ML23115A004

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Presentation Slides - Periodic Advanced Reactor Stakeholder Meeting 04262023
ML23115A004
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Issue date: 04/26/2023
From: Katie Wagner
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Advanced Reactor Stakeholder Public Meeting April 26, 2023 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 575 470 255#

Time Agenda Speaker 10:00 am - 10:10 am Opening Remarks / Advanced Reactor Integrated Schedule NRC 10:10 am - 10:50 am Insights from Nuclear Innovation Alliance (NIA) Workshop on Improving Advanced Reactor Licensing Efficiency Advanced Reactor Licensing Review Enhancements NIA NRC 10:50 am - 11:50 am Alternative Approaches to Address Population-Related Siting Considerations -

White Paper NRC 11:50 pm - 12:10 pm NRC Engagement with Tribal Nations NRC 12:10 pm - 1:25 pm Lunch Break All 2

Time Agenda (continued)

Speaker 1:25 pm - 1:40 pm Guidance for Reviewing Facility Training Programs NRC 1:40 pm - 2:20 pm Joint NRC/Canadian Nuclear Safety Commission (CNSC) Report on TRI-structural ISOtropic (TRISO) Fuel Qualification NRC 2:20 pm - 2:35 pm Break NRC 2:35 pm - 3:35 pm CNSC-NRC Memorandum of Cooperation Topic of Safety Classification of Structures Systems and Components: Interim Report NRC 3:35 pm - 3:40 pm Future Meeting Planning and Concluding Remarks NRC 3

Advanced Reactor Integrated Schedule of Activities The updated Advanced Reactor Integrated Schedule is publicly available on NRC Advanced Reactors website at:

https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 4

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 5

Advanced Reactor Licensing Efficiency Workshop Summary Report Patrick White (pwhite@nuclearinnovationalliance.org)

NRC Periodic Advanced Reactor Stakeholder Meeting April 26, 2023 6

  • NIA is a think-and-do tank working to ensure advanced nuclear energy can be a key part of the climate solution.
  • NIA identifies barriers, performs analysis, engages with stakeholders and policy makers, and nurtures entrepreneurship through its Nuclear Innovation Bootcamp.

Who is Nuclear Innovation Alliance (NIA)?

7 4/26/23

NIA Licensing Efficiency Workshop was based on prior NIA work with stakeholders on ensuring efficient advanced reactor licensing Nuclear Regulatory Commission Advanced Reactor Applicants Congressional Oversight Improving licensing processes Improving license applications Enabling effective regulation Public Participation Ensuring regulator accountability 8

4/26/23

September 2022 workshop goal was to identify barriers to efficient and effective licensing and share best practices, lessons learned Identify barriers and solutions to efficient advanced reactor licensing April NRC Periodic Advanced Reactor Stakeholder Meeting Share workshop findings Discuss recommendations Solicit stakeholder feedback Discuss possible next steps Public engagement with NRC on specific recommendations Share best practices and lessons learned (Link to Summary Report) 9 4/26/23

September 2022 workshop was held under Chatham House Rules to facilitate open, constructive discussion of licensing experiences Licensing Efficiency Workshop Sessions Session 1: Enhancing communication and project management Session 2: Effectively utilizing regulatory engagement plans and optimizing pre-application interactions Session 3: Ensuring effective and efficient safety evaluation reviews Non Governmental Organization Advanced Reactor Developer Potential Owner/Operator 3

16 8

Workshop Participant Affiliation 10 4/26/23

Major theme: effective communication is key to efficient licensing Advanced Reactor Licensing Activities NRC Technical Staff NRC Management NRC Commission

Public, Policymakers,
Industry, NGOs License Applicant
ACRS, OGC Commissioner Executive Director for Operations Office Director Division Director Branch Chief Project Manager Technical Expert Applicants External Communication Internal Communication 11 4/26/23

Advanced Reactor Licensing Efficiency Workshop presentations and discussions provided insights across 5 major topic areas

1. Achieving and maintaining alignment between applicant and NRC on the licensing review process and creating clear lines of communication
2. Preparing the application content and performing the safety review based on clear, definitive, and consistent expectations
3. Ensuring efficient use of staff resources as the NRC receives an increasing number of advanced reactor license applications
4. Developing processes to identify and resolve challenges encountered during reviews
5. Ensuring uniform understanding and expectations on the role of specific NRC offices and committees in the licensing process 12 4/26/23
1. Achieving and maintaining alignment between applicant and NRC on the licensing review process and creating clear lines of communication NRC Technical Staff NRC Management NRC Commission
Public, Policymakers,
Industry, NGOs License Applicant
ACRS, OGC Communication breakdowns between applicants and NRC or within the NRC can significantly complicate or delay licensing reviews 13 4/26/23
1. Achieving and maintaining alignment between applicant and NRC on the licensing review process and creating clear lines of communication Proactively develop lines of communication at all levels as early as practicable Maintain lines of communication throughout the review process Improve internal NRC communication to ensure alignment, clarity, and predictability on technical and policy positions both:

Within a specific license review Across different license reviews Detailed regulatory engagement plants facilitate staff interaction Milestones help hold applicants and NRC accountable on processes Communication and plan updates based on licensing progress help maintain alignment Recommendation for Applicants Recommendation for NRC Focus: Regulatory engagement plans and specific milestones 14 4/26/23

2. Preparing the application content and performing the safety review based on clear, definitive, and consistent expectations NRC Technical Staff License Applicant Inadequate or incomplete applications and unclear questions or feedback can result in costly and lengthy iteration cycles between applicants and NRC License applications and supporting materials Application feedback and additional questions 15 4/26/23
2. Preparing the application content and performing the safety review based on clear, definitive, and consistent expectations Focus on providing information that enables the NRC staff review Prepare applications that reduce barriers to the reviewer reaching a safety determination Focus on providing clear feedback and information requests to applicants Ensure internal agency alignment on key technical and policy issues Licensing audits can facilitate more effective staff reviews of complex issues Applicants and NRC should document best practices for licensing audits processes Lessons learned should be incorporated into general NRC guidance and process Recommendation for Applicants Recommendation for NRC Focus: NRC Licensing Audits 16 4/26/23
3. Ensuring efficient use of staff resources as the NRC receives an increasing number of advanced reactor license applications Participants report some NRC staff resource challenges for current advanced reactor licensing activities, but licensing review workload could increase dramatically to support commercial deployment in the 2030s Figure from 2023 DOE Report Pathways to Commercial Liftoff - Advanced Nuclear 17 4/26/23
3. Ensuring efficient use of staff resources as the NRC receives an increasing number of advanced reactor license applications Prioritize meeting licensing submittal deadlines provided to NRC staff Inform NRC of changing schedule or resource needs for reviews as early as possible Facilitate NRC management and planning of resources NRC management must keep NRC staff accountable for the technical review:
Depth,
Breadth, Scope, and Regulatory basis NRC PM performance can have significant effects on licensing process outcomes NRC should prioritize the training and organizational management of NRC PMs Additional resources, training, and tools could help promote PM excellence Recommendation for Applicants Recommendation for NRC Focus: NRC Project Managers (PM) 18 4/26/23
4. Developing processes to identify and resolve challenges encountered during reviews Commissioner Executive Director for Operations Office Director Division Director Branch Chief Project Manager Technical Expert Applicants Applicants and NRC have multiple levels of decisionmakers involved when resolving technical or policy questions, and resolution paths for issues may be unclear 19 4/26/23
4. Developing processes to identify and resolve challenges encountered during reviews Proactively share concerns about the licensing process at increasing levels of NRC management Avoid the intentional or inadvertent early escalation to senior management or the Commission Provide regular updates to applicants on both major and minor challenges or questions as they emerge Avoid holding of concerns or question until the end of a review to discuss with applicants Develop or expand guidance for staff on preliminary decisions Assess expedited review procedures for applicants to obtain consistent regulatory interpretations Assess an official escalation or appeal process for technical or policy decisions Recommendation for Applicants Recommendation for NRC Focus: Resolving regulation interpretations and issues 20 4/26/23
5. Ensuring uniform understanding and expectations on the role of specific NRC offices and committees in the licensing process NRC Technical Staff NRC Management NRC Commission
Public, Policymakers,
Industry, NGOs License Applicant
ACRS, OGC Reviews and decisions from ACRS and OGC can have significant impacts on licensing reviews, but their relationship and interactions with other entities may be unclear to applicants 21 4/26/23
5. Ensuring uniform understanding and expectations on the role of specific NRC offices and committees in the licensing process

- Clarify the role of Office of General Counsel (OGC) in licensing reviews so that applicants and staff understand the roles, responsibilities, and scope

- Clarify the role of Advisory Committee on Reactor Safeguards (ACRS) to applicants and staff so they can maximize Committee effectiveness in licensing Clarify expectations for ACRS reviews, interactions with NRC staff and applicants, and the scope of ACRS reviews activities Commission should take a more active oversight role on ACRS activities to ensure it maximizes effectiveness Recommendations for Commission Focus: Aligning stakeholder expectations for ACRS reviews 22 4/26/23

23 Next steps: soliciting feedback, discussing recommendations, and identifying opportunities for sharing lessons learned, best practices 4/26/23 April NRC Periodic Advanced Reactor Stakeholder Meeting Share workshop findings Discuss recommendations Solicit stakeholder feedback Discuss possible next steps Public engagement with NRC on specific recommendations Applicant, NRC, and Commission consideration and possible incorporation of report recommendations Identification of additional opportunities for sharing lessons learned and best practices with applicants, utilities, public, and other stakeholders Next steps on Licensing Efficiency

