ML23200A301

Presentation Slides - Periodic Advanced Reactor Stakeholder Meeting 07/20/2023

ML23200A301
Person / Time
Issue date:	07/20/2023
From:	Katie Wagner NRC/NRR/DANU/UARP
To:
References
Download: ML23200A301 (1)

Text

Advanced Reactor Stakeholder Public Meeting July 20, 2023 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 501 432 683#

Time Agenda Speaker 10:00 am - 10:10 am Opening Remarks / Advanced Reactor Integrated Schedule NRC 10:10 am - 10:45 am Environmental Center of Expertise Licensing Review Overview and Enhancements NRC 10:45 am - 11:00 am Introduction to the New Fuels Atlas NRC 11:00 pm - 11:30 am Update on SCALE/MELCOR Non-LWR Source Term and Fuel Cycle Demonstration NRC Project 11:30 am - 12:00 pm Nuclear Supplier QA Program Qualification: ISO 9001 Supplemental Nuclear Energy Requirements Institute (NEI) 12:00 pm - 1:00 pm Lunch Break All 1:00 pm - 1:30 pm Insights on Role of Advisory Committee on Reactor Safeguards (ACRS) During NRC Initial Licensing Reviews 1:30 pm - 2:30 pm Advanced Reactor Population-Related Siting Considerations NEI 2:30 pm - 2:45 pm Break All 2

Time Agenda (continued) Speaker 2:45 pm - 3:15 pm MACCS Consequence Analysis Demonstration Calculations NRC 3:15 pm - 4:15 pm Factory-Fabricated Transportable Micro-Reactor Licensing and Deployment NRC Considerations 4:15 pm - 4:30 pm Break All 4:30 pm - 5:00 pm Risk-Informed Approach to Package Approval for Transportable Microreactors NRC 5:00 pm - 5:05 pm Future Meeting Planning and Concluding Remarks NRC 3

Availability of Interim Joint Report on:

• Interim report available through link on:

Classification of https://www.nrc.gov/reactors/new-reactors/advanced/international-Structures, cooperation/collaboration-with-canada.html

• Also available in NRC ADAMS at ML23172A201 Systems, and

• Feedback is welcome, especially related to enhancing usefulness for potential license applicants Components

• Final report expected to be issued this fall

Contact:

Steve.Jones@nrc.gov or Jorge.Hernandez@nrc.gov 44 4

NRCs Advanced Reactor Readiness By the Numbers Statistics since 2018 Work on more than Completed more than Completed more than 35 policy issues 10 advanced reactor design 75 topical reference models to make report/white paper created more than future assessments more reviews 60 guidance documents.

efficient. 33% faster than the generic schedule goal.

Completed 10 NRC/DOE Kairos MOUs construction focused on permit safety advanced reactor review 50%

collaboration. Established core faster than review teams of the generic More than 140 public schedule goal.

Canada collaboration generated 8-10 technical staff engagements per year more than 10 work plans, per application, on advanced reactor-9 NRC/CNSC joint reports. based on recent related topics new reactor review experience.

The NRCs strategic transformation and modernization enables the safe deployment of ADVANCED REACTORS

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

https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 6

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 7

Environmental Center of Expertise Licensing Review Overview and Enhancements Kenneth Erwin Office of Nuclear Material Safety and Safeguards Division of Rulemaking, Environmental, and Financial Support Environmental Center of Expertise Advanced Reactor Stakeholder Meeting July 20, 2023 8

Principle Legislation, Regulations, and Outcome

• National Environmental Policy Act (NEPA)

• Atomic Energy Act (1954)

(1969)

• Energy Reorganization Act of (1974)

• National Historic Preservation Act (1966),

• 10 CFR Parts as applicable (e.g., 20, 40, 50, 52, Endangered Species Act (1973), others 70, 71, 72, 100)

• 10 CFR Part 51

• Risk Informed

• Impact-focused analyses

• Reasonable assurance of adequate protection

• Disclosure document (CatEx, EA/FONSI, (SER)

EIS/ROD)

Other Important Statutes

• National Waste Policy Act (1982)

CatEx - Category Exclusion

• Energy Policy Act (2005)

EA - Environmental Assessment

• Nuclear Energy Innovation and Modernization Act (amended 2019) FONSI - Finding of No Significant Impact

• Title 41 of Fixing Americas Surface Transportation Act (FAST-41) (2015) EIS - Environmental Impact Statement

• Fiscal Responsibility Act (2023) ROD - Record of Decision SER - Safety Evaluation Report 9 9

Licensing Process and the Environmental Review ACRS Review ACRS Letter Safety Safety Review Evaluation Report Pre-Application Opportunity for Mandatory Commission Application Contested Hearing Hearing*

Activities Decision Environmental Scoping Draft Comments on Final Review Activities NEPA Draft NEPA Document Document ACRS - Advisory Committee on Reactor Safeguards NEPA - National Environmental Policy Act Consultations Public Participation

• Required for early site permits, construction permits, or combined licenses 10

Resource Areas Analyzed in NRCs NEPA Reviews*

Describe the affected environment (baseline conditions) for each resource area and then the consequences of the action (impact level).

Analyze cumulative impacts from past, present, or reasonably foreseeable future actions.

• Typical resources analyzed in new reactor NEPA reviews 11 11

Coordination with Federal, State, local, and Tribal Government Agencies 12 12

NRC Environmental Center of Expertise (ECOE)

• Genesis: Established in October 2019.

• Organization: Centralized Branches From Different Offices Into One Division in the Office Material Safety and Safeguards (NMSS)

• Mission: Performs NRC-wide NEPA Reviews

- Streamlined processes, procedures, and guidance

- Developed common skillsets, knowledge management, project tracking

- Created NEPA Toolbox, Project Management Handbook, and internal NRC web pages (Nuclepedia)

• Strategic Plan: Implement NRCs Mission and Vision (Link)

- Ensure the safe and secure use of radioactive materials

- Continue to foster a healthy organization

- Inspire stakeholder confidence in the NRC 13

NRC Environmental Center of Expertise (ECOE)

• Transformation: Improve the efficiency and effectiveness of NRCs environmental review process

• Implemented Formal Internal Self Assessments and Transformation Efforts

- Streamlined reviews and focus more on impacts that matter

- Solicited input from staff, management, business line owners, stakeholders

- Ensured NRCs NEPA obligations are met and defensible 14

Transformation Efforts

• Manage Projects using commercial project management software and apply agile project management techniques for workload optimization

- Track project schedules, skillsets, and priority across all business lines in NRC (>200 projects)

- Allows for prioritization, flexibility, and agility

- Ensure targets and goals are met

• Incorporation of Lessons Learned

- Clinch River Early Site Permit (ML19190A078)

- Feedback from Public Meetings (ANR Stakeholder, Scoping, Draft EIS meetings)

- Regulatory Information Conference (RIC) Sessions in 2020 (RIC 2020), 2021 (RIC 2021), and 2023 (RIC 2023 Environmental and RIC 2023 Siting)

- Various stakeholder input (ML20065N155) and responses (ML20147A540) and (ML20183A475)

- Regulatory Guide 4.2 Update and Feedback (2018) (ML18071A400)

- Internal Review of New Reactor Reviews Lessons Learned Report for Environmental Reviews conducted in 2017 15

Transformation Outcomes

• Developed Advanced Nuclear Reactor (ANR) Generic Environmental Impact Statements (GEIS) (ML21222A044)

- Technology Neutral, Performance Based (Site and Plant Parameter Envelopes) Framework

- Use of bounding analysis and related concepts

- Under review by the Commission

• Update to License Renewal Generic Environmental Impact Statements (GEIS) (FRN)

