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{{#Wiki_filter:Advanced Reactor Stakeholder Public Meeting September 29, 2021 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#
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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,          INL Regulatory Research Planning for Micro-Reactor Development 11:10 - 11:40 am          Summary of Technical Reports on Transportation and Storage of            NRC Advanced Reactor Fuel 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  NRC and CNSC Inclusive and Risk-Informed Reviews for Advanced Reactors 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      NRC Staffing for Advanced Reactors 4:10 - 4:55 pm        Draft Regulatory Guide to Endorse ASME Section XI, Division 2, Reliability    NRC and Integrity Management (RIM) Programs for Non-Light Water Reactors 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
* Possession and Transportation of Special Nuclear Material
* 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
 
Identify external hazards (EXH)      Flowchart for an acceptable site hazard parameters in region of Micro-Rx assessment in the micro-reactor design process Site Investigation to characterize      Identify Micro-Rx vendors EXH EXH site (EXH-C)                        postulated site parameters (EXH-P)
Compare EXH-C with EXH-P EXH-C within EXH-P?                    End assessment; Y
document results N
Design reanalyzed and EXH-C                        End assessment; Y                                        Key:
within modified EXH-P?                              document results Y = yes N = no N                                                    U = unacceptable performance A = acceptable performance Evaluate mitigating actions                  End assessment; A
document results U
Site not suitable for deployment of this Micro-Rx Slide 14
 
External Hazards and Siting
* 10 CFR Part 100.21 & 100.23 requires characterization of several potential hazards
  - 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
* Regulatory Guide 4.2 Rev. 3 (ML18071A400)
* Interim Staff Guidance ISG-029 (ML20252A076)
  - Generic Environmental Impact Statement
* Potential Change to Framework
  - Enhancing Review Flexibility
* Proposed Update to Part 51 (SECY-21-0001)
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
* Cyber Security
* 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
* Pre-application meetings are crucial Slide 38
 
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
* TRISO: Tristructural isotropic
* LWR: Light water reactor Slide 52
 
Stakeholders Meeting Advanced Reactor Content of Application Project (ARCAP)
Risk-informed, Performance-Based Fire Protection Program (for Operations)
Interim Staff Guidance (Draft) 1 Slide 53
 
===Background===
* 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                          2
 
ARCAP and Technology Inclusive Content of Application Project (TICAP) - Nexus Outline Safety Analysis Report (SAR) -                                                                            Additional Portions of Application Based on TICAP Guidance
* Technical Specifications
: 1. General Plant Information, Site
* Technical Requirements Manual Description, and Overview of the Safety
* Quality Assurance Plan (design)
Case
* Fire Protection Program (design)
: 2. Methodologies and Analyses
* Quality Assurance Plan
: 3. Licensing Basis Event (LBE) Analysis                                                                          (construction and operations)
: 4. Integrated Evaluations                                    Audit/inspection of Applicant Records
* Emergency Plan
: 5. Safety Functions, Design Criteria, and
* Calculations
* Physical Security Plan SSC Safety Clasification
* Analyses
* SNM physical protection program
: 6. Safety Related SSC Criteria and
* P&IDs
* SNM material control and Capabilities
* System Descriptions                              accounting plan
: 7. Non-safety related with special treatment
* Design Drawings
* Cyber Security Plan SSC Criteria and Capabilities
* Design Specs
* Fire Protection Program
: 8. Plant Programs
* Procurement Specs                                (operational)
* Probabilistic Risk Assessment
* Radiation Protection Program Additional SAR Content -Outside the Scope
* Offsite Dose Calculation Manual of TICAP
* Inservice inspection/Inservice
: 9. Control of Routine Plant Radioactive                                                                            testing (ISI/IST) Program Effluents, Plant Contamination, and Solid
* Environmental Report Waste
* Site Redress Plan
: 10. Control of Occupational Doses
* Exemptions, Departures, and
: 11. Organization                                                                                                  Variances
: 12. Initial Startup Programs
* Facility Safety Program (under consideration for Part 53
* Safety Analysis Report (SAR) structure based on clean                                                        applications) sheet approach
*Additional contents of application outside of SAR are still under discussion. The above list is draft and for illustration purposes only.          3 Slide 55
 
Background (cont.)
* 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                                4
 
Requirements
* 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                                    5
 
Referenced Guidance
* 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                                  6
 
Referenced Guidance (cont.)
* 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                            7
 
Scope
* 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                            8
 
Objectives
* 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                                9
 
Objectives (cont.)
* The fire protection program description should:
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                          10
 
Objectives (cont.)
* 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                        11
 
Guidance Topics
* 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                          12
 
Guidance Topics (cont.)
* 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                            13
 
Acceptance Criteria
* 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                              14
 