Advanced Reactor Licensing Review Enhancements John Segala NRR/DANU Advanced Reactor Stakeholder Meeting April 26, 2023 24

NRC Lessons Learned Efforts The Advanced Reactor Program is informed by stakeholder feedback and several NRC staff lessons learned efforts including:

  • Lessons Learned from the NRC Staffs Review of the NuScale Design Certification Application (ML22088A161)
  • Response to the NuScale Design Certification Application Lessons Learned Report (ML22294A144) 25

Enhancing Advanced Reactor Reviews

- Regulatory Review Roadmap (ML17312B567) - Encourages Regulatory Engagement Plans (REPs)

  • NEI 18-06, Guidelines for Development of a Regulatory Engagement Plan (non-public NEI document)

- Pre-application Engagement to Optimize Advanced Reactors Application Reviews white paper

  • Expanded Use of Regulatory Audits

- NRC Office Instruction LIC-111

- Optimization based on lessons learned

  • Optimized use of Requests for Additional Information (RAIs)

- NRR Office Instruction LIC-115

- Management review of RAIs before issuance

  • Transparency through use of Dashboards 26

Enhancing Staff Capability and Capacity

  • Multidisciplinary core review teams to focus reviews
  • Qualification Program for Project Managers

- Office Instruction updated April 2023

  • Building capacity for multiple ongoing reviews

- Hiring new staff

- Training staff on advanced reactor technology

- Use of contractors for flexibility and agility

  • Standardized applications will facilitate efficient reviews
  • Timely information on industry plans supports effective NRC resource planning 27

Successfully Implementing Enhancements

  • Kairos Hermes Test Reactor Construction Permit (CP) review

- Successfully executing 21-month review schedule

  • Dashboards
  • Maximizing the use of audits to optimize RAIs
  • Internal project controls
  • Multidisciplinary core review team
  • Abilene Christian University Molten Salt Research Reactor CP review

- Building off the lessons learned from Kairos review

- Regulatory Engagement Plans

- Successful completion of Topical Report reviews

- Preapplication assessments enhance readiness and quality of application (NuScale, Atomic Alchemy) 28

Next Steps

  • Continue stakeholder engagement through our periodic advanced reactor public meetings and meetings with developers
  • Continue to assess our review processes during ongoing reviews
  • Share best practices with prospective applicants
  • Continue to make enhancements to internal processes based on lessons learned from ongoing reviews and stakeholder input 29

April 2023 30 Alternative Approaches to Address Population-Related Siting Considerations

31 Background

  • SECY-20-0045, Population Related Siting Considerations for Advanced Reactors
  • SRM-SECY-20-0045 dated July 13, 2022

- ML22194A885 The Commission has approved the staffs recommended Option 3, to revise the guidance in Regulatory Guide 4.7, General Site Suitability Criteria for Nuclear Power Stations, related to Title 10 of the Code of Federal Regulations Part 100, Reactor Site Criteria, Section 100.21(h). That provision states that reactor sites should be located away from very densely populated centers and that areas of low population density are generally preferred. The revised guidance will provide technology-inclusive, risk-informed, and performance-based criteria to assess population-related issues in siting advanced reactors. With respect to the traditional dose assessment approach, the staff should provide appropriate guidance on assessing defense-in-depth adequacy and establishing hypothetical major accidents to evaluate.

32 10 CFR 100.21(a) 10 CFR 50.34/52.79 Exclusion Area boundary (EAB) 25 rem - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 10 CFR 100.21(a) 10 CFR 50.34/52.79 Low Population Zone (LPZ) 25 rem - duration 10 CFR 100.21(b)

Population Center Distance (PCD) 11/3 dLPZ Background - Requirements/Guidance 10 CFR 100.21(h)

Located away from populated centers and low population density preferred Regulatory Guide 4.7 (Population Density)

33 Two potential issues identified:

1) 500 persons per square mile (ppsm) out to a distance of 20 miles
2) 500 ppsm close to reactor site used for small communities

- Background and references in ORNL/TM-2019/1197 (ADAMS Accession No. ML19192A102)

Staff developed several options for consideration:

- Option 1 - Status Quo

- Option 2 - Source Term Factor

- Option 3 - Offsite Dose Calculation

- Option 4 - Develop Societal Risk Measures

Background

34 Description Maintain EAB and LPZ for event sequence doses of 25 rem TEDE over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and course of event respectively Maintain distance from densely populated center of more than about 25,000 residents For plants with event sequence doses > 1 rem TEDE over a month beyond the site boundary (DBEs and BDBEs as defined under licensing modernization project (LMP)),

population density < 500 ppsm over the radial distance equal to twice the radius at which 1 rem over a month is estimated Option 3 (Offsite Dose Calculation)

