ML21274A076

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Sept 29 2021 Adv Rx Stakeholder Meeting Slides
ML21274A076
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Issue date: 09/29/2021
From: Prosanta Chowdhury
NRC/NRR/DNRL/NRLB
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Hoellman J
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Download: ML21274A076 (158)
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Advanced Reactor Stakeholder Public Meeting September 29, 2021 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#

Slide 1

Time Agenda Speaker 10:00 - 10:10 am Opening Remarks NRC 10:10 - 10:40 am Draft White Paper on Licensing Strategies for Micro-Reactors NRC 10:40 - 11:10 am Overview of Idaho National Laboratory (INL) Report, INL-EXT-21-61847, Regulatory Research Planning for Micro-Reactor Development INL 11:10 - 11:40 am Summary of Technical Reports on Transportation and Storage of AdvancedReactor Fuel NRC 11:40 am - 12:10 pm Draft White Paper, ARCAP Guidance for Fire Protection (Operations)

NRC 12:10 - 1:00 pm Lunch Break All 1:00 - 1:45 pm NRC and Canadian Nuclear Safety Commission (CNSC) Joint Report: Technology Inclusive and Risk-Informed Reviews for Advanced Reactors NRC and CNSC 1:45 - 3:15 pm Update on Draft NUREG-2246, Fuel Qualification for Advanced Reactors" NRC 3:15 - 3:25 pm Break All 3:25 - 4:10 pm Update on White Paper for Scalable Human Factors Engineering and Flexible Staffing for Advanced Reactors NRC 4:10 - 4:55 pm Draft Regulatory Guide to Endorse ASME Section XI, Division 2, Reliability and Integrity Management (RIM) Programs for Non-Light Water Reactors NRC 4:55 - 5:00 pm Future Meeting Planning and Concluding Remarks NRC Slide 2

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/details#advSumISRA Slide 3

Advanced Reactor Stakeholder Meeting Draft White Paper on Micro-Reactor Licensing Strategies ADAMS Accession No. ML21235A418 September 29, 2021 Slide 4

Overview

  • Purpose and Scope
  • Discussion
  • Licensing Strategies
  • Operational Programs
  • External Hazards and Siting
  • Environmental Review
  • Next Steps and Questions Slide 5

Purpose

  • Develop optional strategies to streamline licensing of micro-reactors to support growing interest in micro-reactors Slide 6

Scope

  • Licensing strategies under existing regulations
  • Potential efficiencies gained at Combined License (COL) stage
  • Strategies for streamlining environmental reviews
  • Providing recommendations to advanced reactors rulemaking working group on potential rulemaking changes Slide 7

Licensing Strategies

  • Standardization through design certification:

- No site-specific design features relied on for safety

- COL Items not used

- Bounding site parameters at design

- Operational programs reviewed at design (e.g.,

topical reports)

Slide 8

Licensing Strategies

  • Standardization through COL:

- Applications are fully standardized

- Advanced Nuclear Reactor Generic Environmental Impact Statement (ANR GEIS) fully leveraged

- Design center review approach utilized Slide 9

Licensing Strategies

  • Standardization through manufacturing license:

- Could be utilized to build micro-reactors in a factory

- Could reduce need for inspections and verifications at deployment site Slide 10

Operational Programs

  • Use of standardized operational programs to streamline COL review through:

- NRC approved topical reports and templates

- NRC review of operational programs within design certification being considered

- Two Groups of Op. Programs -

  • Group 1 - Material to the findings on the design

- Part 53 rulemaking working group exploring options to provide finality to technical specifications at design certification stage

  • Group 2 - Not material to adequacy of the design Slide 11

Operational Programs

- Providing finality in design certification for Groups 1 and 2 programs would constitute policy change and require Commission approval

- May address Group 2 programs in topical reports proposing standardized operational programs, which a COL application could reference Slide 12

External Hazards and Siting

  • Under Part 52 design certification, site parameters compared with actual site characteristics
  • Actual characteristics of the site should fall within site parameters in the design certification
  • If design certification identifies parameters for which design is bounding for several sites, then COL review can be reduced in this area
  • Simplified flow chart developed to facilitate external hazards screening Slide 13

EXH-C within EXH-P?

Identify external hazards (EXH) in region of Micro-Rx N

Y Identify Micro-Rx vendors EXH postulated site parameters (EXH-P)

Site Investigation to characterize EXH site (EXH-C)

Compare EXH-C with EXH-P End assessment; document results Design reanalyzed and EXH-C within modified EXH-P?

Y N

A Evaluate mitigating actions End assessment; document results End assessment; document results U

Site not suitable for deployment of this Micro-Rx Key:

Y = yes N = no U = unacceptable performance A = acceptable performance Flowchart for an acceptable site hazard parameters assessment in the micro-reactor design process Slide 14

External Hazards and Siting

- Geology, seismology, meteorology, hydrology, etc.

  • Recent advances in seismic hazards include:

- Graded approach depending on seismic design category

- Development of probabilistic seismic source and ground motion models for Central and Eastern United States (CEUS)

- Expanded Senior Seismic Hazards Analysis Committee (SSHAC) guidance for Study Levels 1 and 2

- Updated American Nuclear Society (ANS) Standards 2.27, 2.29 and American Society of Civil Engineers (ASCE) 43-19 Slide 15

Possession and Transportation of Special Nuclear Material

  • Fuel reactor in factory and then transport to site of usage

- Domestic licensing of byproduct material (10 CFR Part 30)

- Domestic licensing of source material (10 CFR Part 40)

- Domestic licensing of special nuclear material (10 CFR Part 70)

- Transportation package certification (10 CFR Part 71) (see slides from August 26, 2021, meeting - ML21237A463)

- Physical protection of plants and materials (10 CFR Part 73)

- Material control and accounting of special nuclear material (10 CFR Part 74)

Slide 16

Environmental Review

  • Statutory Requirements

- National Environmental Policy Act

- Additional Laws - "The NEPA Umbrella"

  • Consultation
  • Compliance
  • Administrative
  • Implementing Regulatory Framework

- 10 CFR Part 51 Slide 17

Environmental Review

  • Actions Under Current Framework

- Guidance

- Generic Environmental Impact Statement

  • Potential Change to Framework

- Enhancing Review Flexibility

Slide 18

Next Steps

  • We welcome your questions and feedback on the draft micro-reactor white paper.

Prosanta.Chowdhury@nrc.gov Amy.Cubbage@nrc.gov Margaret.OBanion@nrc.gov Slide 19

Regulatory & Licensing Challenges in Microreactors NRC Advanced Reactor Stakeholder Meeting September 29, 2021 Jason Christensenl Regulatory and Licensing Engineer, Idaho National Lab Slide 20

Regulatory Research Plan In collaboration, INL and ORNL have developed a Regulatory Research Plan (RRP) for the Microreactor Program This work links major research activities in microreactor technologies, as sponsored by DOE Office of Nuclear Energys (DOE-NE), to key regulatory requirements and licensing challenges likely to affect deployments in the domestic commercial energy market The RRP examines and prioritizes key research and development opportunities and recommends activities to DOE-NE R&D planners and investigators concerning the expected impact of such work to the microreactor licensing critical path Initial report was completed March 5, 2021 Slide 21

Industry Survey Industry Survey was sent to members of the Advanced Reactor and Microreactor NEI Working Groups Stakeholders were asked to rank the following items by criticality:

Autonomous and Remote Control/Monitoring Grid Interaction Factory Assembly Transportation Staffing Digital Controls Instrumentation Modeling and Simulation Siting and Environmental Impact Security and Safeguards Slide 22

Industry Survey (Contd)

In addition to the rankings, stakeholders were asked:

Is this list complete? If not, what subject areas should be added?

Please provide any specific R&D items that DOE can support through funding and laboratory expertise, technology, and assistance, and that would benefit the licensing of microreactor designs.