- Incorporated lessons learned from previous License Renewal reviews

- Accelerated schedule per Commission direction

• Improved NEPA documentation

- Developed Templates

- Improved Readability/Reduction in Redundancy

- Increased use of Incorporation by Reference 16

Transformation Outcomes

• Continued Full Participation in Congressional and Administration Efforts and Situational Awareness, As Appropriate

- Nuclear Energy Innovation Capabilities Act (Law)

- Nuclear Energy Innovation and Modernization Act (Law)

- Advance Act (Proposal)

- FRA (Law)

- FAST-41 Implementation through Federal Permitting Improvement Steering Council (FPISC)

(Administration Website)

- Council on Environmental Quality (CEQ) Guidance Updates (FRN)

- CEQ Regulatory Changes (CEQ Website)

- Executive Orders

• 10 CFR Part 51 SECY Paper and Alternatives Analysis Concepts Therein (ML20212L393)

• Initiated Update to Environmental Standard Review Plan (January 2023)

• Preparation of Brownfield Paper (November 2022)

• Developed MOUs with Other NEPA Responsible Agencies (i.e., Department of Energy) 17

Enhancing Advanced Reactor Reviews

• Robust Pre-application Engagement and Readiness Assessments

- Support and Implement Business Line Owners Efforts

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

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

- Pre-application Engagement to Explain NEPA Requirements and Support Readiness Assessments and use of NRR Office Instruction LIC 116

• Expanded Use of Public Meetings and Regulatory Audits

- NMSS follows NRR Office Instruction LIC-111

- Optimization based on lessons learned

• Optimized use of Requests for Additional Information (RAIs)

- NMSS follows NRR Office Instruction LIC-115

- Management review of RAIs before issuance

- Coordination with business line owner

• Transparency through use of Dashboards 18

Enhancing Staff Capability and Capacity

• Multidisciplinary core review teams to focus reviews, as appropriate

• Formal Qualification Program for Project Managers and Technical Reviewers

- Building capacity for multiple ongoing reviews

- Hiring new staff

- Training staff on advanced reactor, fuel cycle, and license renewal technology

- Use of contractors for flexibility and agility

• Pre-application engagement with staff regarding site-specific NEPA resource technical review areas will support efficient reviews

• Timely information on industry plans supports effective NRC resource planning 19

Successfully Implementing Enhancements

• Kairos Hermes Test Reactor Construction Permit (CP) review

- Draft EIS (Link) issued September 29, 2022, Final EIS expected September 2023; With Significant Improvements to Transparency, Accountability, and Readability

- Supporting 21-month review schedule

• Dashboards

• Audits

• Internal project controls

• Multidisciplinary core review team

• Abilene Christian University Molten Salt Research Reactor CP review

- Innovative use of Environmental Assessment (FRN)

• Pre-application reviews ongoing with multiple developers/site owners

- Regulatory Engagement Plans

- Successful completion of Pre-application Readiness Assessments, Public Meetings, and Site Visits

- Pre-application assessments enhance readiness and quality of applications 20

Successfully Implementing Enhancements

• Support for Rulemakings and Policy Issues

- 10 CFR Part 53 EA (FRN)

- Catex (FRN)

- Fusion (Website)

- Accident Tolerant Fuel (Website) 21

Next Steps

• Continue stakeholder engagement through our pre-application readiness assessments/engagements and periodic advanced reactor public meetings

- Share best practices with prospective applicants

• Continue to make enhancements to internal processes based on lessons learned from ongoing reviews and stakeholder input

- Continue to assess our review processes during ongoing reviews

• Examine use of programmatic or sitewide EISs 22

Abbreviations/Acronyms 10 CFR - Title 10 of the Federal Codes of Regulations ACRS - Advisory Committee on Reactor Safeguards ANR - Advanced Nuclear Reactor BL - Business Line CatEx - Category Exclusion CEQ - Council on Environmental Quality EA - Environmental Assessment ECOE - Environmental Center of Expertise EIS - Environmental Impact Statement FAST Title 41 of Fixing Americas Surface Transportation Act FONSI - Finding of No Significant Impact FRN - Federal Register Notice GEIS - Generic Environmental Impact Statement MOU - Memorandum of Understanding NRC - Nuclear Regulatory Commission NEPA - National Environmental Policy Act NMSS - Office of Nuclear Material Safety and Safeguards RAI - Request for additional information RIC - Regulatory Information Conference ROD - Record of Decision SLR - Subsequent License Renewal 23

Introduction to Chris Markley NMSS/DFM the New Fuels Advanced Reactor Atlas Stakeholder Meeting July 20, 2023 24

New Fuels Environment New fuels arena is evolving quickly

Purpose:

Enhance ability to identify and process information Outcome: New Fuels Atlas Enhanced communications

• Infographic

• New Fuels Website Enhanced organization

• Regulatory Planner 25

New Fuels Infographic Looks at all phases of the front and back end of the fuel cycle Provides the who, the what, and the how Highlights information for public stakeholders Framework supports current environment NRC has tools available to regulate 26

NEW FUELS Readiness for new non-light water reactor FUELS the Who -

The Nuclear Regulatory Commission (NRC) is an independent agency that oversees the civilian use of nuclear materials.

Regulatory tools include:

• Licensing - adequate protection

• Oversight - to ensure compliance the What -

• Research - to support Industry is expected to deploy new nuclear development of technical bases fuel technologies. Here we focus on

• International Activities -

advanced non-light water reactor fuel like multilateral and bilateral metals and salts. information exchanges the How -

• Rulemaking - to codify safety requirements The NRCs current regulatory framework can support deployment of anticipated the Who, the What, and the How... new nuclear fuels.

The existing tools can accommodate new fuels!

New Fuels Licensing Activities Inspection &

Enforcemen Regulatory Inspection Regulatory Related Information: t Framework & Oversight Framework

• Accident Tolerant Fuel

• High-Assay Low-Enriched Uranium (HALEU)

Cr, Ch, MC&A Cr, M, T Cr, M,

• NEIMA Review Schedules WM

• Hearing Opportunities

• Public Involvement Ch: Chemical Analysis, Cr: Criticality Analysis, M: Materials Properties & Compatibility, MC&A: Material Control and Accounting, T: Thermal Analysis, WM: Waste

New Fuels Website Enrichment Fabrication Transportation Utilization Safety Environmental Protection Security and Safeguards Stakeholder Engagement 28 https://www.nrc.gov/materials/new-fuels.html

The Regulatory Planner Organizational tool For each technology Fuel cycle phase Programmatic area 29

Any Questions?

30

Update on SCALE / MELCOR non-LWR Source Term and Fuel Cycle Demonstration Project Advanced Reactor Stakeholder Meeting July 2023 Lucas Kyriazidis & Shawn Campbell Office of Nuclear Regulatory Research Division of Systems Analysis Fuel & Source Term Code Development Branch 31

NRCs Strategy for Preparing for non-LWR Applications

• NRCs Readiness Strategy for Non-LWRs Volume #1 Systems Analysis Volume #5 Volume #2 Nuclear Fuel Fuel Cycle IAP Strategy #2 Performance

• IAPs are planning tools that describe: Computer

- Work, resources, and sequencing of work to achieve readiness Codes and Tools

• Strategy #2 - Computer Codes and Review Tools Volume #4 Volume #3 Source Term,

- Identifies computer code & development activities Licensing &

Dose Consequence

- Identifies key phenomena

- Assess available experimental data & needs 32

NRCs Non-LWR Demonstration Projects & Codes NRCs comprehensive neutronics package NRCs comprehensive severe accident

• Cross-section processing progression and source term code

• Decay heat analyses

• Accident progression

• Criticality safety

• Thermal-hydraulic response

• Radiation shielding

• Core heat-up, degradation, and

• Radionuclide inventory & depletion relocation generation

• Fission product release and transport

• Reactor core physics behavior Volume #3 Volume #5 Goal of Volume 3 & 5 is demonstration of SCALE & MELCOR for simulating non-LWRs 33

General Approach Representative Initial and Boundary

1. Build SCALE core models and MELCOR full- Conditions plant models Code

2. Select scenarios that demonstrate code Identify &

Address Development Simulating Accidents capabilities for key phenomena Modeling Gaps

& around Key Phenomena Assessment

3. Perform simulations & code assessments on SCALE & MELCOR Sensitivity Studies

• The scenarios and design assumptions were chosen to show capabilities of the new modeling features added to the codes.