Acceptance Criteria (cont.)
* 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                            15
 
QUESTIONS?
Slide 68 16
 
Advanced Reactor Stakeholder Public Meeting Break Meeting will resume at 1pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#
Slide 69
 
NRC-CNSC MOC Technology Inclusive and Risk-Informed Reviews for Chantal Morin, P. Eng Canadian Nuclear Safety Commission Advanced Reactors:
Eric Oesterle, Senior PM Nuclear Regulatory Commission                  Comparing the US Licensing Modernization Project with the Canadian Regulatory Approach September 29, 2021 Slide 70
 
OUTLINE
* Background on NRC-CNSC MOC
* Key Messages
* Report Overview and Background
* Comparison observations and conclusions 2
Slide 71
 
NRC-CNSC MOC - Background
* 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.
3 Slide 72
 
NRC-CNSC MOC - Background
* 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 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 4
Slide 73
 
Key Messages
* 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.
5 Slide 74
 
Historical Overview of LMP
* 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
* NEI 18-04 was endorsed in Regulatory Guide (RG) 1.233 in June 2020.
NRC public website link: https://www.nrc.gov/reactors/new-reactors/advanced/details.html#flexLearn 6 Slide 75
 
REPORT - Overview
* 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 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 7 Slide 76
 
Background on NRC LMP Approach
* 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 NRC public website link: https://www.nrc.gov/reactors/new-reactors/advanced/details-1.html 8 Slide 77
 
Background on CNSC Approach
* 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.
CNSC public website link: http://www.nuclearsafety.gc.ca/eng/acts-and-regulations/regulatory-documents/index.cfm 9 Slide 78
 
Report - General Observations 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
                                                    *NOTE: Probabilistic Risk Assessment (PRA) is the US Multi-unit event sequences are addressed      terminology for Probabilistic Safety Assessment (PSA)
                                                    - interchangeable
* 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
* Licensing basis events/postulated initiating events Abnormal Operation Occurrences (AOOs)
Design Basis Events (DBEs)/Design Basis Accidents (DBAs)
Beyond Design Basis Events (BDBEs)/Beyond Design Basis Accidents (BDBAs) 10 Slide 79
 
Report - General Observations High level similarities (continued):
* Fundamental safety functions (3Cs: Control, Cool, Confine)
US NRC                                                      CNSC
* Control of reactivity Reactivity and Power Control
* Monitoring of safety-critical parameters to guide operator actions Heat Removal
* Removal of heat from the fuel
* Confinement of radioactive material Radionuclide Retention
* 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                                                                    11 Slide 80
 
Report - General Observations 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
* Threshold levels and definitions for LBEs DBE vs. DBA BDBE vs. BDBA
* Dose limits associated with DBEs/DBAs NRC - 25 rem (max)
CNSC - 2 rem 12 Slide 81
 
Report - General Observations High level differences:
* Single failure criterion Embedded within NRC LMP DID assessment                                CNSC (flexible to CNSC maintains use of SFC and includes            NRC LMP applicant proposals) some limited relaxation flexibility Safety Related      Important to Safety High level observation:
* Safety classification of SSCs                    Not Safety Related with Important to Safety PSA metrics or DID importance                  Special Treatment CNSC accepts licensee proposals on safety classification                                                          Not important Not Safety Related to Safety 13 Slide 82
 
Report - General Observations Event sequency frequency (per plant year)
Dose acceptance criteria and event classification 14 Slide 83
 
Report - General Observations Dose acceptance criteria and event classification NRC LMP                                          CNSC Event                                                                                Dose Limits Frequency Ranges            Dose Limits          Frequency Ranges Category      (per year)                  (Rem)                (per year)
(Rem)
(Sv)
Frequency 1-10-1/y: 0.1 to 1                            50mRem AOO            >10-2 Frequency 10-1-10-2/y: 1
                                                                      >10-2 0.5 mSv DBE: 1 - 25 2 Rem DBA/DBE        10 10-4        DBA: 10 CFR 50.34 limit        10 10-5 20 mSv 25 Rem
                                                                      <10-5 No dose limit defined, BDBA/BDBE      10 5x10-7 25 - 750            (no lower limit Safety Goals applied defined) 15 Slide 84
 
Report - General Observations
* 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.).
16 Slide 85
 
Report - General Conclusions
* 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 17 Slide 86
 
THANK YOU Questions?
18 Slide 87
 
ACRONYMS 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 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
 
===Background===
* 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 Slide 90
 
Purpose and Considerations
* 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 Slide 91
 
Purpose and Considerations
* 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)
Slide 92
 
NUREG-2246, Comment Submittal Received
* 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 Slide 93
 