Slide 23

Initial Survey Results The initial survey results were grouped into bands based on importance and time criticality The areas of highest priority are band 1 and are needed before subsequent bands Band Topic Area 1

Autonomous and Remote Control/Monitoring Modeling and Simulation 2

Transportation Siting and Environmental Impact Security and Safeguards Factory Assembly 3

Operations, Maintenance, and Security Staffing Grid Interaction Digital Controls Instrumentation Slide 24

Items with Multiple Areas of Regulatory Need Transportation:

There are multiple stages of transportation throughout the life of a microreactor

  • From factory to use site (fueled but not yet operational)
  • Between use sites (post-operation)
  • From use site to disposition process facility (spent fuel for disposal)

Each of these stages of transport are unique and will require the applicant to meet different regulations, such as:

  • Transport Container Design
  • Shielding for each operational stage
  • Shipping Type (air, train, ship, truck)
  • Emergency response Slide 25

Items with Multiple Areas of Regulatory Need (contd)

Autonomous and Remote Control/Monitoring Many consider this to be one area This encompasses many different licensing areas

  • Autonomous reactor operation
  • Remote Control/Monitoring
  • Digital Controls
  • Operator Licensing
  • Number of Operators per Reactor Most of these areas are new or would require new licensing approaches Slide 26

Additional Survey Results Other areas of regulatory framework development need identified by stakeholders include:

Aircraft Impact Assessment Emergency Planning (currently an NRC rulemaking in progress-see 10CFR 50.160)

Control Room Design Siting and Environmental concerns Radiation Protection Waste Management Regulatory Oversight (including oversight of remote and autonomous reactors)

Manufacturing Licenses (including fueling and defueling in a factory setting)

Slide 27

Path Forward INL staff will utilize this report to continue expanding on the survey results to better identify specific R&D projects to recommend INL and ORNL staff have completed updating the RRP draft with the survey information and have proposed R&D activities that will directly support the Microreactor Program Some identified areas do not provide detailed R&D activities because of proprietary information withheld by the developers Final Regulatory Research Plan draft is being used to support FY22 DOE Microreactor Regulatory Development Planning Slide 28

Recommendation for Future Work: Transition from Shipping to Operational Status Microreactors are slated to have modular construction techniques that will ship major components or possibly near complete units from the factory to approved siting locations Reduces complexity of on-site assembly and construction The transition from shipment regulations (10CFR Part 71) to an operating license under 10CFR Parts 50/52 is not currently addressed by regulations 10CFR Part 53 will need to address this missing piece (likely under Subpart E, which contains construction aspects and inspections, tests, analyses, and acceptance criteria)

Key Question: What regulatory process will be used during the transition of a mobile reactor from 10 CFR Part 71 to 10CFR Part 50/52 or Part 53?

Deliverable: Regulatory Strategy Report that outlines the transition that could be submitted to NEI/NRC for comment or endorsement Slide 29

FY22 Planned Work: Guidance for Manufacturing Licenses Some microreactor developers are planning to fabricate and fuel their reactors at a manufacturing facility NRC Manufacturing Licenses are covered under 10CFR Part 52 Subpart F Factory fueling is not covered in Subpart F 10CFR Part 53 will contain a subpart covering manufacturing licenses (Subpart E)

At this time, no factory fueling regulations are being included Key Question: What type of regulatory requirements for the factory construction and fueling of a microreactor are required? ?

Deliverable: A report that provides recommendations for the development of regulations to allow factory fueling and guidance for developers for those recommendations Slide 30

Guidance for Manufacturing Licenses INL Staff will review current manufacturing license regulations to determine what areas would need to be modified to allow for factory fueling Propose changes to regulations to allow for factory fueling, refueling, and defueling a microreactor Develop guidance for obtaining a manufacturing license to perform these tasks The report(s) generated by this work will be submitted for review and potential endorsement by NEI and/or NRC Slide 31

Conclusions INL and ORNL staff continue to seek new and updated needs to support the development and deployment of microreactors For more information, request a survey, or to provide specific research input, please contact:

Jason Christensen INL Regulatory Support Engineer Jason.Christensen@inl.gov Slide 32

Slide 33

Approach for Assessing Storage and Transportation of Advanced Reactor Fuels September 29, 2021 Nick Hansing Materials Engineer Division of Fuel Management Office of Nuclear Material Safety and Safeguards Slide 34

Assessment Process for Advanced Reactor Fuels

  • Survey state of knowledge on candidate technology
  • Identify information needs
  • Determine approach to address needs
  • Take action as appropriate on desired approach Slide 35

Technical Reports

  • NRC used the Center for Nuclear Waste Regulatory Analyses (CNWRA) for support in assessment process
  • CNWRA provided a series of reports that surveyed the state of knowledge and identified potential challenges/information needs to enhance the efficiency and effectiveness of NRC regulatory reviews
  • Focus narrowed to potential information needs of storage and transportation applications for the following advanced reactor fuels (ARF):

- tristructural isotropic (TRISO) fuel

- nuclear metal fuel Slide 36

Subject Reports

  • Transportation of Fresh (Unirradiated) ARF types

- Literature Review: ML20184A151

- Identification of Potential Challenges: ML20209A541

- Identification of Potential Information Needs: ML21021A326

  • Transportation of Spent (Irradiated) ARF types

- Literature Review and Identification of Potential Challenges: ML20237F393

  • Storage of Spent (Irradiated) ARF types

- Literature Review: ML20211L885

- Identification of Potential Challenges: ML20022A217

  • Final Summary Report (not yet publicly available)

Slide 37

Key Messages

  • Wide variability in potential designs
  • Regulatory framework currently provides sufficient requirements and guidance to allow the NRC staff to make a finding of reasonable assurance of adequate protection to public and workers for applications for advanced reactor fuels
  • Some topics could use enhancement

TRISO Results: Fresh Fuel Transportation

  • Very few documents on the topic, but no record of observed or postulated degradation mechanisms
  • Versa-Pac VP-55 and VP-110 (CoC 9342)
  • Already certified for fresh TRISO fuel transport
  • Challenges/information needs? Criticality safety Slide 39

TRISO Results: Spent Fuel Transportation

  • Very limited information - no records of observed degradation or damage of the fuel or any safety issues during transportation
  • TN-FSV cask (CoC 9253) could possibly be used to transport spent TRISO fuel
  • Challenges/information needs? Criticality safety Slide 40

TRISO Results: Storage

  • Fort St. Vrain Experience - no observed aging effects or degradation of fuel
  • Arbeitsgemeinschaft Versuchsreaktor (AVR) Experience - steel cans leaked in wet storage; casks retained potential releases
  • Challenges/information needs? Criticality safety, Corrosion of non-fuel hardware, Mechanical properties of materials Slide 41

Metal Fuel Results: Fresh Fuel Transportation

  • Very few documents on the topic, but no record of observed or postulated degradation mechanisms

- UNC-2600 (CoC 5086)

- ES-3100 (CoC 9315)

- Versa-Pac Models VP-55 and VP-110 (CoC 9342)

  • Challenges/information needs? Criticality safety, Sodium behavior, Mechanical properties of materials Slide 42

Metal Fuel Results: Spent Fuel Transportation

  • No records of fuel degradation/damage or cask safety issues during:

- FFTF moves using T-3 cask

- Fermi 1 using stainless steel canisters

- EBR-II moves using TN-FSV, NAC-LWT and HFEF-5 and HFEF-14

  • Challenges/information needs? Criticality safety, Sodium behavior, Mechanical properties of materials, and See next slide Slide 43

Metal Fuel Results: Spent Transportation Options

- NLI-1/2 cask

  • (CoC No. 9010, spent metal fuel - expired CoC)

- T-3 cask

  • (CoC No. 9132, spent metal fuel - expired CoC)

- NAC-LWT cask

  • (CoC No. 9225, spent metal fuel)

- RH-TRU 72-B cask

  • (CoC No. 9212, remote-handled transuranic wastes - outside of CoC scope )

- TRUPACT-II cask

  • (CoC No. 9218, contact-handled transuranic wastes - shielding needed, possibly requiring drop tests, likely outside of CoC scope)