35 Option 3 - Example Cases

36 Option 3 - Example Cases

37 Option 3 - Example Cases

38 Prepared to support public meetings and discussion of future changes to Regulatory Guide 4.7 Preliminary approaches for

- Non-light water reactors under LMP-type methodology

- Light water reactors under traditional methodology

- Non-light water reactors under traditional (non-LMP) methodology Distinctions between:

- Analyses related to estimated doses at EAB/LPZ

- Analyses related to alternative to existing population density guidance (500 ppsm out to 20 miles)

White Paper

39 Source term for siting analysis Additional Information provided in:

Regulatory Guide 1.183 Regulatory Guide 1.233 Footnote (6) - 10 CFR 50.34 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

40 LMP approach for non-LWRs was primary focus of SECY-20-0045 Preliminary white paper methodology

- Analyses related to estimated doses at EAB/LPZ

  • Design Basis Accidents

- Analyses related to alternative to existing population density guidance (500 ppsm out to 20 miles)

  • Design Basis Events and Beyond Design Basis Events

- Outputs used to determine distance at which an event results in 1 rem TEDE over 30 days Non-LWR under LMP

41 SECY-20-0045 mentions using traditional approach (RG 1.183)

SRM directed staff to provide guidance on assessing defense-in-depth adequacy and establishing hypothetical major accidents to evaluate Preliminary white paper methodology

- Analyses related to estimated doses at EAB/LPZ

- Analyses related to alternative to existing population density guidance (500 ppsm out to 20 miles)

  • Accounting for potential containment performance under severe accident conditions LWR under traditional approach

42 SECY-20-0045 mentions using traditional approach (RG 1.183)

SRM directed staff to provide guidance on assessing defense-in-depth adequacy and establishing hypothetical major accidents to evaluate Guidance prepared for non-LWRs relying on containment type design feature as a primary means to limit the release of radionuclides Preliminary white paper methodology

- Analyses related to estimated doses at EAB/LPZ

  • Regulatory Guide 1.183 like analysis for source term used for assessing containment and site-specific information

- Analyses related to alternative to existing population density guidance (500 ppsm out to 20 miles)

  • Accounting for potential severe accidents that challenge the containment Non-LWR under traditional approach

43

  • Prepare draft revision 4 to RG 4.7

- DG 4031

  • Publish draft guidance for public comment

- Target: Fall 2023

  • Resolve public comment

- Target: 1st quarter CY 2024 Moving forward

44 Questions and Discussion

45 Backup Slide - Dose Falloff Source:

NUREG-0396 Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support Of Light Water Nuclear Power Plants, Published Dec 1978 Figure I-1 Dose Falloff with Distance

46 Consequence Based Security (SECY-18-0076)

EP for SMRs and ONTs (SECY-18-0103)

Functional Containment (SECY-18-0096)

Insurance and Liability Siting near densely populated areas Environmental Reviews Licensing Modernization Project Backup Slide - Integrated Approach

47 Backup Slide - Integrated Approach SECY-19-0117: 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

NRC Engagement with Tribal Nations Kevin Williams, Director Division of Materials Safety, Security, State, and Tribal Programs April 26, 2023 Advanced Reactor Public Stakeholder Meeting

Tribal Policy Statement In 2017, the NRC published the Tribal Policy Statement is centered on the following principles:

1.

The NRC recognizes the Federal trust relationship with and will uphold its trust responsibility to Indian Tribes.

2.

The NRC recognizes and is committed to a government-to-government relationship with Indian Tribes.

3.

The NRC will conduct outreach to Indian Tribes.

4.

The NRC will engage in timely consultation, as applicable.

5.

The NRC will coordinate with other Federal agencies, as applicable.

6.

The NRC will encourage participation by State-recognized Tribes.

This principles guide the NRCs government to government interactions with the Tribal Nations.

The NRC does its part in implementing this duty in the context of our jurisdiction and in honoring treaties.

49

Licensing Reviews 50

Key Differences in Tribal Consultation between the National Historic Preservation Act Section 106 and the NRCs Tribal Policy Statement

  • NRCs Tribal Consultation Information Tool ( ML23019A328) 51

NRC Tribal Program Contacts

  • Email Kevin.Williams@nrc.gov
  • Phone: 301-415-3340
  • Email: Booma.Venkataraman@nrc.gov
  • Phone: 301-415-2934
  • Contact the NRCs Tribal Program Team
  • Email: Tribal_Outreach.Resource@nrc.gov
  • NRC General Contact Information

Questions?