• There is no significance to the magnitude of the releases in the MELCOR demonstration calculations.

• The results are not intended to provide accident source terms for use in licensing decisions.

34

Volume 3 Severe Accident Progression & Source Term 35

Availability of Volume 3 Reference Material Five Major Types of Non-LWRs Analyzed under Volume 3 2021

• Heat Pipe Reactor - INL Design A

• High-Temperature Gas-Cooled Pebble-bed Reactor - PBMR-400

• Molten-salt-cooled Pebble-bed Reactor - UCB Mark 1 2022

• Molten-salt-fueled Reactor - MSRE

• Sodium-cooled Fast Reactor - ABTR Public workshop videos, slides, reports at advanced reactor source term webpage SCALE input models available here.

MELCOR input models available upon request.

36

Non-LWR Designs Considered & Project Scope High-Temp. Gas Cooled Reactor Sodium-Cooled Fast Reactor Molten Salt-Cooled Reactor Molten Salt-Fueled Reactor Heat Pipe Reactor PBMR-400 ABTR UCB Mk1 PB-FHR MSRE INL Design A

• 400 MWth reactor, graphite

• 250 MWth pool-type reactor,

• 236 MWth reactor at atmospheric

• 10 MWth reactor, graphite

• 5 MWth with a 5-year operating moderated utilizing metallic U / HT-9 fuel rods pressures moderated at near atmospheric lifetime

• Helium-cooled & TRISO-particle

• Reactor fueled with U-Pu-Zr fuel

• Flibe cooled & Pebble fueled pressures

• 1,134 heat pipes fueled with UO2 pebble-fueled at 10 wt.% U-235 slugs (TRISO) at 19.9 wt.% U-235

• Reactor fueled with liquid dissolved fuel (19.75 wt.% U-235)

• Fuel discharged at high burnup (90

• Liquid sodium coolant

• Online refueling fuel in molten salt (34.5 wt. % U-

• Reactivity controlled via control GWd/MTIHM) 235) drums 37

Heat Pipe Reactor - INL Design A Accidents Modeled

• Transient overpower

• Loss-of-heat sink

• Anticipated transient w/o SCRAM Insights Gained on BDBA Behavior

• Following scram, passive heat dissipation into reactor cavity ends the release from fuel

• Reactor building bypass requires two failures in a single heat pipe - one in the condenser region and another in the evaporator region

• Significant uncertainty in the release fractions depending upon the assumptions. No significance to the magnitude of release.

38

Pebble-Bed Gas-Cooled Reactor - PBMR-400 Accidents Modeled

• Depressurized loss-of-forced circulation Insights Gained on BDBA Behavior

• Graphite oxidation from air ingress does not generate sufficient heat to impact fuel

• Passive heat dissipation into reactor cavity limits release from fuel failure

• A low graphite conductivity has the largest impact on the peak fuel temperature 39

Pebble-Bed Molten-Salt-Cooled - UCB Mark 1 ATWS with variable DRACS Accidents Modeled

• Anticipated transient w/o SCRAM

• Station-wide blackout

• Loss-of-coolant accident Insights Gained on BDBA Behavior

• For ATWS, fuel heat-up was limited by reactivity feedback & passive decay heat removal system

• For SBO, with failure of the passive decay heat removal system, coolant boiling occurred over the course of several days

• For LOCA, with one train of decay heat removal system operating, coolant boiling was possibly averted.

• For LOCA, with failure of the passive decay heat removal system, fuel damage occurred.

End of the Xenon transient and a return to power.

40

Molten-Salt-Fueled Reactor - MSRE Accidents Modeled

• Full reactor inventory molten salt spill without water

• Full reactor inventory molten salt spill with water Insights Gained on BDBA Behavior

• Auxiliary filter operation increases the release rate of noble gases to the environment while also filtering airborne aerosols

• Aerosol releases to the environment are reduced due to settling in the reactor cell, capture in the filter, and capture in the condensing tank in the water spill cases

• The aerosol mass in the reactor building also spanned many orders of magnitude depending on scenario assumptions 41

Sodium-Cooled Fast Reactor - ABTR Accidents Modeled

• Unprotected transient overpower

• Unprotected loss-of-flow

• Single blocked assembly Insights Gained on BDBA Behavior

• With ULOF, core power eventually converges on the DRACS heat removal rate

• Single blocked assembly leads to rapid fuel damage 42

What have we learned & where are we going?

1. SCALE & MELCOR Volume 3 Models Support Readiness for NRC Non-LWR Licensing Reviews

• Leveraged UCB Mk1 to Support NRRs review of the Hermes Construction Permit Application

2. Additional SCALE & MELCOR Code Enhancements & Capabilities In-Progress

• Integration of SCALEs ORIGEN module into MELCOR for higher fidelity MSR transient analyses

• Capability to model multiple working fluids in the same MELCOR plant model

• Demonstrate capability for horizontal heat pipe reactors

• Refinement of specialized models (e.g., fluid freezing and cascading heat pipe failures)

• Fission product chemistry refinement

3. New Upcoming NRC Public Works for Additional Studies

• SCALE & MELCOR Demonstration Calculations for a Molten Chloride Fast Spectrum Reactor

• Public report - SCALE modeling of the sodium-cooled fast-spectrum ABTR

• Public report - MELCOR Accident Progression and Source Term Demonstration Calculations for a Sodium-Cooled Fast-Spectrum Reactor 43

Volume 5 Radionuclide Characterization, Criticality, Shielding, and Transport for the Nuclear Fuel Cycle 44

Reminder: General Approach Representative

1. Build SCALE core models and MELCOR Initial and Boundary models Conditions

2. Select scenarios that demonstrate code Code capabilities for key phenomena Identify &

Address Development Simulating Accidents Modeling Gaps

& around Key Phenomena Assessment

3. Perform simulations & code assessments on SCALE & MELCOR Sensitivity Studies 45

LWR Nuclear Fuel Cycle & Regulations

• Protect workers, public and the environment against radiological and non-radiological hazards that arise from fuel cycle operations.

• Radiation hazards

• Radiological hazards

• Non-radiological (chemical) hazard 46

Project Scope - Non-LWR Fuel Cycle Stages in scope for Volume 5 Enrichment Fuel Utilization Fresh Fuel UF6 Transportation Fuel Fabrication (including on-site UF6 enrichment Transportation spent fuel storage)

Stages out of scope for Volume 5 Uranium Mining & Milling

• Not envisioned to change from current methods.

Power Production

• Addressed in Volume 3 - Source Term& Consequence work Spent Fuel Off-site Storage &

Transportation

• Large uncertainties & lack of information Spent Fuel Final Disposal

• Large uncertainties & lack of information 47

Non-LWR Designs Considered High-Temp. Gas Cooled Reactor Sodium-Cooled Fast Reactor Molten Salt-Cooled Reactor Molten Salt-Fueled Reactor Heat Pipe Reactor PBMR-400 ABTR UCB Mk1 PB-FHR MSRE INL Design A

• 400 MWth reactor, graphite

• 250 MWth pool-type reactor,

• 236 MWth reactor at atmospheric

• 10 MWth reactor, graphite

• 5 MWth with a 5-year operating moderated utilizing metallic U / HT-9 fuel rods pressures moderated at near atmospheric lifetime

• Helium-cooled & TRISO-particle

• Reactor fueled with U-Pu-Zr fuel

• Flibe cooled & Pebble fueled pressures

• 1,134 heat pipes fueled with UO2 pebble-fueled at 10 wt.% U-235 slugs (TRISO) at 19.9 wt.% U-235

• Reactor fueled with liquid dissolved fuel (19.75 wt.% U-235)

• Fuel discharged at high burnup (90

• Liquid sodium coolant

• Online refueling fuel in molten salt (34.5 wt. % U-

• Reactivity controlled via control GWd/MTIHM) 235) drums 48

Representative Fuel Cycle Designs

• Developed 5 non-LWR fuel cycle design concepts

• HPR - INL Design A

• HTGR - PBMR-400

• FHR - UCB Mark 1

• MSR - MSRE Designs to be documented in publicly

• SFR - ABTR available report later this year

• Design concepts identify potential processes & methods

• What shipping package could transport HALEU-enriched UF6?