Slide 44

Metal Fuel Results: Storage

  • ~2 metric tons of heavy metal EBR-II fuel was in wet storage
  • Some water in-leakage resulted from improper sealing
  • Department of Energy is moving fuel to dry storage
  • Challenges/information needs? Criticality safety, Corrosion of non-fuel hardware, Sodium behavior, Mechanical properties of materials Slide 45

Corrosion of non-fuel hardware

  • New fuel designs may have new materials and new container environments (TRISO use of graphite, for instance)
  • Analysis for chemical, galvanic, or other reactions needed
  • One such aspect could be the presence of solid fluoride salt residue from fluoride salt cooled reactors, which generates fluorine under radiolysis Slide 46

Mechanical properties of materials

  • New materials are envisioned for both TRISO and metal fuels
  • Legacy data for metal fuel
  • New information for new materials and designs - modeling, testing, analyses Slide 47

Criticality Safety

  • Criticality performance for these fuels can be complicated by variations in enrichment and burnup
  • Criticality experiments may be needed to address variations
  • If burnup credit is sought, methods may need to be developed
  • Isotopic depletion & criticality (keff) computer codes may need validation Slide 48

Sodium Behavior

  • Limited testing or analysis for sodium-containing metal fuel
  • Sodium creep and location shift due to vibration and/or drop
  • Reactions with water
  • Thermal performance of bonding Slide 49

Path Forward

  • Obtain input from stakeholders on reports, particularly challenges and information needs

- Letters sent to various stakeholders encouraging pre-application interactions

- Feedback sought in this meeting as to the desirability of a public meeting/workshop to discuss what would be expected in fuel facility and storage and transportation applications?

  • Develop review guidance enhancements (TBD)
  • Continue assessment process for additional technologies Slide 50
  • NRC welcomes pre-application engagement on potential storage and transportation applications.

Nicholas.Hansing@nrc.gov Slide 51

Acronyms

  • ARF: Advanced Reactor Fuel
  • AVR: Arbeitsgemeinschaft Versuchsreaktor
  • CNWRA: Center for Nuclear Waste Regulatory Analyses
  • EBR-II: Experimental Breeder Reactor-II
  • FFTF: Fast Flux Test Facility
  • HALEU: high-assay low-enriched uranium
  • LWR: Light water reactor Slide 52

Stakeholders Meeting 1

Advanced Reactor Content of Application Project (ARCAP)

Risk-informed, Performance-Based Fire Protection Program (for Operations)

Interim Staff Guidance (Draft)

Slide 53

=

Background===

2 The Advanced Reactor Content of Application Project (ARCAP) has been developing guidance to support the review of non-light water reactors (non-LWRs), modular LWRs and stationary micro-reactors.

The guidance has been developed as draft Interim Staff Guidance (ISG) documents in the form of draft white papers.

One of those draft documents is the ISG on Risk-informed, Performance-Based Fire Protection Program (for Operations)

(ML21253A134).

The purpose of this ISG is to facilitate the review of advanced reactor applications that use a risk-informed, performance-based approach to develop their fire protection programs.

Slide 54

ARCAP and Technology Inclusive Content of Application Project (TICAP) - Nexus 3

Outline Safety Analysis Report (SAR) -

Based on TICAP Guidance

1. General Plant Information, Site Description, and Overview of the Safety Case 2.

Methodologies and Analyses 3.

Licensing Basis Event (LBE) Analysis 4.

Integrated Evaluations 5.

Safety Functions, Design Criteria, and SSC Safety Clasification

6. Safety Related SSC Criteria and Capabilities 7.

Non-safety related with special treatment SSC Criteria and Capabilities 8.

Plant Programs Additional Portions of Application

  • Technical Specifications
  • Technical Requirements Manual
  • Quality Assurance Plan (design)
  • Quality Assurance Plan (construction and operations)
  • Physical Security Plan
  • SNM material control and accounting plan
  • Radiation Protection Program
  • Inservice inspection/Inservice testing (ISI/IST) Program
  • Environmental Report
  • Site Redress Plan
  • Exemptions, Departures, and Variances
  • Facility Safety Program (under consideration for Part 53 applications)

Audit/inspection of Applicant Records Calculations Analyses P&IDs System Descriptions Design Drawings Design Specs Procurement Specs Probabilistic Risk Assessment Safety Analysis Report (SAR) structure based on clean sheet approach Additional SAR Content -Outside the Scope of TICAP

9. Control of Routine Plant Radioactive Effluents, Plant Contamination, and Solid Waste
10. Control of Occupational Doses
11. Organization
12. Initial Startup Programs
  • Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.

Slide 55

Background (cont.)

4 The guidance in this ISG can be applied to any non-LWR, small modular LWR or stationary micro-reactor applying for an operating licensing (OL) and combined license (COL) under Title 10 of the Code of Federal Regulations (10 CFR) Part 50 or 52.

This document does not provide guidance regarding the licensing requirements for fire protection programs needed to be in place prior to receipt of byproduct, source, or special nuclear material under 10 CFR Parts 30, 40, and 70. A construction permit (CP) applicant may address these fire protection licensing requirements within its CP application (in accordance with 10 CFR 50.31) or separately from the CP application.

The ISG will be updated to apply to applications under 10 CFR Part 53, when 10 CFR Part 53 is issued.

Slide 56

Requirements 5

10 CFR 50.48(a), which requires that each operating nuclear power plant have a fire protection plan that meets the requirements of either 10 CFR Part 50, Appendix A, Criterion 3 for LWRs or the applicants proposed principal design criteria that have been deemed acceptable by the NRC. Proposed principal design criteria are required per regulations in 10 CFR 50.34 and 10 CFR 52.79.

10 CFR 50.48(c), risk-informed, performance-based fire protection program requirements, which incorporates by reference National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, with certain exceptions.*

  • Although 10 CFR 50.48(c) is not applicable to non-LWRs, stationary micro reactors, and small modular LWRs, elements and concepts in NFPA 805 can be applied to these reactor types with justified exceptions and/or deviations, where appropriate.

Slide 57

Referenced Guidance 6

RG 1.189, Revision 4, Fire Protection for Nuclear Power Plants, which is a comprehensive fire protection guidance document, identifies the scope and depth of fire protection that the NRC staff would consider acceptable for nuclear power plants.

NFPA 804, Standard for Fire Protection for Advanced Light Water Reactor Electric Generating Plants.

Provides useful information when used in conjunction with NRC regulations and guidance - the NRC has not formally endorsed NFPA 804, and some of the information in the NFPA standard may conflict with regulatory requirements.

Provides a deterministic approach to the fire protection program.

NFPA 805, 2001 Edition.

NFPA 806, Performance-Based Standard for Fire Protection for Advanced Nuclear Reactor Electric Generating Plants Change Process.

Slide 58

Referenced Guidance (cont.)

7 RG 1.205, Revision 2, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, which provides NRC guidance on an acceptable approach to meeting 10 CFR 50.48(c).

Note that Nuclear Energy Institute (NEI) published implementing guidance for the specific provisions of NFPA 805 and 10 CFR 50.48(c) in NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c). RG 1.205, Revision 1, endorsed, with clarifications and exceptions, NEI 04-02, Revision 2. RG 1.205, Revision 2., updates the previous staff positions in RG 1.205, Revision 1, and endorses NEI 04-02, Revision 3.

Slide 59

Scope 8

The scope of this ISG addresses the review of the application content regarding the fire protection program for operations including application descriptions of:

Management policy and program direction and the responsibilities of those individuals responsible for the program/plans implementation.

The integrated combination of procedures and personnel that will implement fire protection program activities.

This ISG does not address fire protection system design or the fire hazards analysis within the probabilistic risk assessment (PRA).

Slide 60

Objectives 9

  • The fire protection program description should establish the fire protection policy for the protection of safety-significant* SSCs at each plant and the procedures, equipment, and personnel required to implement the program at the plant site.
  • 10 CFR 50.48 and Appendix A to 10 CFR Part 50 both refer to structures, systems, and components (SSCs) that are important-to-safety; in place of this term, this ISG refers to safety-significant SSCs to be consistent with terminology used in TICAP and ARCAP reference documents.