53

Lunch Break Meeting will resume at 1:25 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 575 470 255#

Advanced Reactor Stakeholder Public Meeting

Facility Training Program Guidance DRO-ISG-2023-04

  • This ISG is intended to support both applications under the proposed Part 53 as well as near-term applications under Parts 50 and 52.
  • The guidance supports the NRC staff review of the portion of an application associated with the training program for plant personnel, including licensed operator initial and requalification training programs.
  • This guidance also facilitates the review of non-accredited training programs at commercial nuclear plants. This guidance may also be used to support training program inspection needs as currently specified in NUREG-1220.
  • This guidance covers:
  • Scope of facility training programs
  • The 5 phases of the systems approach to training 55

Draft Final Report U.S. NRC Advanced Reactor Stakeholders Meeting April 26, 2023 Kelly Conlon, Canadian Nuclear Safety Commission (CNSC)

Jeff Schmidt, U.S. Nuclear Regulatory Commission (NRC)

Objective and Status

  • Collaboration on pre-application activities to ensure mutual preparedness to efficiently review advanced reactor and SMR designs
  • A number of vendors proposing to use TRISO fuel are engaged in pre-licensing or licensing activities
  • The TRISO assessment is a joint white-paper that can be used to develop regulatory guidance
  • Currently under final management review
  • Completion expected in the second quarter of 2023 57

Assessment Scope

  • Develop a shared, evidentiary basis to support regulatory findings for items that are generically applicable to TRISO
  • Identify items that are design dependent
  • Highlight areas where additional information or testing is needed
  • Focused on the U. S. Department of Energys (DOEs) Advanced Gas Reactor (AGR) program
  • Most applicants intend on using uranium oxycarbide (UCO) fuel kernels of the same or similar design 58

Assessment Scope

  • Leverages the U.S. NRC staffs review of EPRI-AR-1-NP-A, UCO TRISO Coated Particle Fuel Performance topical report (TR) (ML20336A052)
  • TR scope included the AGR-1 and 2 test programs
  • TR focused on TRISO particle attributes that produced AGR program failure fractions and fission product releases
  • The CNSC/NRC assessment will provide an overview of the following:
  • UCO TRISO particle assessment
  • Fuel compact or pebble form attributes
  • Evaluation model capabilities and model assessment
  • The AGR test envelope and the adequacy of AGR test data 59

NUREG-2246 Fuel Qualification for Advanced Reactors NUREG-2246 is a technology inclusive framework that provides criteria that when satisfied support a regulatory finding that a nuclear fuel is qualified

  • Qualified fuel refers to fuel that if built within specifications will perform as described in the safety analysis
  • Primarily developed as a guide for advanced reactor fuel development since extensive guidance already exists for light-water reactor fuel
  • Can be used by applicants to develop or assess existing fuel qualification plans or data
  • Focused on solid fuel forms, NUREG/CR-7299 (ML22339A161) addresses fuel qualification for molten salt fueled reactors 60

UCO TRISO Particle and Fuel Form

  • UCO particle attributes described in EPRI-AR-1-NP-A are sufficient to produce AGR fission product release performance
  • Qualitative standard for the SiC microstructure is subjective and can be difficult to implement
  • Work is currently being performed to characterize the AGR microstructure to better understand as-built grain size distribution
  • Data could be used to develop a quantitative standard for the SiC microstructure
  • Fuel Form (Compact and Pebble) Assessment
  • Review is design specific
  • Need to provide data/testing to demonstrate safety functions are met
  • 40% upper bound packing fraction limit based on the AGR program 61

Evaluation Model Assessment

  • Identifies important geometry, material, physical modeling considerations necessary to develop a TRISO evaluation model Some failure modes may be excluded based on meeting the AGR manufacturing specifications precluding certain failure mechanisms Some failure modes cannot be modeled based on the lack of sufficient data
  • Use experimental data to account for failure modes not modeled
  • Provide justification that the overall failure fraction is sufficiently conservative to account for the mechanisms not modeled
  • Over the tested temperature ranges, there is likely sufficient AGR data to support model validation though the final justification of data sufficiency is the responsibility of the applicant
  • Design-specific evaluation models are anticipated 62

Test envelope should be consistent with irradiation tests covering expected design-specific normal operation and transient conditions (i.e., the performance envelope)

