new-reactors/advanced.html

Contents

  • Goals of this presentation
  • Background
  • Regulatory approaches for standardized operational programs
  • Alternative approaches for environmental reviews
  • Other licensing and deployment topics
  • Next steps 75

Goals of this Presentation

  • Inform stakeholders about regulatory approaches the NRC staff is developing, for Commission consideration, regarding standardized operational programs and alternative environmental reviews
  • Inform stakeholders about other licensing and deployment topics, potential near-term strategies, and next steps the NRC staff is considering.
  • Engender a dialogue with stakeholders 76

=

Background===

  • For licensing purposes, micro-reactors are commercial power reactors licensed under Section 103 of the AEA.
  • Micro-reactors typically use non-light-water reactor technologies, are expected to have power levels on the order of tens of megawatts thermal, small site footprints, low potential consequences in terms of radiological releases, and may have increased reliance on passive systems and inherent characteristics to control power and heat removal.
  • Factory-fabricated micro-reactors are a subset of micro-reactors that would rely heavily on standardization and mass production to simplify licensing and deployment.*

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  • See SECY-24-0008, Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, dated January 24, 2024 (ML23207A252).

=

Background===

  • For the purposes of this presentation, the term Nth-of-a-Kind (NOAK) micro-reactor generally means a micro-reactor of a standard common design that has been previously approved by the NRC through, a design certification, manufacturing license, or final safety analysis report for a first-of-a-kind (FOAK) combined license or operating license.
  • Nth-of-a-Kind micro-reactor licensing refers to licensing micro-reactors of a standard common design for operation as power reactors at fixed sites.

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Conceptual Deployment Model for Factory-Fabricated Transportable Micro-Reactors 79

Up-Front Approval of the Standard Plant

  • Standard design approved in a manufacturing license, design certification, construction permit and operating license, or combined operating license
  • Technical issues resolved
  • Standardized operational programs
  • Generic environmental review (to the extent practicable)
  • Hearings covering the standard design and environmental review Nth-of-a-Kind Licensing
  • Streamlined administrative processes
  • Confirmation of site suitability for the standard design
  • Site-specific environmental review (applying the generic environmental review, as appropriate)
  • Closure of ITAAC/license conditions
  • Confirmatory site-specific inspection
  • Site-specific hearing
  • Operating decision An approach by which a robust up-front approval of a standard design enables efficient, predictable licensing of Nth-of-a-Kind reactors 80

Regulatory Approaches for Standardizing Operational Programs

  • The NRC staff is exploring approaches to review operational matters at the design approval stage (ML or DC) for a standard micro-reactor design considering two general groups of operational programs:

Group 1: Design-related (e.g., technical specifications (TS), design quality assurance, and portions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV) in-service inspection and in-service testing programs)

Group 2: Site specific/operational (e.g., operator training, security programs, emergency preparedness, etc.) developed in the operating license or combined license under current processes 81

Regulatory Approaches for Environmental Reviews The current regulatory process is to perform an environmental impact statement (EIS) for nuclear power reactors*

The NRC staff is considering alternative approaches for environmental reviews for micro-reactors of a common design Any EIS, generic environmental impact statement (GEIS), or environmental assessment (EA) under any alternative may be tiered from the New Reactor GEIS 82 Alternative Environmental Review for First-of-a-Kind Environmental Review for Nth-of-a-Kind Design-Specific GEIS GEIS or Generic EA w/ Exemptions Supplemental EA tiered from FOAK GEIS or EA Environmental reviews associated with a design approved in a ML or DC EIS or EA with Exemptions Supplemental EA tiered from FOAK EIS or EA Micro-Reactor Online Environmental Review Portal EIS or EA with Exemptions Develop Design-Specific Portal Applicant supplies site-specific data into Portal NRC develops EA tiered from First-of-a-kind EIS/EA or from New Reactor GEIS based on applicant submission on Portal Design-Specific Categorical Exclusions (CATEX)

EIS or EA with Exemptions Develop CATEX and Checklist Applicant supplies site-specific data using Checklist NRC determines if CATEX applies

  • See SECY-24-0046, Implementation of the Fiscal Responsibility Act of 2023 National Environmental Policy Act Amendments (https://www.nrc.gov/docs/ML2407/ML24078A013.html), for discussion of the current process for environmental reviews and potential changes related to the Fiscal Responsibility Act of 2023.