Slide 61

Objectives (cont.)

10

Describe the overall fire protection program for the facility; Identify the various positions within the licensee's organization that are responsible for the program; State the authorities that are delegated to each of these positions to implement those responsibilities; and Outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage.

Slide 62

Objectives (cont.)

11

  • The program description should also describe specific features necessary to implement the program such as:

Administrative controls and personnel requirements for fire prevention and manual fire suppression activities; and The means to limit fire damage to safety-significant SSCs so their capability to perform safety functions is maintained.

Slide 63

Guidance Topics 12 Applicable Industry Codes and Standards Organization, Staffing, and Responsibilities Fire Protection Staff Training and Qualification General Employee Training Fire Brigade Training and Qualification Chemical Fires (Training and Qualification Program related to the Unique Features of Chemical Fires associated with the Reactor Design)

Fire Protection Program Documentation and Changes, Configuration Control and Quality Assurance Verification and Validation (V&V) of Fire Models Slide 64

Guidance Topics (cont.)

13 Review of physical plant changes and procedure changes for impact on the fire protection program Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection hazard analysis Monitoring program Defined strategies for fighting fires Reporting Control of combustibles, hazardous materials, and ignition sources Housekeeping Manual firefighting capabilities Slide 65

Acceptance Criteria 14 The fire protection program description meets applicable Regulatory Positions of RG 1.189 or provides a basis for any deviations.

The program description provides clear delineations of organization, staffing, and responsibilities, consistent with RG 1.189.

The fire protection program change process identifies changes that require prior NRC review and approval, consistent with RG 1.189 and RG 1.205.

The monitoring program includes bases for failure probability assumptions used in the fire PRA, methods used to monitor availability, reliability, and performance of fire protection systems and features, and processes for identifying and implementing corrective actions (NFPA 805 and RG 1.205 are referenced).

Slide 66

Acceptance Criteria (cont.)

15 The V&V of fire models is consistent with RG 1.189 and RG 1.205.

The fire protection program QA requirements are consistent with RG 1.189.

The fire protection program description adequately addresses the evaluation of compensatory measures for interim use for adequacy and appropriate length of use.

Training of fire protection staff, fire brigade staff, and general employees regarding the fire protection program is consistent with the applicable sections of RG 1.189 and NFPA 805.

Note that NFPA 804 and NFPA 806 are not specifically referenced in the Acceptance Criteria, but the ISG acknowledges that the use of these or other referenced standards allows an applicant a way to describe what they have done without having to include a detailed description in their submittal.

Slide 67

16 QUESTIONS?

Slide 68

Break Meeting will resume at 1pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#

Advanced Reactor Stakeholder Public Meeting Slide 69

NRC-CNSC MOC September 29, 2021 Technology Inclusive and Risk-Informed Reviews for Advanced Reactors:

Comparing the US Licensing Modernization Project with the Canadian Regulatory Approach Chantal Morin, P. Eng Canadian Nuclear Safety Commission Eric Oesterle, Senior PM Nuclear Regulatory Commission Slide 70

  • Background on NRC-CNSC MOC
  • Key Messages
  • Report Overview and Background
  • Comparison observations and conclusions OUTLINE 2

Slide 71

  • On August 15, 2019, the Canadian Nuclear Safety Commission (CNSC) and the Nuclear Regulatory Commission (NRC), signed a Memorandum of Cooperation (MOC) to increase collaboration on the technical reviews of advanced reactor and small modular reactor technologies.
  • Under the MOC, the NRC and CNSC plan to issue joint reports that cover topics that are generally applicable to advanced reactor developers and designers, as well as targeted reports addressing specific technical aspects for individual vendors or potential applicants.

NRC-CNSC MOC - Background 3

Slide 72

  • The NRC and CNSC approved several work plans under the MOC which began collaborative efforts in mid-2020
  • Work plans included specific topical areas and technical topics associated with specific advanced reactor and small modular reactor vendors as well as work plan focused on process
  • Since the initiation of the MOC, the NRC and CNSC have completed a joint technical report associated with the X-energy design and a report comparing the technology-inclusive, risk-informed NRC and CNSC review approaches for advanced reactors NRC-CNSC MOC - Background 4

Public website links:

https://www.nrc.gov/reactors/new-reactors/nrc-cnsc-memorandum-of-cooperation.html https://nuclearsafety.gc.ca/eng/resources/news-room/feature-articles/Sharing-our-expertise-with-the-US-Nuclear-Regulatory-Commission.cfm Slide 73

  • Both the Canadian and U.S. regulatory approaches effectively protect the health and safety of the public and the environment during nuclear power plant operation.
  • Both the Canadian and U.S. regulatory approaches include some degree of technology-inclusive, risk-informed, performance-based review.
  • Comparing the Canadian and U.S. regulatory approaches is a first step in enhancing understanding of each regulators framework to enable greater cooperation and collaboration on advanced reactor designs that are contemplated for deployment in both countries.
  • The existing common ground between the Canadian and U.S. regulatory approaches can be leveraged to pursue joint technical reviews of advanced reactor applications in the future.

Key Messages 5

Slide 74

  • The recently completed report specifically focused on the Licensing Modernization Project (LMP) process and the CNSC review approach for Advanced Reactors.
  • LMP was a U.S. Department of Energy (DOE) funded and nuclear industry led effort to develop a technology-inclusive, risk-informed, and performance-based (TI-RIPB) approach to developing a safety case.
  • Elements of the LMP approach were developed over two decades and the completed process was finalized in NEI 18-04, Rev. 1, dated August 2019.

Next Generation New Plant (NGNP) table-top exercises with advanced reactor developers vetted and informed LMP process Several separate DOE reports formed basis for NEI 18-01, Rev. 1

Historical Overview of LMP 6

NRC public website link: https://www.nrc.gov/reactors/new-reactors/advanced/details.html#flexLearn Slide 75

  • Purpose
  • Overview of Regulatory Processes for New Designs Pre-application Interactions Application Interactions Overview of Regulatory Safety Objectives and Dose Limits Technology inclusive, risk-informed, performance-based licensing approaches
  • Assessment of Technology-Inclusive, Risk-Informed Approaches
  • Suggestions for Future Work
  • Conclusions REPORT - Overview 7

Public website links:

https://www.nrc.gov/reactors/new-reactors/nrc-cnsc-memorandum-of-cooperation/joint-reports-of-the-nrc-and-cnsc.html https://nuclearsafety.gc.ca/eng/resources/news-room/feature-articles/Sharing-our-expertise-with-the-US-Nuclear-Regulatory-Commission.cfm Slide 76

  • NRCs LMP approach is a technology-inclusive, risk-informed, and performance-based (TI-RIPB) approach for non-LWR advanced reactors
  • NRCs LMP approach is frequency-consequence oriented approach using PRA that can be used to establish the licensing basis and content of applications Background on NRC LMP Approach 8

NRC public website link: https://www.nrc.gov/reactors/new-reactors/advanced/details-1.html Slide 77

  • CNSC has a long history of risk-informed approach aligned with IAEA generally not very prescriptive
  • Regulatory Documents describe expectations for:

Postulated Initiating Events determination Safety Analysis: both deterministic and probabilistic Safety System Classification Defence in Depth, etc.