  • Maximum steady-state irradiated parameters per EPRI-AR-1-NP-A
  • Based the low failure rate at 1600 oC during AGR safety testing
  • AOO peak particle temperature < 1600 oC could be warranted based on design-specifics
  • Applicant required to demonstrate that SARRDL and appropriate dose criteria or limits are met
  • Higher peak AOO TRISO particles temperatures could be justified
  • 1700 oC target peak design basis accident (DBA) particle temperature
  • Based on AGR data showing an increase in failure rate from 1700 to 1800 oC
  • DBA peak particle temperature < 1700 oC could be warranted based on design-specifics
  • Applicant required to demonstrate the appropriate dose criteria or limits are met
  • Higher peak DBA TRISO particles temperatures could be justified 63 Test Envelope
  • AGR safety testing did not include overpower transient testing such as rod withdrawal or rod ejection type reactivity insertions
  • Failure fractions assumed to be a function of absolute temperature, but rate of change could lead to other failure modes (e.g., melt, kernel swelling induced coating stresses)
  • Based on NGNP project, transients 1 second have a negligible temperature change across the particle due to the thermal time constant
  • Short time constant allows for energy dissipation to the surrounding environment
  • Overpower transients 1 second expected to have a negligible increase in failure fractions as compared to other means (e.g., absolute temperature)
  • Overpower transients should still be evaluated based on the failure mechanisms associated with absolute temperature
  • For overpower transients < 1 second, additional justification is needed to demonstrate a non-conservative failure fraction is predicted 64 Test Envelope
  • The quality of the AGR 1 and 2 test data (and hence TRISO particle development) judged to be of sufficient quality for licensing applications
  • Experimental uncertainties in EPRI-AR-1-A, Section 6.5 provide acceptable measurement uncertainties for use in licensing applications
  • AGR program test conditions constructed to match the expected operating condition of HTGRs with full scale TRISO particles
  • Test conditions match the expected operating conditions
  • No particle scale distortion
  • Distortions caused by compact or pebble geometry can be accommodated analytically if the matrix material is well characterized 65 Test Envelope

Conclusions This report establishes a common regulatory position on TRISO fuel qualification based on existing knowledge (e.g., AGR program) and identifies design-specific analytical or testing gaps that should be addressed to enable TRISO use in licensing applications.

  • AGR program provided end-state attributes and established manufacturing specifications to produce fuel with fission product retention capabilities to support expected licensing applications
  • The extent and quality of the AGR 1 and 2 data, both steady-state irradiation and safety testing, may be sufficient for evaluation model development over the range of conditions tested
  • Additional test data, beyond the current AGR program safety test data, is not needed for overpower transients with durations 1 second
  • For overpower transients < 1 second, additional justification needed to address potential failure mechanisms based on a large temperature differential across the particle
  • Fuel compact or pebble is expected to be design-specific and the applicant will be responsible for qualifying compact/pebble designs that meet their safety functions
  • 40% upper bound packing fraction established 66

Questions?

Questions for U.S. NRC:

jeffrey.schmidt2@nrc.gov; 301-415-4016 Questions for CNSC:

mediarelations-relationsmedias@cnsc-ccsn.gc.ca; 613-996-6860 67

Break Meeting will resume at 2:35 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 575 470 255#

Advanced Reactor Stakeholder Public Meeting

NRC-CNSC MOC Interim Joint Report on Classification of Structures, Systems and Components 69

  • Work Plan
  • Scope of Safety Classification Project
  • Interim Report Findings Safety Classification Comparison and Effects Pilot Design Rule Comparisons
  • Engineering Design Rule Inputs
  • Questions Agenda 70
  • Identify key similarities and differences in the engineering design rules and specifications applied to each safety class and how this impacts the outcomes
  • Review how each organization applies existing codes and standards and interacts with Standards Development Organizations (SDOs) to verify appropriate codes and standards are being developed, applied, and endorsed.

Objectives of Work Plan 71

Scope of Work 72

  • New Water-Cooled Small Modular and Advanced Non-Water-Cooled Reactors
  • Safety Classification Processes:

Programmatic Specific Design Hazard Protection Reliability Assurance (Design, Maintenance, and Availability)

Pressure Retaining Components Seismic Design Quality Assurance (Construction)

Civil Structures Fire Protection Testing and Inspection Electrical and I&C Equipment Qualification

  • Application of Engineering Design Rules:
  • Safety Analysis Deterministic Probabilistic
  • Initiating Event Determination
  • Safety Functions
  • Consequence Assessment
  • Classification of Structures, Systems, and Components (SSCs)
  • Assignment of Engineering Design Rules by Classification Safety Classification Process 73
  • 10 CFR 50.40, Common Standards, states: In issuing a construction permit or operating license under 10 CFR Part 50 or an early site permit, combined operating license, or manufacturing license under Part 52, the Commission will be guided, in part, by:

reasonable assurance of compliance with the regulations of 10 CFR Part 50 adequate protection of the public health and safety NRC Licensing Approach 74

  • A safety assessment of the site and facility, including:

contained radioactive materials application of engineering standards safety features and barriers to release of radioactive material analysis of a postulated fission product release

  • An assessment of the design of the facility, including:

principal design criteria (PDC) relationship of the facility design bases to the PDC analysis and evaluation of the design and performance of SSCs to assess the risk to public health and safety NRC Safety Analysis Elements 75

  • 10 CFR 50.2: Safety-related SSCs means those SSCs relied upon to remain functional during and following design basis events to assure:

The integrity of the reactor coolant pressure boundary; The capability to shut down the reactor and maintain it in a safe shutdown condition; or The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures.