Other Licensing and Deployment Topics Maximal design standardization

  • The regulations in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, provide several regulatory pathways for design standardization, including manufacturing licenses, design certifications, and standard design approvals, under which most safety issues would be resolved.
  • Maximal standardization would involve approval of a standardized micro-reactor design and subsequent deployment under a combined license or construction permit and operating license without any significant deviations or departures from the standardized design.
  • Maximal design standardization could allow micro-reactors of a common design to be deployed to most sites in the U.S. with minimal need for site-specific features or the associated additional NRC reviews and approvals.

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Other Licensing and Deployment Topics Grading the level of site characterization

  • A standardized design for a micro-reactor could establish bounding parameters for site characteristics that are important to the safety review so that micro-reactors of the standard design could be deployed at suitable sites throughout the U.S.
  • The NRC staff is considering approaches for grading the level of site characterization for NOAK micro-reactors of a standard design based on the applicable hazards for the specific micro-reactor design, the amount of margin included in the design for each bounding site parameter, and the amount of margin to appropriate dose reference values.
  • A graded approach could focus on how a construction permit or combined license applicant can provide the required site characterization information and demonstrate that the bounding parameters are met for the candidate site.

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Other Licensing and Deployment Topics Deployment site security

  • The existing requirements for security apply to licensing micro-reactors of a common design, including various regulations in 10 CFR Part 73, Physical Protection of Plants and Materials.
  • The NRC staff is considering additional approaches for streamlining the review of security for licensing Nth-of-a-Kind micro-reactors, including the possibility to standardize operational aspects of security, to the extent practical.

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Other Licensing and Deployment Topics Deployment site emergency preparedness

  • The existing regulations for emergency preparedness in 10 CFR Part 50, Domestic licensing of production and utilization facilities, apply to licensing micro-reactors of a common design.
  • The NRC staff is exploring approaches for streamlining the review of emergency preparedness for licensing NOAK micro-reactors based on considerations such as the possibility that potential accidents would result in low doses at the site boundary and, under certain circumstances, might not require extensive off-site response.

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Other Licensing and Deployment Topics Streamlined processing of license applications and licensing documents

  • Licensing applications referencing an approved micro-reactor design that leverages maximal design standardization will likely be nearly identical, with some possible minor variations related to licensee-specific or site-specific information.
  • NRC-generated licensing documents, such as the NRC staff safety evaluation, environmental review, license, and required Federal Register notices, will likely be very similar for licensing each individual micro-reactor of a common design.
  • The NRC staff is considering approaches for using electronic licensing forms, licensing document templates, and automation to streamline processing and review of micro-reactor applications to reduce the timeframes for acceptance review, docketing, safety review, environmental review, concurrence, license issuance, and other steps.

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Construction inspection Micro-reactors of a common design might be self-contained in that they would be almost entirely fabricated at a factory and require minimal site preparation or construction activities at the deployment site, or they might consist of a core module that is fabricated in a factory and then incorporated into or connected to permanent structures and systems constructed at the deployment site, such as a reactor building and power conversion equipment.

In either case, it will be necessary for the NRC staff to verify completion of ITAAC in support of a finding for authorization to operate under 10 CFR 52.103(g) or to verify substantial completion of construction for issuance of an operating license under 10 CFR 50.56 and 50.57(a)(1).

As discussed in SECY-23-0048*, the NRC staff is considering approaches for risk-informed and performance-based inspections at both the fabrication facility and deployment site that can be completed within the expected timeframes for licensing and deployment of Nth-of-a-Kind micro-reactors.

88 Other Licensing and Deployment Topics

  • SECY-23-0048, "Vision for the Nuclear Regulatory Commission's Advanced Reactor Construction Oversight Program" (ML23061A086)

Next Steps

  • Publish a draft white paper, expected in Fall 2024, to further stakeholder engagement
  • Develop a Commission paper on licensing and deployment considerations for Nth-of-a-kind micro-reactors:
  • Request Commission direction on regulatory approaches for standardizing operational programs and alternate environmental reviews
  • Provide information on other topics, including the NRC staffs related near-term strategies and potential next steps 89

Discussion Items

  • Are there other approaches that the NRC staff should consider for NOAK micro-reactor licensing and deployment?
  • Are there additional strategies the NRC staff should consider for the other identified topics?
  • Other feedback or questions for the NRC staff?

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Closing Remarks September 18, 2024, Periodic Advanced Reactor Stakeholder Public Meeting October 30, 2024, Periodic Advanced Reactor Stakeholder Public Meeting

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