What are the hazards associated?

• How is spent SFR fuel moved? What are the hazards associated?

• How is fissile salt manufactured for MSRs? What are the various kinds of fissile salt that may be used? What are the hazards?

Used as the Initial and Boundary Conditions for developing SCALE & MELCOR models 49

non-LWR Fuel Cycle Demonstration Project -High Temperature Gas-Cooled Reactors

• HTGR Fuel Cycle Highlights

• Use of HALEU (19.75 wt.% U-235)

• No approved commercial-sized transport packages (UF6 &

fresh pebbles)

• New chemicals and processes for TRISO particle and pebble manufacturing

• TRISO fabrication Sol gel process ; pebble manufacturing

• Continuous fuel circulation, loading, and removal

• Accidents Modeled

1. Criticality due to HALEU-enrichments - UF6 and fuel pebble operations

2. Hazards associated with new chemicals (e.g., spills, water interaction, fire)

3. Fission product release from damaged fuel pebbles during fuel handling 50

SCALE & MELCOR Analyses for Selected Accidents from the HTGR Fuel Cycle Criticality-related Analyses Spent Fuel Pebble Inventory & Fission In-Facility UF6 Release Product Release Water ingress into the DN30-X during UF6 transport UF6 Cylinder Rupture

• Simulated UF6 enriched to 10 & 20 wt. % U-235 Spent Fuel Storage Tank Release

• Spent fuel tank holding 620,000 pebbles simulated

• UF6 cylinders are overfilled & heated resulting in

• Shown to be subcritical rupture and release

• Approx. 500 pebbles discharged daily / 1284 days to fill SFT

• Total decay heat and inventory of SFT determined 51

Future Work for Volume 5

1. New Upcoming Planned Workshop

• September 2023 - Sodium Fast Reactor Nuclear Fuel Cycle Analyses

• 2024 - Molten Salt Fueled Reactor Nuclear Fuel Cycle Analyses

2. Upcoming Public Report(s)

• Summer 2023 - Non-LWR Fuel Cycles for Severe Accident Simulations

3. Additional SCALE & MELCOR Code Enhancements & Capabilities In-Progress

• New capabilities planned in SCALE for handling irregular geometries in SCALE (fuel reprocessing)

• Leveraging newly developed capabilities to SCALE & MELCOR from Volume 3 52

Nuclear Supplier QA Program Qualification: ISO 9001 Supplemental Requirements Mark Richter-Technical Advisor Nuclear Energy Institute July 20, 2023

©2023 Nuclear Energy Institute 53

Context Development plans for advanced reactors is of a scope and scale not before seen in our industry The current operating fleet must be supported for 40-60 years or more of safe and reliable operations The current supply chain will be challenged to meet the dynamic and growing demand as well as aggressive timelines for new parts and components Anticipated supply chain challenges will require new and transformative quality management approaches Opportunity for NRC to demonstrate regulatory leadership as a modern regulator, seeking new efficiencies in regulatory processes during a period of dynamic industry growth

©2023 Nuclear Energy Institute 54

ISO-9001 Approach to Meet Appendix B ISO-9001 is already employed in many other industries and some nuclear suppliers already have ISO-9001 programs NEI is developing a process whereby an ISO 9001 QA program, with enhancements, could be used as a framework for meeting the requirements of 10CFR50, Appendix B, leading to a more nimble and responsive supply chain. It is anticipated that this will be helpful in both maintaining the operating fleet as well as developing and deploying several hundred advanced reactors over the next decade.

NOT proposing ISO 9001 as a replacement for 10 CFR Part 50 Appendix B

©2023 Nuclear Energy Institute 55

Not Starting From Scratch NRC SECY-03-0117 Approaches for Adopting More Widely Accepted International Quality Standards compares ISO-9001-2000 against the existing 10 CFR Part 50 Appendix B requirements and recommends that supplemental requirements would be needed EPRI 1007937 Analysis and Comparison of ANSI/ISO/ASQ Q9001:2000 with 10CFR50 Appendix B: ISO-9001 Gap Analysis and EPRI 1002976, An In-Depth Review of Licensee Procurement Options for Use with ISO-9001 Suppliers Other regulated industries utilizing an ISO-9001 based quality program and their regulating bodies have recognized the need for and implemented supplemental requirements

©2023 Nuclear Energy Institute 56

Implementation of NEI 22-04 Decomposes each Appendix B criterion into discrete requirements and identifies comparable requirements from ISO 9001 Note potential gaps for compliance with the regulation Provide recommendations for addressing the gaps and are implemented contractually by the Appendix B purchaser and potential ISO-9001 supplier Purchaser maintains 10 CFR Part 21 responsibilities, and supplier/vendor will impose the reporting requirements for nonconformances with sub-tier supplier(s)

©2023 Nuclear Energy Institute 57

Status and Next Steps NEI 22-04 Nuclear Supplier QA Program Qualification: ISO 9001 supplemental requirements rev. 0 is essentially complete ISO-9001 supplier dry run assessments informed final draft (Assessments of Pioneer Motor Bearing and Penn United Complete)

NEI 22-04 undergoes broad industry review (Q3 2023)

NEI participates in pre-submittal meeting with NRC (Q3 2023)

NEI submits for NRC review and endorsement (Q4 2023)

©2023 Nuclear Energy Institute 58

Questions?

mar@nei.org

©2023 Nuclear Energy Institute 59

Advanced Reactor Stakeholder Public Meeting Lunch Break Meeting will resume at 1:00 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 501 432 683#

NRC Staff Interactions with ACRS on Kairos Hermes Construction Permit Safety Review Advanced Reactor Stakeholder Meeting July 20, 2023 Matthew Hiser Senior Project Manager Advanced Reactor Licensing Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities U.S. Nuclear Regulatory Commission

Background

• Kairos submitted 11 topical reports prior to the Hermes construction permit (CP) application

- All 11 topicals were approved prior to issuing the final Hermes CP safety evaluation

- 8 of the 11 topical reports were reviewed by ACRS between 2020 and early 2023 (see Appendix II of ACRS letter)

• ACRS had strong familiarity with the Kairos technology and key technical topics involved in the Hermes CP application review 62

Timeline of ACRS Interactions on Hermes CP

• April 2022: Kairos and NRC staff presented Hermes CP overview

• January - March 2023

- NRC staff provided preliminary Hermes CP SE chapters and key appendices to ACRS for review (all available in ADAMS under docket #05007513)

• March - May 2023: NRC staff and Kairos briefed ACRS on Hermes CP safety analysis and review

- Kairos subcommittee: March 1, March 23-24, April 4, April 18

- Full committee: May 3-4

• May 16, 2023: ACRS letter issued

• June 20, 2023: NRC staff response to ACRS letter issued 63

Staff Insights from ACRS Review of Hermes CP

• Staff appreciates the timely and thorough review of the Hermes CP application and safety evaluation

• ACRS used a risk-informed approach to focus on the most safety significant aspects of the design to ensure that the review was efficient and thorough

• ACRS review of preliminary safety evaluation chapters while the final safety evaluation was being assembled expedited ACRS review and accelerated project schedule 64

Questions?