Background on CNSC Approach 9

CNSC public website link: http://www.nuclearsafety.gc.ca/eng/acts-and-regulations/regulatory-documents/index.cfm Slide 78

High level similarities: Technology-inclusive, risk-informed approaches

  • Safety analysis process includes both probabilistic and deterministic methods PRA* is foundational to the NRC LMP approach; PRA is also an important part of Canadian approach Multi-unit event sequences are addressed
  • Safety goals and objectives Demonstrate that frequently occurring plant events have minor consequences Demonstrate that events with severe potential consequences have a very low frequency of occurrence

Design Basis Events (DBEs)/Design Basis Accidents (DBAs)

Beyond Design Basis Events (BDBEs)/Beyond Design Basis Accidents (BDBAs)

Report - General Observations 10

- interchangeable Slide 79

High level similarities (continued):

  • Fundamental safety functions (3Cs: Control, Cool, Confine)

Report - General Observations 11 US NRC CNSC Reactivity and Power Control

Control of reactivity

Monitoring of safety-critical parameters to guide operator actions Heat Removal

Removal of heat from the fuel Radionuclide Retention

Confinement of radioactive material

Control of operational discharges and hazardous substances, as well as limitation of accidental releases

Shielding against radiation

  • Defense-in-Depth (DID) adequacy approach generally consistent with IAEA concept of layers of defense layers associated with design, programs, procedures, risk assessments no overreliance on a single layer of defense Slide 80

High level differences:

  • PRA and Risk Metrics CNSC uses Level 2 PRA results; NRC LMP approach uses Level 3 NRC LMP approach includes some cumulative risk metrics
  • Dose limits associated with DBEs/DBAs NRC - 25 rem (max)

CNSC - 2 rem Report - General Observations 12 Slide 81

High level differences:

  • Single failure criterion Embedded within NRC LMP DID assessment CNSC maintains use of SFC and includes some limited relaxation flexibility High level observation:
  • Safety classification of SSCs PSA metrics or DID importance CNSC accepts licensee proposals on safety classification Report - General Observations 13 NRC LMP CNSC (flexible to applicant proposals)

Safety Related Important to Safety Not Safety Related with Special Treatment Important to Safety Not Safety Related Not important to Safety Slide 82

Dose acceptance criteria and event classification Report - General Observations 14 Event sequency frequency (per plant year)

Slide 83

Dose acceptance criteria and event classification Report - General Observations 15 Event Category NRC LMP CNSC Frequency Ranges (per year)

Dose Limits (Rem)

Frequency Ranges (per year)

Dose Limits (Rem)

(Sv)

AOO

>10-2 Frequency 1-10-1/y: 0.1 to 1 Frequency 10-1-10-2/y: 1

>10-2 5mRem 0.5 mSv DBA/DBE 10 10-4 DBE: 1 - 25 DBA: 10 CFR 50.34 limit 25 Rem 10 10-5 2 Rem 20 mSv BDBA/BDBE 10 5x10-7 25 - 750

<10-5 (no lower limit defined)

No dose limit defined, Safety Goals applied Slide 84

  • The NRC LMP approach is more similar to the CNSCs approach than the traditional NRC licensing approach for LWRs.
  • Although consistent at a high level, NRCs LMP approach uses PRA more extensively while CNSCs approach considers other factors in decision making (operating experience, regulatory authority input, etc.).

Report - General Observations 16 Slide 85

  • Excellent exercise for both regulators to become aware of each others approaches for new reactor safety reviews
  • Benefits the facilitation of future cooperative efforts on joint technical reviews
  • Better general understanding at the CNSC of this new approach which could facilitate the conduct of some Vendor Design Reviews
  • Ongoing work plans under the CNSC-NRC Memorandum of Cooperation (MOC) involve joint reviews of limited technical topics for specific advanced reactor designs that will further develop our capabilities to perform joint technical reviews Report - General Conclusions 17 Slide 86

Questions?

THANK YOU 18 Slide 87

AOO: Anticipated Operational Occurrence BDBE: Beyond Design Basis Events (BDBA also)

DBA: Design Basis Accident (only safety related system are credited)

DBE: Design Basis Events (all systems are credited)

DEC: Design Extension Conditions (subset of BDBA)

DiD: Defence in Depth F-C: Frequency-Consequence LBE: Licensing Basis Events LMP: Licensing Modernization Project TI-RIPB: Technology-Inclusive Risk Informed Performance Based PIE: Postulated Initiating Event PSA/PRA: Probabilistic Safety Assessment/ Probabilistic Risk Assessment QHO: Quantitative Health Objectives SSC: Structures, Systems, Components ACRONYMS 19 Slide 88

Draft NUREG-2246, Fuel Qualification for Advanced Reactors Discussion of Comments Received during Public Comment Period (June 30-August 30, 2021)

September 29, 2021 Slide 89

  • Early Stakeholder Engagement

- Framework presented at May 7, 2020, periodic advanced reactor stakeholder meeting

- Draft white paper released in support of October 1, 2020, periodic advanced reactor stakeholder meeting

  • International coordination with the Nuclear Energy Agency (NEA) -

Working Group on the Safety of Advanced Reactors (WGSAR)

- Regulatory Perspectives on Nuclear Fuel Qualification for Advanced Reactors

  • Draft NUREG-2246 is an iteration of draft white paper, with adjustments to address feedback received at stakeholder meetings and NEA-WGSAR input

Background

Slide 90

  • NUREG-2246 provides a framework for evaluating a nuclear fuel design and attempts to enable a transparent, efficient, and thorough safety review
  • Informed by staff experience gained from licensing solid fuel reactor designs (particularly LWR designs), advanced reactor fuel testing performed to-date, and the accelerated fuel qualification (AFQ) considerations
  • Focused on areas where irradiated fuel tests have been required Purpose and Considerations Slide 91
  • An attempt was made to develop generically applicable criteria

- Some criteria may not be applicable to all fuel types or reactor designs

- Criteria may not be sufficient in some cases

  • Additional activity

- Separate guidance being developed for molten salt reactors

- NUREG-2246 is being exercised for a generic assessment of metal fuel (Idaho National Lab) and TRISO fuel (Pacific Northwest National Lab)

Purpose and Considerations Slide 92

  • Draft NUREG-2246, Fuel Qualification for Advanced Reactors

- FRN published on June 30, 2021

  • Three comment submittals received

- (1) Public, ML21243A353

- (2) Nuclear Energy Institute (NEI), ML21243A356

- (3) Public, ML21246A124

  • NEIs comment submittal requested a public meeting to further discuss the fuel qualification guidance NUREG-2246, Comment Submittal Received Slide 93

Status of Comment Review Submitter Comments Review Status Changes to NUREG-2246 1-Public 1

Complete None 2-NEI 30 Complete 17 comments resulted in edits to NUREG-2246, Revisions are still being developed 3-Public 8

5 of 8 Comments Reviewed Ongoing Slide 94

Outline of Highlighted Comments Comment Topic NEI-1, NEI-24, Public 3-1 Non-LWR examples NEI-5 Role of research literature NEI-7 Regulatory basis for fuel qualification NEI-9 Scope of degradation mechanisms NEI-10 GDC/ARDCs NEI-14 Fuel safety functions NEI-16 First core applications Comment Topic NEI-17 TRISO manufacturing NEI-18 Safe shutdown NEI-23 Coolable geometry NEI-26 Control element insertion NEI-29 Test envelope and Performance envelope NEI-30 Quality assurance Slide 95

There are numerous advanced reactor designs, with many different fuel designs, and each fuel design has its own unique challenges and advantages. This draft NUREG has a lot of guidance heavily based on existing LWR fuel designs, which may not be applicable to all advanced reactor designs. Please make the guidance more applicable to non-traditional fuel designs.

Please add examples for non-traditional LWR fuel designs.

Proposed Response Summary:

Staff acknowledges that NUREG-2246 is informed by lessons-learned from experience with LWR fuel but disagrees with the characterization that it is based on existing LWR fuel designs.

Section 1.3, Scope, states that assessment criteria draws from results from advanced reactor fuel testing performed to-date, and accelerated fuel qualification (AFQ) considerations On-going work to exercise NUREG-2246 framework through generic assessments of metal fuel (Idaho National Lab) and Tristructural Isotropic (TRISO) fuel (Pacific Northwest National Lab)

NEI Comment #1 and #24 and Public Comment 3-1, General Slide 96

The NUREG primarily references the Crawford report when establishing the objective of fuel qualification and its corresponding basis. A technical paper (the Crawford report) does not establish a regulatory basis and is not appropriate to establish the regulatory need for fuel qualification.

Furthermore, the noted economic operation in the Crawford report is not something that NRC should be regulating on.

Please replace the reference of a technical paper with a regulatory one (when identified as a means to do so) and remove reference to economics as a regulatory goal.