  • Influences requirements for traditional safety analysis and application of engineering design rules Definition of Safety-Related 76
  • Deterministic structure Single failure criterion Conservative analytical methods Reliance on safety-related SSCs Acceptance criteria related to initiating event frequency
  • Design-specific probabilistic analyses provide risk insights and confirm safety goals would be met NRC Traditional Approach 77

Initiating Event Category AOO DBA SSC Availability Safety-Related SSCs with Single Failure; with and without Offsite Power; other SSCs with Technical Justification Safety-Related SSCs with Single Failure; with and without Offsite Power Pressure Boundary Within 110% of Design Within Acceptable Design Limits Fuel Within Specified Acceptable Fuel Design Limits Cladding Failure if Specified Acceptable Fuel Design Limit Exceeded Dose 10 CFR Part 20 Accident Dose Limit (25 Rem TEDE) or Small Fraction of Limit Consequential Failures No Escalation without other Independent Faults No Consequential Failures of SSCs Necessary to Mitigate Fault Loss of Coolant Accident Not Applicable 10 CFR 50.46 Criteria Analysis Acceptance Criteria 78

  • Safety-Related SSCs relied on to meet analysis acceptance criteria for safe shutdown (including pressure boundary)

SSCs credited for mitigation of dose consequences

  • Important to Safety Functions identified in PDC Special purpose regulations for defense in depth Regulatory Treatment of Non-Safety Systems (RTNSS)
  • Risk-informed safety classification per 10 CFR 50.69 NRC Traditional Classification 79
  • NRC regulations associate application of certain rules based on SSC safety classification, for example:

Quality assurance for activities affecting the safety-related functions of SSCs Seismic design criteria for safety-related SSCs Inservice testing and inspection of safety-related SSCs (ASME Code per 10 CFR 50.55a, water-cooled reactors)

Environmental qualification of important to safety SSCs

  • Other rules applied on a graded basis (GDC-1)

Engineering Design Rules 80

  • Technology-inclusive, risk-informed, and performance-based licensing process NEI 18-04 endorsed for licensing of advanced reactors within NRC regulatory framework (RG 1.233)

Establishes methods for the following:

Definition, categorization, and evaluation of events SSC classification, performance criteria, and design rules Evaluation of defense in depth adequacy

  • Informs safety design to demonstrate compliance Licensing Modernization Project (LMP) 81

LMP Event Classification 82 Anticipated Operational Occurrences (AOOs) have frequencies of 10-2 per plant-year or higher Design basis events (DBEs) have frequencies of 10-4 or higher and less than 10-2 per plant-year Beyond design basis events (BDBEs) have frequencies less than 10-4 per plant-year Design basis accidents are event sequences derived from DBEs to set safety related SSC performance criteria

LMP Safety Classification 83

  • Risk-informed with consideration of uncertainty
  • Events evaluated with consideration of sequence frequency and consequences
  • Criteria address cumulative and sequence risk
  • Explicit consideration of defense in depth
  • Assignment of engineering design rules considers safety classification and SSC safety function LMP Attributes 84
  • NRC regulations provide for specific exemptions that:

Are authorized by law Will not present undue risk to public health and safety Are consistent with the common defense and security Supported by one or more special circumstances

  • Special circumstances include:

Compliance not necessary to achieve the underlying purpose Safety benefit compensates for any decrease in safety NRC Exemption Process 85

  • Nuclear Safety Control Act (NSCA) compliance required
  • CNSC promulgates REGDOCs to meet NSCA REGDOCs include requirements and guidance Applicants may show intent of requirement has been addressed by other means; Commission determines if requirement is met
  • Safety analysis expectations captured in:

REGDOC 2.4.1, Deterministic Safety Analysis REGDOC 2.4.2, Probabilistic Safety Assessment for Nuclear Power Plants REGDOC 2.5.2, Design of Reactor Facilities: Nuclear Power Plants CNSC Approach 86

Analysis Acceptance Criteria 87 Event Category AOO DBA (or AOO with DID Level 2 Failure)

BDBA SSC Availability No Single Failure Single-Failure Affecting Safety System Group No Single Failure Analysis Methods Best Estimate (DID Level 2)

Conservative Analysis or Best Estimate plus Evaluation of Uncertainties (DID Level 3)

Best Estimate (DID Level 4)

Fuel and SSC Limits Within Specified Acceptable Design Limits; No Unanalyzed Conditions Within Specified Acceptable Design Limits; No Unanalyzed Conditions Evaluate Ability to Restore or Maintain Safety Functions Dose 0.5 millisievert (mSv) 20 mSv Safety Goals Consequential Failures Prevented to the Extent Practicable Prevented to the Extent Practicable Avoid Cliff-Edge Effects; Prevent Early Containment Failure