Contact me by e-mail at Matthew.Hiser@nrc.gov or by telephone at (301) 415-2454 65

U.S. NRC ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS) REVIEWS OF NEW FACILITY APPLICATIONS Joy Rempe, Chairman Advanced Reactor Stakeholder Meeting July 20, 2023 66

ACRS Overview

• Provides Commissioners independent technical reviews of, and advice on, safety of proposed or existing reactor facilities, adequacy of proposed safety standards, and adequacy of NRC safety research program

- Statutorily mandated by Atomic Energy Act of 1954, as amended

- Operational practices governed by Federal Advisory Committee Act (FACA)

• Independent of NRC staff. Reports directly to the Commission, which appoints ACRS members For additional information about ACRS, see: https://www.nrc.gov/about-nrc/regulatory/advisory/acrs.html 6

7

ACRS Review of Proposed Facility

• Integrated review of applicant submittals (and associated staff safety evaluation) including:

- Safety Analysis Reports (for construction permit, operating license, early site permits, design certification, and standard design approvals)

- Topical Reports and possibly other supporting documents (white papers, technical reports, etc.)

• Typically includes one or more subcommittee meetings and at least one full committee meeting prior to issuance of ACRS letter report.

- Portion of meetings open to public, allowing opportunities for public comments

- Portion of meetings may be closed to allow discussion of proprietary information 6

8

ACRS Review of Proposed Facility (contd)

• ACRS developing best-practices guidance* to promote streamlined reviews focused on safety and risk significant aspects

- Implements lessons learned from recent design-centered reviews

- Lead ACRS member and ACRS staff work with cognizant NRR staff to develop committee engagement plan to optimize review schedule

- ACRS review completed after staff draft safety evaluation report completed**

• Typically includes topics*** such as: overall design (emphasizing unique and novel aspects), safety functions and principal design criteria, safety-related structures, systems, and components, licensing basis event selection, fuel qualification, safety analysis methods and results, and source term.

• Guidance, along with name of lead ACRS member for each design-centered subcommittee to be posted on https://www.nrc.gov/about-nrc/regulatory/advisory/acrs.html.

• • ACRS reviews may be concurrent with staff completion of some safety evaluation report chapters.

• • • Representative topics, not comprehensive. 6 9

Advanced Reactor Population-Related Siting Considerations July 20, 2023

© 2023 Nuclear Energy Institute

Acronyms BWR - Boiling Water Reactor CFR - Code of Federal Register DiD - defense-in-depth EAB - exclusion area boundary LPZ - low population zone LWR - light water reactor MWth - megawatt thermal PCD - population center distance PDD - population density distance ppsm - persons per square mile PWR - Pressurized Water Reactor SAMA - Severe Accident Mitigation Alternatives SECY - Commission paper SOARCA - State-of-The-Art Reactor Consequence Analysis SRM - Staff Requirements Memorandum ©2023 Nuclear Energy Institute 71

=

Background===

10 CFR Part 100.21 requirement:

(h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable3.

©2023 Nuclear Energy Institute 72

Guidance for Implementation Regulatory Guide (RG) 4.7, General Site Suitability Criteria for Nuclear Power Stations

• Current criterion: population density not exceeding 500 persons per square mile (ppsm) out to 20 miles

• Based on large light water reactor experience SRM-SECY-20-0045 directed NRC staff to revise RG 4.7 guidance relating to 10 CFR 100.21(h) to include provisions for advanced reactor designs

• No greater than 500 ppsm out to twice the distance at which 1 rem dose for the 30-day exposure period is calculated based on design-specific events

©2023 Nuclear Energy Institute 73

Industrys Effort to Provide Feedback Goal - Assess NRCs pre-decisional white paper, Alternative Approaches to Address Population-Related Siting Considerations made public in April 2023 and provide industrys observations to inform further discussions Objectives

• Put the population-density siting consideration in context with other siting elements and defense-in-depth considerations

• Compare the level of protection afforded as proposed by NRC for advanced reactors to that currently applied to existing LWRs

• Identify whether NRCs guidance would result in undue burden (i.e., excessive restrictions on siting) for advanced reactors Scope - Cover all advanced reactor designs except the following

• Large (gigawatt scale) designs

©2023 Nuclear Energy Institute 74

Population-Density Siting Consideration in Context

NRC Advanced Reactor Policy Statement At least the same degree of protection of the environment, public health &

safety, and common defense and security Enhanced margins of safety and/or use of simplified, inherent, passive or other innovative means to accomplish safety & security functions Designs that incorporate the defense-in-depth philosophy

©2023 Nuclear Energy Institute 76

Siting Criteria / Limitations Power reactor siting has typically involved assessment of a variety of distances, most of which are depicted in Fig. 1 Each provides functional and defense-in-depth (DiD) purposes Siting criteria protect from societal impacts & provides DiD to minimize societal impacts should containment fail*

• as we understand it; based on TID 14844 (1962) Source: NRC SECY 20-0045 Figure 1 ©2023 Nuclear Energy Institute 77

Siting Criteria / Limitations (cont)

Prior to 1973, PDD was not considered (no d20 - only dEAB , dLPZ and dPCD)

WASH-1308 (1973) suggested need for an RG and that there should be a PDD, but different approach (equivalent to ~1600 ppsm)

RG 4.7, R0 (1974) did not include PDD (no d20 - only dEAB , dLPZ and dPCD)

RG 4.7, R1 (1975) added PDD, but it was d30

• 500 ppsm, different than what was proposed in WASH-1308

• No reference to a dose basis RG 4.7, R2 (1998) changed PDD to d20

• No reference to a dose basis

• States only that Numerical values in this guide are generally consistent with past NRC practice and reflect consideration of severe accidents, as well as demographic and geographic conditions characteristics of the United States.

ORNL/TM-2019/1197 (2019) evaluated various alternatives to the current requirements that would achieve a reduction of about an order of magnitude versus large LWRs, including ratio of thermal output and case-be-case, design-specific review, but did not recommend a specific dose criterion ©2023 Nuclear Energy Institute 78

Compare the Level of Protection How does the proposed PDD criterion compare to the existing 20-mile PDD for Large LWRs?

Existing basis for 20-mile PDD

• Deterministically set for current large LWRs; no calculation of the dose at that distance required.

• NRC states in the pre-decisional white paper that the 20-mile distance was based on insights from probabilistic risk assessments and other studies associated with light-water reactor designs.

• Have not found documentation for original basis other than as summarized on previous slide.

NRC proposed alternative advanced reactor PDD

• Distance that is equal to twice the distance at which a hypothetical individual could receive a calculated TEDE of 1 rem over a period of 1 month from the release of radionuclides following postulated accidents

• We could not find a clear technical basis for this criterion and so the basis is not well-understood

• Dose at PDD (twice the distance to 1 rem) will be much less than 1 rem, on the order of doses from background radiation

©2023 Nuclear Energy Institute 80

What model inputs and assumptions does NRC expect applicants to use?