Proposed Response Summary:

Staff disagrees with the characterization that a technical paper served as a primary basis for NUREG-2246, but acknowledges that the research by Crawford, Porter, Hayes, Meyer, and Petti informed their judgement on an applicable definition of fuel qualification, for use in NUREG-2246.

Definition in Section 3 was selected because its an accurate characterization of fuel qualification based on the NRC regulations, staffs experience from licensing solid fuel reactor designs, results from advanced reactor fuel testing performed to-date, and AFQ considerations.

NEI Comment #5, Page 1-1 Slide 97

The text states "this framework relies on regulatory requirements that are applicable to applications for design certifications, combined licenses, manufacturing licenses, or standard design approvals."

However, fuel qualification itself is not required for any of the mentioned licensing approvals. Fuel qualification is therefore only necessary to meet the requirements in Section 2.1 to the extent that the fuel is specifically relied upon as a safety feature.

Please clarify the connection of fuel qualification to the noted licensing approvals, or modify the text to more clearly denote more context to how potentially applicable requirements would inform what is necessary for fuel qualification.

Proposed Response Summary:

The term fuel qualification is not explicitly defined or used in NRC regulations.

There are regulatory requirements generically applicable to power reactor applications that are generally associated with nuclear fuel behavior under conditions of normal operation, including the effect of anticipated operational occurrences, and accident conditions.

Regulatory requirements and connections to regulation is provided in Section 2.1, Regulatory Basis NEI Comment #7, Page 1-1 Slide 98

Fuel life-limiting failure and degradation mechanisms are not just due to irradiation during reactor operation. Other degradation mechanisms, chemical attacks, hydrogen pickup, high temperature, and time at temperature during AOOs or Design Basis Accidents also impact fuel performance.

Please remove "due to irradiation" to expand the applicability of the statement to include other failure mechanisms.

Proposed Response Summary:

Propose change from due to irradiation to due to irradiation and irradiation assisted phenomena.

Agree that there are degradation mechanisms beyond irradiation

- Primary obstacle to qualifying nuclear fuel has generally been demonstrating fuel performance at the desired exposure

- most fuel degradation phenomena are impacted by irradiation NEI Comment #9, Page 1-2 and Section 3.1.1 Slide 99

The regulatory basis denoted is 50.43(e) and the design criteria (GDC and ARDC). However, the GDC and ARDC are not requirements for, and as guidance are not required to be met by, non-LWRs. Thus, non-LWRs may choose to develop PDCs through another method. The guidance should clarify that fuel qualification is only necessary if it is determined to be one of the PDCs for the design, based upon the fuel being relied upon as a safety feature.

Please clarify the text to indicate how fuel qualification could be used to demonstrate compliance to 50.43 but is not necessary if fuel is not relied upon as a safety feature. Additional context on how potentially applicable requirements would inform what is necessary for fuel qualification would be helpful.

Proposed Response Summary:

Agree that GDCs and ARDCs are not necessarily requirements for non-LWRs but are instead considered guidance for non-LWR advanced reactor applicants in developing proposed PDCs.

The cited GDC/ARDC are associated with safety functions generally involving nuclear fuel that are not otherwise captured in NRC regulations (e.g., fuel safety limits, maintaining coolable geometry) and are expected to be addressed as part of fuel qualification.

Considering revisions to Section 2.1 to accurately reflect requirements associated with PDCs NEI Comment #10, Section 2.1 Slide 100

The following text is misleading because fuel may have safety functions as noted in the text, but it is not required to. It is possible to not credit the fuel and instead credit other mechanisms outside of the fuel matrix. Fuel qualification is therefore only necessary to meet the requirements in Section 2.1 to the extent that the fuel is specifically relied upon as a safety feature.

"Fuel qualification partially addresses the fundamental safety functions of control of reactivity, cooling of radioactive material, and confinement of radioactive material" Please clarify the role of fuel qualification and its necessity only if being relied upon and/or credited in the safety analysis as some designs may not. Additional context on how potentially applicable requirements would inform what is necessary for fuel qualification would be helpful.

Proposed Response Summary:

Staff recognizes that the role of fuel in the protection against the release of radioactivity can vary depending upon the reactor design.

Considering revising text to clarify that the role of fuel in the safety functions can be addressed in a graded manner in accordance with the degree to which the fuel is credited.

NEI Comment #14, Section 2.2.3 Slide 101

The text here on lead test specimen programs is applicable only if we had existing/operating advanced reactors. It does not discuss alternatives for fuel to be qualified for first core applications where lead test specimens are not possible.

Please add information on fuel qualification for first core applications.

Proposed Response Summary:

  • Considering adding a new Section 2.5, First Core Applications (and renumbering the current Section 2.5 to 2.6)

NEI Comment #16, Section 2.4 and Section 3.4.2 Slide 102

The TRISO SER allows for manufacturing independence as long as the final product has properties that fall within the specification range. Please note TRISO as an example of "insensitivity to manufacturing processes."

Please add text to denote that the TRISO is an example of fuel that has an insensitivity to manufacturing processes and instead measurable criteria can be used to justify predicted performance.

Proposed Response Summary:

Revise Sections 3.1 and 3.1.3 to reference the EPRI TRISO Topical Report and associated safety evaluation report, reflecting that key end-state parameters for TRISO particles have been identified that provide assurance of fuel performance during normal operation NEI Comment #17, Section 3.1.3 Slide 103

Criterion G2.3 includes a statement for the "ability to achieve and maintain safe shutdown can be assured." The NUREG further defines safe shutdown as "a state in which the reactor is subcritical, decay heat is being removed, and radionuclide inventory is contained." However, safe shutdown and safe state are not interchangeable. Not all reactors must be subcritical to be safe and therefore it is not necessary for all fuel types to be subcritical. Industry recommends aligning the NUREG with other documentation that use the phrase "a safe stable end-state" instead.

Please change criterion G2.3 to "Ability to achieve and maintain a safe, stable, end state" and revise the text throughout to be consistent with this revised criterion.

Proposed Response Summary:

Relatively recent NRC policy papers have clarified that maintaining subcriticality with only safety-related structures, systems, and components may not be required (SECY-18-0099)

Staff expects that nuclear fuel be designed such that forces on the nuclear fuel, resulting from internal or external events, will not preclude the eventual achievement of a subcritical state NEI Comment #18, Figures 3-3 and 3-8, Section 3.2.3 Slide 104

No specific criterion is set as to what the term "coolable geometry" means for non-traditional LWR fuel designs. For example, some advanced reactor fuel designs like the MSFR and the MCFR are planning on using liquid salt with fissile product as both their coolant and their fuel, which would require clarification/flexibility on the definition of "coolable geometry."

Please clarify what a coolable geometry criterion would be for fuel designs that do not have a containment (e.g., LWR fuel cladding) and how this will ensure ability to attain safe, stable, end state.

Proposed Response Summary:

Section 3.2.3.1.1, G2.3.1(a) - Identification of Phenomena provides examples of the types of phenomena that could cause a loss of coolable geometry.

Staff recognizes that these criteria cannot be specified generically for all fuel designs Section 1.3, Scope, clarifies that some criteria may not apply to liquid fuel forms (e.g., MSR fuel), and these fuel forms may require additional or alternative criteria.

NEI Comment #23, Page 3-8 Slide 105

For designs that do not have neutron control element insertion (e.g., control drums),

having criteria to ensure control element insertion paths would not be necessary.

Please revise text to indicate criteria should be specified only for designs with neutron control elements whose insertion is credited in accident response models.

Proposed Response Summary:

Revise Section 3 to reference Section 1.3, Scope

- The assessment criteria draws on regulatory experience gained from licensing solid fuel reactor designs (particularly LWR designs), results from advanced reactor fuel testing performed to-date, and AFQ considerations.

- An attempt has been made to develop generically applicable criteria.

- Staff recognizes that some criteria may not be applicable to all fuel types or reactor designs.

NEI Comment #26, Page 3-10 Slide 106

The text does not address methods for justifying when fuel can be used beyond its performance envelope when lead test specimens are not available.

Please add information on what adequate justification is needed to expand the performance envelope using experimental data without the use of lead test specimens.