Status of Comment Review Submitter    Comments          Review Status Changes to NUREG-2246 1-Public        1              Complete              None 17 comments resulted in edits to NUREG-2246, 2-NEI          30              Complete Revisions are still being developed 5 of 8 Comments 3-Public        8                                    Ongoing Reviewed Slide 94
 
Outline of Highlighted Comments Comment        Topic                                Comment Topic NEI-1, NEI-24, Non-LWR examples                    NEI-17  TRISO manufacturing Public 3-1 NEI-5          Role of research literature          NEI-18  Safe shutdown Regulatory basis for fuel NEI-7 qualification                        NEI-23  Coolable geometry Scope of degradation NEI-9 mechanisms                          NEI-26  Control element insertion GDC/ARDCs NEI-10                                                      Test envelope and NEI-29 Performance envelope NEI-14          Fuel safety functions NEI-16          First core applications              NEI-30  Quality assurance Slide 95
 
NEI Comment #1 and #24 and Public Comment 3-1, General 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)
Slide 96
 
NEI Comment #5, Page 1-1 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.
Slide 97
 
NEI Comment #7, Page 1-1 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 Slide 98
 
NEI Comment #9, Page 1-2 and Section 3.1.1 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 Slide 99
 
NEI Comment #10, Section 2.1 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 Slide 100
 
NEI Comment #14, Section 2.2.3 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.
Slide 101
 
NEI Comment #16, Section 2.4 and Section 3.4.2 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)
Slide 102
 
NEI Comment #17, Section 3.1.3 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 Slide 103
 
NEI Comment #18, Figures 3-3 and 3-8, Section 3.2.3 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 Slide 104
 
NEI Comment #23, Page 3-8 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.
Slide 105
 
NEI Comment #26, Page 3-10 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.
Slide 106
 
NEI Comment #29, Page 3-16 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 Section 3.4.2) 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).
Slide 107
 
NEI Comment #30, Page 3-17 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 Slide 108
 
Next Steps
* Continue to develop formal responses to public comment submittals
* Additional public meeting(s), if needed
* Advisory Committee on Reactor Safeguards (ACRS) Full Committee Meeting
  - Tentatively scheduled for November 2021
* Issue final NUREG-2246
  - Expected to be published in January/February 2022
* Stakeholder engagement on on-going work to exercise NUREG-2246 frameworkmetal fuel and TRISO fuel Slide 109
 
Advanced Reactor Stakeholder Public Meeting Break Meeting will resume in 10 minutes Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 970 740 540#
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
 
Current Framework 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 Slide 113
 
Current Framework NUREG-0711, Human Factors Engineering Program Review Model Planning and        Design            Verification    Implementation Analysis                            and Validation    and Operation HFE Program Management Acceptance Criteria: 300+,
Operating                                developed for LLWRs, integrated Experience Review with review process Human -System Interface Design Function                                                  Design Analysis &                                              Implementation Human Factors Allocation      Procedure          Verification and Development Validation Human Task Analysis                                            Performance Monitoring Training Program Development Staffing &
Qualification Human Reliability Analysis 4
Slide 114
 
===
Background===
White Paper: Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors*
* Released March 25, 2021 (ML21069A003)
* Presented to advanced reactor stakeholders on April 8, 2021
* Presented as Key Guidance to ACRS S/C on May 20, 2021 5
Slide 115
 
===
Background===
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
Slide 116
 
===
Background===
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
Slide 117
 
Implications for human-system integration Passive Safety Features, Inherent Safety Characteristics & Automation
* 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
Slide 118
 
Implications for human-system integration Automation of Plant Operations
* 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 Slide 119
 
Implications for human-system integration Staffing and Qualifications
* 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 Slide 120
 
Implications for human-system integration The Evolving Concept of the Control Room
* 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 Slide 121
 
Implications for human-system integration Summary 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 Slide 122
 
Part 53 working groups addressing human-system considerations Tailored Flexible Operator Staffing Licensing Scalable Human Factors Engineering Reviews 13 Slide 123
 
Scalable Human Factors Engineering Reviews Technical Basis Development:
* 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 Slide 124
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Characterization Targeting  Screening    Grading  Review Plan 15 Slide 125
 
Scalable Human Factors Engineering Reviews Staff Guidance :
* 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 Slide 126
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Characterization  Targeting  Screening Grading Review Plan 17 Slide 127
 
Scalable Human Factors Engineering Reviews Characterization :
* 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
* Compliance with CFR requirements / planned exemption requests 18 Slide 128
 
Scalable Human Factors Engineering Reviews Characterization :
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 Slide 129
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Review Characterization Targeting  Screening    Grading Plan 20 Slide 130
 
Scalable Human Factors Engineering Reviews Targeting:
* 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 Slide 131
 