Some do not appear to be specified in NRCs pre-decisional white paper, Alternative Approaches to Address Population-Related Siting Considerations Based on past practice, we assumed the NRC might expect the most conservative case

• No credit for shielding with normal activity; individual is modeled as residing outdoors unprotected for 30 days

• 95% Weather - WD Maximum; very conservative weather assumptions Note: industry does not believe the worst case is necessary, but wanted to start from the worst case that we believed the NRC might expect

• In any case, we compare models done with consistent assumptions

©2023 Nuclear Energy Institute 81

Population Density Distance (PDD) Dose Historical Effectiveness Jensen Hughes conducted a scoping atmospheric dispersion consequence analysis using the WinMACCS code Scoping model purpose: to estimate dose calculated using the existing (LWR based) PDD siting criterion (i.e., the level of protection actually provided by the 20-mile limit for the current fleet) for comparison with the NRC proposal for advanced reactors of twice the distance to 1 rem dose over 30 days

©2023 Nuclear Energy Institute 82

LWR Scoping WinMACCS Model Inputs and Assumptions Started with NRC Linear No-Threshold (LNT) Point Estimate Sample Problem distributed with WinMACCS version 3.10 Core inventory taken from the NRCs SOARCA for the Surry reactor and ratioed up by 50% (effectively representing 3819 MWth)

• reasonably representative of a typical large LWR Radionuclide release magnitudes (to the environment) based on calculating an average of the total release fraction of the highest releases from 13 recent SAMA analyses (both PWR and BWR) using a frequency screening

• WinMACCS typical three plume model/assumptions used for total release Surry meteorology data per example data with WinMACCS code Dose was calculated for 30 days, with no credit for protective actions

©2023 Nuclear Energy Institute 83

LWR Scoping WinMACCS Model Eight cases

• Cases 1 - 4 represent the 30-day dose for individuals who maintain normal activity during the release and thereafter

• Cases 5 - 8 eliminate credit for shielding with normal activity; individual is modeled as residing outdoors unprotected for 30 days

• Cases reflect four different levels of weather conditions assumptions Case 8 results are the most restrictive

©2023 Nuclear Energy Institute 84

LWR Scoping WinMACCS Model Results Observation: The Commission has previously stated that the current generation of plants is adequately safe

• Current 20-mile distance results in acceptable societal risks

• Doses for large LWR at 20 miles for 30 days do not pose undue health risks Applying NRC proposed twice the distance to 1 rem over 30 days

• This dose criterion is orders of magnitude less than what is currently accepted by NRC

• Distances produced using this criterion would be an order of magnitude more than 20 miles

• NRC proposed criterion is excessively restrictive on siting

©2023 Nuclear Energy Institute 85

Advanced Reactor Scoping WinMACCS Model Estimates calculated for two advanced reactor designs based on the generic LWR WinMACCS model (assumed representative of others)

• High-Temperature Gas Reactor with tri-structural isotropic (TRISO) fuel

• Molten Salt Reactor Release-related inputs for these scoping HTGR and MSR models were based on data developed by Sandia in SAND2020-0402 for performing simplified scoping assessments of advanced reactors Used a postulated maximum credible accident (MCA) based upon review of the Sandia study

• Small core inventory, e.g., 250 MWth

• Assumed percentage of fuel damage; percentage of fission product migration from fuel to environment; degraded containment Dose was calculated for 30 days, with no credit for protective actions (same as large LWR model) ©2023 Nuclear Energy Institute 86

Advanced Reactor Scoping WinMACCS Model Results Observations from large LWR model results (slide 16) also applicable to advanced reactors NRC proposed PDD estimates for advanced reactors would be excessively more conservative than what the NRC currently finds acceptable for large LWRs If the current level of protection for the existing LWR fleet is used, the PDD for these advanced reactors could be well within a site boundary NRCs proposed PDD approach would create excessive restrictions on the ability to site advanced reactors - far beyond what NRC imposes on large LWRs today when the comparative size of source terms in considered

©2023 Nuclear Energy Institute 87

Identify Whether NRC Approach Results in Undue Restrictions on Siting

Does NRC approach Appropriately Credit Safety Features of Advanced Reactors?

SRM-SECY-20-0045 approved the staffs recommended option to revise the guidance in RG 4.7 to include technology-inclusive, risk-informed, and performance-based criteria for population densities that are based on estimates of radiological consequences from design-specific events

• NRCs proposed PDD criterion would result in siting restrictions that do not reflect the potential for enhanced safety and reduced risks associated with radiological releases from advanced reactor designs

• More guidance needed to balance realistic/conservative assumptions

• The impact on micro-reactors was not analyzed, but can be anticipated as consistent with other observations

• Layers of DiD applied in the design deserve credit, as applicable, e.g.,

small source term, slow accident progression, low operating pressure

©2023 Nuclear Energy Institute 89

Summary Observations Key Take-Aways for Further Discussion Alternate dose criterion should be developed to be more representative of the currently accepted level of protection for large LWR licensing.

• The proposed criterion associated with very low frequency events is excessively conservative compared to previous large LWR licensing and compared to the annual exposure of the public from natural and man-made sources (e.g., medical procedures).

• The undue burden created by excessive conservatism significantly restricts advanced reactor siting as compared to the NRCs currently accepted approach for large LWRs.

Clarity is needed on the modeling assumptions, which heavily influence dose criterion calculations, including consideration with respect to the realistic exposure risk to the public that would be acceptable to the NRC NEI is preparing a white paper with more detail on our observations to inform NRCs consideration of revisions to draft guidance.

©2023 Nuclear Energy Institute 91

Advanced Reactor Stakeholder Public Meeting Break Meeting will resume at 2:45 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 501 432 683#

MACCS Consequence Analysis Demonstration Calculations Advanced Reactor Stakeholders Meeting July 2023 AJ Nosek, PhD Reactor Systems Engineer Accident Analysis Branch, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission

Purpose

• The purpose of this project was to:

- Assess the capabilities of the MACCS code to analyze a selected conceptual advanced reactor design under a postulated accident scenario.

- Identify potential gaps that may exist in conducting such an analysis, both technical and practical.

- Exercise new models (e.g., nearfield models) and new settings (i.e., inventory and source term) in the MACCS code for the selected advanced reactor design.

• The project continued similar code readiness work using SCALE and MELCOR:

- Oak Ridge National Lab (ORNL) developed a SCALE model to compute the radionuclide core inventory of the Idaho National Lab (INL) Design A conceptual reactor design.

- Sandia National Labs (SNL) developed a MELCOR model to simulate postulated accidents of the INL Design A conceptual reactor design.

94

Project Approach

• All MACCS analyses used the following inputs:

- An example inventory from a demonstration SCALE model (ORNL/TM 2021/2021)

- An example source term and reactor building dimensions from a MELCOR demonstration model (SAND2022-2745)

- Weather and regional characteristics from an existing (Sequoyah Nuclear Plant) site

- Other general settings defined in the MACCS parameter guidance report (NUREG/CR-7270) 95

Project Approach

• The main analysis used the following settings:

- A radionuclide list for consequence analysis from SAND2022-12018 based on example inventory from the demonstration SCALE model

- The Regulatory Guide 1.145 full model for nearfield transport

• Sensitivity analyses evaluated the following characteristics:

- Radionuclide lists for consequence analysis

- Dose exposure periods

- Nearfield models

- Release timings 96

Project Approach

• The project used MACCS v4.1.

• All reported doses are projected doses in the ambient environment.

• The MACCS calculations sample from a range of weather conditions as input. The mean, 5th quantile, and 95th quantile MACCS outputs represent the distribution of results due to weather uncertainty.

97

Example Accident Scenario Description

• To demonstrate code capabilities, the MELCOR team selected accident conditions for the MELCOR INL Design A model.

• This effort produced an example accident progression and source term for a transient overpower (TOP) accident scenario.

• The MACCS calculations use the example atmospheric release from the MELCOR demonstration project as input

• The MACCS demonstration calculations do not reflect realistic radiological consequences outside of the conditions assumed in the MELCOR analysis.

INL Design A reactor vessel cross-section (from figure 3-1 of SAND2022-2745) 98

Source Term*

• Two release pathways from building at 5.15 m and 9.15 m (16.9 ft and 64 ft).

• Release begins in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• Relatively small amount of release before MELCOR cutoff time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Project divided plume into 28, 1-hour segments.

• Note: Consequence results are shown to illustrate code capabilities only. Actual 99 consequences would be based on design, site, and scenario-specific factors.