Proposed Response Summary:

Reference Section 3.3.2.2.4, EM G2.2.4-Restricted Domain, Application of an evaluation model outside of the supporting test envelope (see Section3.4.2) may be justified based on physical arguments (e.g., that the evaluation model provides a simplified or bounding treatment of physical phenomena). Justification for extrapolation of a model outside of the test envelope is strengthened by the use of physics-based models, such as those discussed in Section 2.3, which are informed by fundamental information about fuel evolution and behavior, as opposed to empirically derived models (Terrani, et al., 2020).

NEI Comment #29, Page 3-16 Slide 107

ASME NQA-1 is not the only way to qualify fuel data. For data collected at national laboratories, the application of this standard may not be possible, despite the national laboratories using alternative and acceptable quality assurance methods. NRC should accept data from the technical experts at national laboratories if the data was collected under the lab's QA program as noted in ML20054A297, where NRC staff determined that Argonne National Lab's quality assurance program plan is based on the method provided in ASMEs NQA-1-2008/2009 and satisfies the quality assurance requirements of Appendix B to 10 CFR Part 50.

Please either remove or modify the text to allow for data to also be qualified under the commercial grade dedication (CGD) process rather than stating data must be made compliant.

Proposed Response Summary:

Revise Section 3.4.3.1 to clarify that approaches other than those provide in ASMEs NQA-1, including CGD, may be acceptable means for justifying that data is collected under an appropriate QA program NEI Comment #30, Page 3-17 Slide 108

  • Continue to develop formal responses to public comment submittals
  • Additional public meeting(s), if needed

- Tentatively scheduled for November 2021

- Expected to be published in January/February 2022 Stakeholder engagement on on-going work to exercise NUREG-2246 frameworkmetal fuel and TRISO fuel Next Steps Slide 109

Break Meeting will resume in 10 minutes Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#

Advanced Reactor Stakeholder Public Meeting Slide 110

Scalable Human Factors Engineering Review Framework for Advanced Reactors NRR - Division of Advanced Reactors and Non-power Production and Utilization Facilities NRR - Division of Reactor Oversight September 29, 2021 Slide 111

Agenda

  • Background
  • Advanced Reactor Implications for Human-System Integration
  • Scalable HFE Framework
  • Next Steps
  • Questions/Comments 2

Slide 112

10 CFR 50.34(f)(2)(iii)

Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts. (I.D.1) 3 Current Framework Slide 113

NUREG-0711, Human Factors Engineering Program Review Model 4

Current Framework HFE Program Management Operating Experience Review Function Analysis &

Allocation Task Analysis Staffing &

Qualification Human Reliability Analysis Human -System Interface Design Procedure Development Training Program Development Human Factors Verification and Validation Design Implementation Human Performance Monitoring Planning and Analysis Design Verification and Validation Implementation and Operation Acceptance Criteria: 300+,

developed for LLWRs, integrated with review process Slide 114

White Paper: Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors*

  • Presented to advanced reactor stakeholders on April 8, 2021
  • Presented as Key Guidance to ACRS S/C on May 20, 2021

5 Background

Slide 115

White Paper: Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors*

Key messages:

  • If an advanced reactor design presents very low radiological risk then the current regulatory framework for operation of large LWRs may be unnecessary for reasonable assurance of safety.
  • A new regulatory framework for advanced reactors (10 CFR 53) should be capable of addressing novel operational concepts for a wide variety of advanced reactor technologies.
  • A risk-informed, performance-based, and technology-inclusive regulatory framework for advanced reactors must appropriately consider the role of humans and human-system integration

6 Background

Slide 116

White Paper: Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors*

Topics included:

  • diverse and novel operational characteristics, including automation of operations
  • staffing and qualifications of operations personnel
  • evolution in control room concepts

7 Background

Slide 117

  • Will be used to varying degrees and in different combinations among advanced reactor designs
  • Achieve safety through different means
  • Have different implications for the role of personnel in the assurance of plant safety 8

Passive Safety Features, Inherent Safety Characteristics & Automation Implications for human-system integration Slide 118

  • Automation is implemented in levels that span from manual to autonomous operation.
  • Even in an autonomous design, there may still exist a need for humans to implement manual operations under certain circumstances, such as for defense-in-depth.
  • Automation generally enhances operational performance, however other operational effects must be considered as well (e.g., operators losing manual control proficiency).

9 Automation of Plant Operations Implications for human-system integration Slide 119

  • Applicants may propose alternative staffing models.
  • Number of personnel
  • Crew composition
  • Span of control / responsibility
  • Operator licensing for advanced reactors may include
  • allowances for varying licensing examination scope on a facility-specific basis
  • modified simulator requirements.

10 Staffing and Qualifications Implications for human-system integration Slide 120

  • Some advanced reactor facilities may wish to not utilize traditional control rooms
  • Operations functions may become decentralized
  • New missions for reactor facilities may be bring the emergence of functions that have no precedent within traditional control rooms 11 The Evolving Concept of the Control Room Implications for human-system integration Slide 121

Advanced reactors are likely to vary in terms of the

  • Role
  • Number
  • Qualifications
  • Crew composition
  • Work settings of personnel responsible for plant operational safety 12 Summary Implications for human-system integration Slide 122

13 Part 53 working groups addressing human-system considerations Tailored Operator Licensing Scalable Human Factors Engineering Reviews Flexible Staffing Slide 123

  • Brookhaven National Laboratory contracted to develop a method for scaling the scope and depth of HFE reviews for advanced reactors (ML21266A192)
  • Review of advanced reactor design concepts
  • Identification of human performance considerations
  • Review of NRCs existing review plans / methods for non-LLWR nuclear technologies
  • Method enables staff to readily adjust the focus and level of HFE review based upon risk/safety insights and the unique characteristics of the facility design/operation
  • Staff developing guidance for scalable HFE review using BNL technical basis report 14 Technical Basis Development:

Scalable Human Factors Engineering Reviews Slide 124

15 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 125

  • Objective: Provide a standard process for the development of an application-specific, risk-informed HFE review plan that is focused on those aspects of a design and its operation most important to the assurance of facility operational safety
  • Relationship to other HFE guidance:
  • Does not replace existing NRC guidance (e.g., NUREGs 0711 and 0700)
  • Facilitates selective use of existing guidance and standards (e.g., IEEE, ANS) as justified by state-of-the-art 16 Staff Guidance :

Scalable Human Factors Engineering Reviews Slide 126

17 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 127

  • Objective: Develop an understanding of a facility design and its operation from the perspective of their implications for human performance and human factors engineering
  • Process: Staff review of applicant submittals, beginning pre-application when possible, to identify safety and risk insights and those features/characteristics of the design and its operation that will be important to informing the scope and depth of the HFE review
  • Inputs:
  • Facility concept of operations
  • General safety and risk insights for the facility
  • Important human actions (from risk and deterministic analyses)
  • Human factors engineering activities conducted/planned

Scalable Human Factors Engineering Reviews Slide 128

Concept of operations: A concept of operations (ConOps) defines the goals and expectations for the new system from the perspective of users and other stakeholders and defines the high-level considerations to address as the detailed design evolves. An HFE-focused ConOps addresses the following six dimensions:

  • Plant Goals (or Missions)
  • Agents Roles and Responsibilities
  • Staffing, Qualifications, and Training
  • Management of Normal Operations
  • Management of Off-normal Conditions and Emergencies
  • Management of Maintenance and Modifications 19 Characterization :

Scalable Human Factors Engineering Reviews Slide 129

20 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 130

  • Objective: Identify aspects of the applicants facility design and operations that warrant staff HFE review
  • Process: Review the results of the characterization considering
  • risk importance
  • safety significance
  • uncertainty (e.g., potential unknowns introduced by novelty, lack of detail)
  • Inputs: Characteristics of the facility design and operation 21 Targeting:

Scalable Human Factors Engineering Reviews Slide 131

  • Example characteristics:
  • New missions
  • New staffing positions
  • Safety function monitoring
  • Risk-important human actions
  • New hazards
  • High levels of automation 22 Targeting:

Scalable Human Factors Engineering Reviews Slide 132

23 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 133

  • Objective: Identify the applicants HFE activities that warrant staff review
  • Process: Review the results of the characterization and targeting considering
  • HFE activities completed / planned by the licensee
  • Characteristics identified for targeting Assess HFE activities needed to establish technical basis for the use / design of targeted characteristics
  • Inputs:
  • Characteristics of the design and operation targeted for review
  • HFE activities completed / planned by the licensee 24 Screening:

Scalable Human Factors Engineering Reviews Slide 134

  • Example:

25 Screening:

Scalable Human Factors Engineering Reviews Example Targeted Characteristic HFE Activity Selected for Review*

1 Novel technology Operating experience review 2

New staffing model Staffing plan validation 3

New job function Task analysis 4

New mission Integrated system validation

  • Selected activities are for explanatory purposes only Slide 135

26 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 136

H FE standards and guidelines cited by the applicant 27 Grading:

Scalable Human Factors Engineering Reviews Slide 137

  • Example:

28 Scalable Human Factors Engineering Reviews Example Characteristic / HFE Activity Standard / Guideline Cited by Applicant Standard / Guideline Selected by staff*

1 Alarm system NUREG-0700 based style manual NUREG-0700 2

Task Analysis None NUREG-0711, Chapter 5 3

Staffing NUREG-1791 NUREG-1791 as amended for advanced reactors 4

Computer-based procedures IEEE-1023 IEEE-1786

  • Selected standards and guidelines are for explanatory purposes only Slide 138

29 Staff guidance - Process Overview:

Scalable Human Factors Engineering Reviews Characterization Targeting Screening Grading Assemble Review Plan Slide 139

  • Objective: Develop an application-specific plan for an HFE technical review of the applicants submittal.
  • Process: Staff will review results of the targeting, screening and grading activities in conjunction with the applicants schedule for subsequent HFE activities and submittals to develop an integrated plan that optimizes the technical review, use of resources, and schedule for completion.
  • Inputs:
  • Characteristics of the design and operation targeted for staff review
  • Applicants HFE activities selected by for staff review
  • HFE standards and guidelines to be applied to the review
  • Applicants schedule for the conduct of HFE activities and related submittals 30 Assemble the review plan:

Scalable Human Factors Engineering Reviews Slide 140

  • Encourages and facilitates early engagement between applicants and the staff
  • Tailors the HFE technical review to the specific application using risk and safety insights
  • Establishes a standard, technology inclusive process for scaling a review that does not presume a given form or scope of applicant HFE activities /

program

  • Enables the application of standards, guidelines, and criteria according to their fit for the facility, technology, method, or process under review 31

SUMMARY

Scalable Human Factors Engineering Reviews Slide 141

32 Guidance Development:

Next Steps Activity Target Date Issue summary report for technical basis development October 15 Complete draft guidance for scaling HFE technical review plans November 30 Issue draft guidance for stakeholder review and feedback December 15 Slide 142

33 Questions/Comments Slide 143

Endorsement of ASME Section XI, Division 2, Reliability and Integrity Management (RIM) Programs for Non-Light Water Reactors September 29, 2021 Steve Philpott, Project Manager Tim Lupold, Senior Mechanical Engineer Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Slide 144

Objectives for Todays Presentation Alert Stakeholders that DG-1383 is publicly available for comment Overview the staffs review of ASME Code,Section XI, Division 2 Familiarize stakeholders with ASME Code,Section XI, Division 2 (RIM)

Present the unique regulatory aspects of preservice and inservice inspection (PSI / ISI) programs for advanced reactors Identify the conditions the staff has recommended in the draft guide on the use of ASME Code,Section XI, Division 2 Communicate plans for public comment period and final regulatory guide 2

Slide 145

Review Team 3

RES/DE

NRR/DEX

  • Hanh Phan - Senior Reliability and Risk Analysis Analyst - NRR/DANU
  • Steve Philpott - Project Manager -

NRR/DANU

  • Tim Lupold - Senior Mechanical Engineer - NRR/DANU
  • Steve Downey - Senior Reactor Inspector - R-II
  • Isaac Anchondo-Lopez - Reactor Inspector - R-IV The RIM Working Group And other SMEs as needed Slide 146

RIM Review Project Schedule

  • Responded to ASME letter and initiated detailed staff review: Aug 2020
  • Reviewed code and developed initial staff positions: Aug - Dec 2020
  • Developed staff positions and draft regulatory guide (RG): Jan - Sep 2021
  • Published draft RG: Sep 2021 4

Slide 147

Regulatory Guide Project Schedule 5

Federal Register announcement of availability for public comment 45-day public comment period Public comment resolution Issue as Final Regulatory Guide shortly after next ASME Code Edition Rulemaking

  • Goal is by June 30, 2022 Slide 148

ASME Section XI, Division 2 (RIM) -

Introduction RIM is a program to ensure the reliability and integrity of components.

Based on achieving an acceptable level of reliability. Establishes reliability targets from the PRA.

RIM expert panels ensure that components can be managed through RIM strategies to ensure that reliability targets will be met.

Allows flexibility for Owners to implement alternative strategies from Section XI, Division 1 requirements.

Technology neutral - applicable to all reactor designs Technology-specific appendices 6

Slide 149

ASME Section XI, Division 2 (RIM) -

Introduction

  • Step 1: Determine Scope of SSCs for RIM Program
  • Step 2: Evaluate SSC Damage Mechanisms
  • Step 3: Determine Plant and SSC Level Reliability and Capability Requirements
  • Step 4: Identify and Evaluate RIM Strategies to Achieve Reliability Targets
  • Step 5: Evaluate Uncertainties in Reliability Performance
  • Step 6: Implement RIM Program
  • Step 7: Monitor SSC Reliability Performance and Update RIM Program 7

Slide 150

RG 1.246 (DG-1383) Structure

  • Purpose
  • Applicability (non-LWRs)
  • Regulations &

Related Guidance 1

  • Background
  • Bases for NRC Staff Positions 2
  • Staff Regulatory Guidance (Conditions) 3 8

Section A Section B Section C Slide 151

Regulatory Aspects 9

Use of license condition

  • Proposed regulatory tool Difference between LWRs and non-LWRs
  • LWRs: 50.55a governs
  • Non-LWRs: no clear regulation
  • Part 50 Appendix A is guidance for developing PDC for non-LWRs
  • plans for conducting normal operations, including maintenance, surveillance, and periodic testing of SSCs Slide 152

DG-1383 Conditions Use in conjunction with a license condition Use with 2019 Edition of Section XI-Division 1 Document how aspects of Section XI-Division 2 are considered Changes / Information to be sent to the NRC

  • For review and approval
  • For information
  • Other information to be available for audit ANDE-1 not approved for personnel Qualification 10
  • Information to be submitted with the application Slide 153

DG-1383 Conditions - Cont.

Information within Code that is in course of preparation is to be developed and submitted for review and approval as appropriate by the applicant Editions of NQA-1 and RA-S-1.4

  • RG 1.28 and Trial Use RG 1.247, respectively Justify acceptability of the PRA in RIM program Cannot override construction code NDE without approval Pre-Service Inspections not addressed in Code / Encoded examination requirement limited to volumetric exams 11 Slide 154

DG-1383 Conditions - Cont.

Appendix V to be considered for low pressure applications Records retention to be IAW QA program requirements Stress relaxation to be considered as a degradation mechanism Minor errata type corrections Liquid leak test clarifications and hold time limits 12 Slide 155

Next Steps

  • DG-1383 public comment period until mid-November (45 days)
  • Resolution of comments
  • Final RG issued: targeting mid-2022 (ASME Code edition final rule)
  • NRC Reviewer and Inspection Guidance

- Develop plan and vet through management 13 Slide 156

Questions or Comments?

Slide 157

Future Meeting Planning

  • The next periodic stakeholder meeting is scheduled for November 10, 2021.
  • If you have suggested topics, please reach out to Prosanta.Chowdhury@nrc.gov.

Slide 158

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