Scalable Human Factors Engineering Reviews Targeting:
* Example characteristics:
* New missions
* New staffing positions
* Safety function monitoring
* Risk-important human actions
* New hazards
* High levels of automation 22 Slide 132
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Review Characterization Targeting Screening    Grading Plan 23 Slide 133
 
Scalable Human Factors Engineering Reviews Screening:
* 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 Slide 134
 
Scalable Human Factors Engineering Reviews Screening:
* Example:
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 25 Slide 135
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Review Characterization Targeting  Screening    Grading      Plan 26 Slide 136
 
Scalable Human Factors Engineering Reviews Grading:
* Objective: Identify the standards and guidelines to be applied to the review of the applicants submittals
* Process: Review the applicants submittals to identify the HFE standards that the applicant has applied to their design, methods, and processes and assess the adequacy of the identified standard for their intended purpose. Establish the standards and guidelines to be applied in the staff HFE review.
* Inputs:
* Characteristics of the design and operation targeted for staff review
* Applicants HFE activities selected for staff review
* HFE standards and guidelines cited by the applicant 27 Slide 137
 
Scalable Human Factors Engineering Reviews Grading:
* Example:
Example  Characteristic / HFE Standard / Guideline Cited by Standard / Guideline Selected Activity            Applicant                    by staff*
1    Alarm system        NUREG-0700 based style        NUREG-0700 manual 2    Task Analysis        None                          NUREG-0711, Chapter 5 3    Staffing            NUREG-1791                    NUREG-1791 as amended for advanced reactors 4    Computer-based      IEEE-1023                    IEEE-1786 procedures
* Selected standards and guidelines are for explanatory purposes only 28 Slide 138
 
Scalable Human Factors Engineering Reviews Staff guidance - Process Overview:
Assemble Characterization Targeting  Screening    Grading  Review Plan 29 Slide 139
 
Scalable Human Factors Engineering Reviews Assemble the review plan:
* 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 Slide 140
 
Scalable Human Factors Engineering Reviews
 
==SUMMARY==
* 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 Slide 141
 
Next Steps Guidance Development:
Activity              Target Date Issue summary report for technical October 15 basis development Complete draft guidance for scaling November 30 HFE technical review plans Issue draft guidance for            December 15 stakeholder review and feedback 32 Slide 142
 
Questions/Comments 33 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 Slide 145                            2
 
The RIM Working Group
* Bruce Lin - Materials Engineer -
RES/DE
* Ian Tseng - Mechanical Engineer -
NRR/DEX
* Hanh Phan - Senior Reliability and Risk Analysis Analyst - NRR/DANU
* Steve Philpott - Project Manager -
Review    NRR/DANU
* Tim Lupold - Senior Mechanical Engineer - NRR/DANU Team
* Robert Roche-Rivera - Program Manager - RES/DE
* Steve Downey - Senior Reactor Inspector - R-II
* Isaac Anchondo-Lopez - Reactor Inspector - R-IV And other SMEs as needed Slide 146 3
 
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 Slide 147          4
 
Federal Register announcement of availability for public comment Regulatory 45-day public comment period Guide Project  Public comment resolution Schedule Issue as Final Regulatory Guide shortly after next ASME Code Edition Rulemaking
* Goal is by June 30, 2022 5            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 3
Section1      A      Section  2 B    Section C
* Purpose
* Background
* Staff Regulatory
* Applicability (non-LWRs)
* Bases for NRC      Guidance Staff Positions    (Conditions)
* Regulations &
Related Guidance 8
Slide 151
 
Use of license condition
* Proposed regulatory tool Difference between LWRs and non-LWRs Regulatory
* LWRs: 50.55a governs
* Non-LWRs: no clear regulation Aspects
* Part 50 Appendix A is guidance for developing PDC for non-LWRs
* RG 1.232 ARDC
* 10 CFR 50.34(b)(6)(iv) and 52.79(a)(29)(i)
* plans for conducting normal operations, including maintenance, surveillance, and periodic testing of SSCs 9            Slide 152
 
DG-1383 Conditions Use in conjunction with a license condition
* Information to be submitted with the application 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 Slide 153
 
DG-1383 Conditions - Cont.
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 Information within Code that is in course of preparation is to be developed and submitted for review and approval as appropriate by the applicant 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 Liquid leak test clarifications and hold time limits Minor errata type corrections 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}}

Revision as of 19:55, 16 January 2022

Sept 29 2021 Adv Rx Stakeholder Meeting Slides
ML21274A076
Person / Time
Issue date: 09/29/2021
From: Prosanta Chowdhury
NRC/NRR/DNRL/NRLB
To:
Hoellman J
References
Download: ML21274A076 (158)
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