  • Defence in depth explicitly considered
  • Five levels of defence:
1. Prevent deviation from normal operation
2. Prevent AOOs from escalating to accident conditions; control systems acting alone prevent SSC damage
3. Minimize accident consequences; safety systems acting alone mitigate all AOOs and DBAs within dose criterion
4. Minimize radiological release from severe accident; probabilistic analyses demonstrate safety goals are met
5. Mitigate consequences of release Defence in Depth 88
  • All SSCs identified as important to safety (ITS) or not important to safety
  • Safety-significance of ITS SSCs based on:

safety function(s) to be performed consequence(s) of failure probability that the SSC will be called upon to perform the safety function time following a initiating event at which the SSC will be called upon to operate, and the expected duration of that operation

  • Applicant may propose graded classification of ITS SSCs CNSC Safety Classification 89
  • REGDOC 2.5.2 provides guidance for assigning engineering design rules
  • Rules should be determined based on safety classification and include the following categories:

Codes and standards Safety margins Reliability Equipment qualification Provisions for inspection, testing, and maintenance Organizational quality assurance CNSC Assignment of Design Rules 90

Compare calculated consequences against performance criteria that vary with event frequency.

Analysis Identification of initiating events; categorization of events by frequency Events Applicant proposes initial design and iterates to meet similar performance goals.

Design Safety Analysis Similarities 91

Framework CNSC NRC Traditional NRC LMP Mitigating SSCs Important to Safety SSCs Safety Related (SR) only SR only - Performance Criteria and Consequence Analysis AOO Analyses Sequence Frequency; Best Estimate Initiating Frequency; Conservative Analysis Sequence Frequency; Best Estimate w/Uncertainty Accident Analyses Sequence Frequency; Conservative or Best Estimate w/Uncertainty Guidance for Event Selection; Conservative Sequence Frequency; Best Estimate w/Uncertainty Beyond Design Basis Sequence Frequency; Best Estimate Special Regulations; Best Estimate Sequence Frequency; Best Estimate w/Uncertainty Probabilistic Analyses Complementary Confirmatory Foundational Safety Analysis Differences 92

Outcome of Safety Classification 93

  • Assumptions:

Identical single-reactor plant for deployment in U.S. and Canada Applicant uses safety analysis method consistent with a selected regulatory framework Applicant develops probabilistic analysis for confirmation of defense in depth and support of risk-informed decision-making Leveraging Prior Approvals 94

  • Leveraging NRC Framework Outcome for CNSC Application Conformance with CNSC regulatory requirements expected; risk-informed processes support justification of alternate means Demonstrate conformance with defence in depth and engineering design rule assignment using risk informed processes Leveraging Prior Approvals (Cont) 95
  • Leveraging CNSC Framework Outcome for NRC Application Development of principal design criteria (PDC), definition of SSCs considered equivalent to safety-related, and application of design rules Reconcile differences in safety analysis necessary to satisfy PDC (analysis of AOOs) and definition of safety-related SSCs credited for mitigation AOO categorization (initiating event or full sequence frequency) and acceptance criteria Conformance with applicable special purpose regulations (exemption)

Address conformance with standard review plan for water-cooled reactors Verify defense in depth Leveraging Prior Approvals (Cont) 96

  • Reliability Assurance Programs Establishes engineering design rules applied to intermediate safety-significance SSCs Program consistent with risk-informed classification processes
  • Pressure Retaining Components and Supports Functional Classification (light water SMRs only):

Functional classification results in the application of ASME BPVC Section III, Division 1 Differences in functional classification increase for lower safety-significance SSCs Risk informed, technology neutral classification guidance likely to support consistent application of codes to individual SSCs (SMRs and Advanced Reactors)

Design Rule Insights 97

  • Finalize engineering design rules topic area input addressing similarities, differences, and impacts Next Steps 98 Expected release of final report in Summer 2023 Programmatic Specific Design Hazard Protection Reliability Assurance (Design, Maintenance, and Availability)

Pressure Retaining Components Seismic Design Quality Assurance (Construction)

Civil Structures Fire Protection Testing and Inspection Electrical and I&C Equipment Qualification

Questions?

THANK YOU 99

Future Meeting Planning

  • The next periodic stakeholder meetings are scheduled for the following dates in 2023: June 7, July 20, and September 14.
  • If you have suggested topics, please reach out to Steve Lynch at Steven.Lynch@nrc.gov 100

How Did We Do?

  • Click link to NRC public meeting information:

https://www.nrc.gov/pmns/mtg?do=details&Code=20230268

  • Then, click link to NRC public feedback form:

101

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