Example of Main Analysis Results*

Conceptual Regulatory Criteria Dose Distance

• Note: Consequence results are shown to illustrate code capabilities only. Actual 100 consequences would be based on design, site, and scenario-specific factors.

Sensitivity Analysis of Radionuclides

• MACCS guidance recommends

• That list was expanded to using a subset of radionuclides include additional radionuclides from the LWR core for analysis from the conceptual HPR core Recommended Radionuclide List for LWR Additional Radionuclides from draft Applications (from NUREG/CR-7270, Table 2-2) SAND2022-12018, Table 4-2 Co-58 Y-90 Ru-103 Te-132 Ba-137m Nd-147 Ag-111 Ge-77 Pm-149 Sn-121 Co-60 Y-91m Ru-105 I-131 Ba-139 Np-239 As-77 Nb-95m Pm-151 Sn-123 Kr-85 Y-91 Ru-106 I-132 Ba-140 Pu-238 Cd-115 Nd-149 Pr-145 Sn-125 Kr-85m Y-92 Rh-103m I-133 La-140 Pu-239 Cd-115m Pd-109 Sb-125 Sn-127 Kr-87 Y-93 Rh-105 I-134 La-141 Pu-240 Eu-154 Pm-147 Sm-151 Te-125m Kr-88 Zr-95 Rh-106 I-135 La-142 Pu-241 Eu-155 Pm-148 Sm-153 U-234 Rb-86 Zr-97 Te-127 Xe-133 Ce-141 Am-241 Eu-156 Pm-148m Sm-156 U-237 Rb-88 Nb-95 Te-127m Xe-135 Ce-143 Cm-242 Sr-89 Nb-97 Te-129 Xe-135m Ce-144 Cm-244 Additional Daughter Radionuclides Sr-90 Nb-97m Te-129m Cs-134 Pr-143 Sb-127 Sr-91 Mo-99 Te-131 Cs-136 Pr-144 Sb-129 In-115m In-115 Th-230 Ra-226 Rn-222 Sr-92 Tc-99m Te-131m Cs-137 Pr-144m 101

Sensitivity Analysis of Radionuclides*

• A sensitivity using the updated radionuclide list shows little change in the results.

• The ratio of the peak doses from the new radionuclide list compared to the LWR list shows minimal increase in consequence.

• Note: this did not evaluate radionuclides important to ingestion consequences, which is based on a separate list

• Note: Consequence results are shown to illustrate code capabilities only. Actual 102 consequences would be based on design, site, and scenario-specific factors.

Sensitivity Analysis of Dose Exposure Period*

• The analyzed exposure periods begin at the start of the accident. Release begins quickly at less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• Longer exposure periods cause greater doses.

• At 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, only a fraction of the release has occurred, and no plume segments have travelled beyond 5 km (3.1 mi).

• Note: Consequence results are shown to illustrate code capabilities only. Actual 103 consequences would be based on design, site, and scenario-specific factors.

Sensitivity Analysis of Dose Exposure Period*

• Exposures during plume passage and exposures to short lived radionuclides occur only in the early phase of the accident.

• Nevertheless, exposure to ground contamination over the long term is the dominant contributor to the overall lifetime dose projection

• Note: Consequence results are shown to illustrate code capabilities only. Actual 104 consequences would be based on design, site, and scenario-specific factors.

Sensitivity Analysis of Nearfield Modeling

• Using the weather conditions of the Sequoyah Nuclear Plant, the project compared the following three MACCS nearfield modeling approaches:

(1) Regulatory Guide (RG) 1.145 Partial Model (with Area Source)

(2) RG 1.145 Full Model (with Point Source)

(3) Ramsdell and Fosmire Model (with Point Source)

• Both options 1 and 2 are based on the nearfield modeling approach described in RG 1.145.

- Option 1 is a partial implementation of RG 1.145. This model does not directly account for building wake. Instead, the project uses an area source based on the building size to model the building wake zone.

- Option 2 is a full implementation of RG 1.145. This modeling approach considers the effects of both building wake mixing and ambient plume meander.

• Option 3 is based on the Ramsdell and Fosmire nearfield modeling approach used by ARCON96. (SAND2021-6924)

• Options 2 and 3 are new models available in MACCS version 4.1 105

Sensitivity Analysis of Nearfield Modeling*

• The amount of plume spread in the different models notably impacts doses in the first 40 km (25 mi).

• Because the Ramsdell and Fosmire Model has the most spread, it has the lowest peak doses of the three nearfield models.

• The two RG 1.145 models have similar peak doses after roughly 1 km (0.62 mi), whereas the Ramsdell and Fosmire Model does not have the same peak doses until approximately 40 km (25 mi).

• Note: Consequence results are shown to illustrate code capabilities only. Actual 106 consequences would be based on design and site-specific factors.

Sensitivity Analysis of Nearfield Modeling*

• There are 64 compass directions in the analysis. Direction 1 represents north, and ascending numbers represent a clockwise direction.

• The double peak results are due to the meteorological conditions at the Sequoyah Nuclear Plant site, which has two dominant wind directions.

• The RG 1.145 Full Model has the narrowest plume, and therefore it has the highest doses in the dominant wind directions.

• The Ramsdell and Fosmire Model has more horizontal plume spread, creating a more uniform dose.

• The Ramsdell and Fosmire Model shows a lower dose in all directions, likely because of vertical plume spread.

• Note: Consequence results are shown to illustrate code capabilities only. 107 Actual consequences would be based on design, site, and scenario-specific factors.

Sensitivity Analysis of Release Timing

• MACCS assumes that reactor shutdown occurs at the beginning of an accident scenario. MACCS begins calculating decay and ingrowth at this time.

• The TOP scenario is different in that shutdown does not immediately occur.

• The reactor is postulated to operate for roughly an hour when the reactor power level steadily increases. This presents a few issues in computing offsite consequences:

- Since reactor shutdown does not immediately occur, the holdup time between reactor shutdown and the start of release is shorter than MACCS anticipates.

- The calculation of the core inventory assumes steady-state operation. If the reactor power level changes, it may not fully represent the new composition from the shift in fission rate before shutdown.

- Release can begin before reactor shutdown. MACCS is designed only to calculate decay and ingrowth from a core inventory given at a fixed time. It does not model production of fission and activation products during release.

108

Sensitivity Analysis of Release Timing

• A shift in the release timing impacts the dose. This sensitivity evaluates the dose over a 7-day exposure period.

• The TOP release has a start time of 0.94 hours0.00109 days <br />0.0261 hours <br />1.554233e-4 weeks <br />3.5767e-5 months <br />; a shift of -0.94 hours0.00109 days <br />0.0261 hours <br />1.554233e-4 weeks <br />3.5767e-5 months <br /> represents an immediate release.

• Despite a shift by up to 4 days, the projected dose remained within about 6 percent.

• Nearly half of this range (3 percent) can be attributed to a change in release timing of just 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (i.e., from -0.94 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).

Note: Consequence results are shown to illustrate code capabilities only. Actual 109 consequences would be based on design and site-specific factors.

Summary Results

• The MELCOR analysis selected accident conditions during a TOP accident scenario that produced an example source term of the INL Design A conceptual design.

• The sensitivity analysis of the nearfield models using the weather conditions of the Sequoyah Nuclear Plant site shows the following:

- The RG 1.145 Full Model has the highest peak dose.

- The Ramsdell and Fosmire Model has the most plume spread.

- The two RG 1.145 Models align quickly after 1 km.

- The Ramsdell and Fosmire Model may not align the RG 1.145 Models until 40 km (25 mi).

• The other sensitivity analyses show the following:

- Exposure to ground contamination over the long term is the dominant contributor to the overall lifetime dose projection in this example.

- The use of an expanded subset of radionuclides in this example consequence analysis showed only a minimal increase in consequence.

- A shift in the release timing for this example source term has a small impact on the projected peak dose.

110

Conclusions

• The results of our evaluation confirm that, despite some limitations, analysts can use the flexibility of the MACCS code to analyze the offsite consequences of an advanced reactor design under a postulated accident scenario.

• The evaluation exercise provided valuable practical experience in implementing new ORIGEN inventories and MELCOR source terms in MACCS.

• As new source terms of new and advanced reactor designs become available, RES staff may assess whether further enhancements to the MACCS code are needed.

• The project has identified several candidate future research activities:

- continue to demonstrate MACCS capabilities using as input the core radionuclide inventory and atmospheric release from the example SCALE and MELCOR demonstration calculations,

- continue the evaluation of radionuclides in non-LWR inventories important to dose and expanding these evaluations to include ingestion doses,

- develop a method to analyze or conservatively bound accidents with simultaneous release and fission, and

- develop methods to analyze or conservatively bound the impact of additional radionuclide chemical and physical forms and how they may transform in the environment.

111

Questions?

• A. Nosek, MACCS Consequence Analysis Demonstration Calculations for an Example Heat Pipe Reactor Source Term, U.S. Nuclear Regulatory Commission, March 2023, Agencywide Documents Access and Management System Accession No. ML23045A044

• E. Walker, S.E. Skutnik, W.A. Wieselquist, A. Shaw, and F. Bostelmann, SCALE Modeling of the Fast-Spectrum Heat Pipe Reactor, ORNL/TM-2021/2021, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 2021, ML22158A054.

• K. Wagner, C. Faucett, R. Schmidt, and D. Luxat, MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor, SAND2022-2745, Sandia National Laboratories, Albuquerque, New Mexico, March 2022, ML22144A188.

• K.A. Clavier, D.J. Clayton, and C. Faucett, Quantitative Assessment for Advanced Reactor Radioisotope Screening Utilizing a Heat Pipe Reactor Inventory, SAND2022-12018, Sandia National Laboratories, Albuquerque, New Mexico, September 2022, ML22270A046.

• Sandia National Laboratories, Technical Bases for Consequence Analysis Using MACCS (MELCOR Accident Consequence Code System), NUREG/CR-7270, U.S. Nuclear Regulatory Commission, Washington, DC, October 2022, ML22294A091.

• D.J. Clayton, Implementation of Additional Models into the MACCS Code for Nearfield Consequence Analysis, SAND2021-6924, Sandia National Laboratories, Albuquerque, New Mexico, June 2021, ML21257A120.

112

Licensing and Deployment Considerations for Factory-Fabricated Transportable Micro-Reactors Advanced Reactor Stakeholders Meeting July 20, 2023 William Kennedy Amy Cubbage Advanced Reactor Policy Branch U.S. Nuclear Regulatory Commission

Contents

• Goals of this presentation

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

• Regulatory Approaches for Fuel Loading at the Factory

• Regulatory Approaches for Operational Testing at the Factory

• Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps

• Next steps 114

Goals of this Presentation

• Inform stakeholders about regulatory approaches the NRC staff is developing for consideration by the Commission for fuel loading and operational testing at the factory

• Inform stakeholders about other licensing and deployment topics and potential near-term strategies and next steps the NRC staff is considering

• Receive feedback from stakeholders 115

Conceptual Deployment Model for Factory-Fabricated Transportable Micro-Reactors 116

Regulatory Approaches for Fuel Loading at the Factory

• The NRC staff is developing approaches for licensing fuel loading at the factory under the existing regulations for consideration by the Commission:

- Facility operating license issued pursuant to 10 CFR Part 50 that limits operation to fuel loading

- Combined license issued pursuant to 10 CFR Part 52 that limits operation to fuel loading

- Manufacturing license for manufacture and possession of the utilization facilities and a license to possess special nuclear material issued pursuant to 10 CFR Part 70 with provisions for the utilization facilities to include features to preclude criticality 117

Regulatory Approaches for Operational Testing at the Factory

• The NRC staff is developing approaches for licensing operational testing at the factory under the existing regulations for consideration by the Commission:

- Construction permit issued pursuant to 10 CFR Part 50, potentially covering many reactors, that would be converted to 10 CFR Part 50 facility operating licenses that limit operation to that needed for operational testing

- Combined licenses issued pursuant to 10 CFR Part 52, potentially issued at the same time based on one application, that limit operation to that needed for operational testing

- Construction permit issued pursuant to 10 CFR Part 50, potentially for many commercial non-power reactors, that would be converted to facility operating licenses that limit operation to that needed for operational testing 118

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Considerations related to initial fuel load and authorization to operate at the deployment site for reactors that arrive pre-loaded with fuel

• Deployment strategies that include loading fuel or operational testing at a manufacturing facility would result in fueled reactors arriving at the deployment site

• Several requirements in the Atomic Energy Act of 1954, as amended (AEA), and 10 CFR Parts 50 and 52 that are related to public notifications, the opportunity for hearing, authorization to operate the facility, and others are premised on fuel being initially loaded at the deployment site

• The NRC staff is considering whether there is a suitable alternative to initial loading of fuel at the deployment site that could be used as an alternate milestone and would accomplish the underlying purpose of the AEA and regulations 119

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Timeframe for authorization to operate at the deployment site

• Factory-fabricated transportable micro-reactors may have significantly simpler and shorter construction activities at the deployment site compared to large light water reactors and could be ready to begin operation in days to weeks to a few months after obtaining a construction permit or combined license

• Several requirements in the AEA and 10 CFR Part 50 and Part 52 that are related to the environmental review, the schedule for intended operation, public notifications, the opportunities for hearing, authorization to operate the facility, and others include timeframes that could add up to many months in total

• For licensing under 10 CFR Part 52, the NRC staff plans to clarify the circumstances under which the schedule for intended operation and initial fuel load can be accelerated and is considering ways to streamline public notifications, hearings, and the authorization to operate, as appropriate

• For licensing under 10 CFR Part 50, the NRC staff is considering opportunities to expedite steps in the processing and review of applications for facility operating licenses, such as acceptance review and docketing, milestones for hearings, and the supplement to the environmental impact statement 120

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Licensing replacement reactors

• Factory-fabricated transportable micro-reactors might be periodically replaced with reactors of the same design at the end of their lives or fuel cycles, and each reactor would be required to have its own combined license or facility operating license

• A licensee might have multiple fueled reactors on site in various states of operation and shutdown to allow for transition from the operating reactor to the replacement reactor with minimal downtime. This would need to be considered in the safety and environmental reviews

• The NRC staff previously addressed similar concepts and considered licensing options for multi-module facilities in SECY-11-0079, License Structure for Multi-module Facilities Related to Small Modular Nuclear Power Reactors, dated June 12, 2011 (ADAMS Accession No. ML110620459)

• The NRC staff is considering approaches under 10 CFR Part 50 and Part 52 where the construction permit application or combined license application would cover all reactors envisioned to be operated at the deployment site and each reactor would be authorized to begin operation under its own facility operating license or combined license once the Commission had made the required findings 121

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Autonomous and remote operations

• Proposed designs for factory-fabricated transportable micro-reactors (and potential designs for other types of reactors) might include autonomous and remote operational characteristics to reduce the number of operators and other categories of personnel at the facility site

• As previously noted in SECY-20-0093, Policy and Licensing Considerations Related to Micro-Reactors, dated October 6, 2020 (ADAMS Accession No. ML20129J985), both autonomous and remote operations raise potential policy-related matters

• The NRC staff plans to further develop its understanding of the industry deployment models for factory-fabricated transportable micro-reactors with respect to industry plans for remote and autonomous operations, identify any gaps in the existing human factors engineering review needed to address the deployment models, and develop the technical bases for any new guidance that may be needed

• As part of the proposed Part 53 rulemaking provided to the Commission, the NRC staff has proposed a new risk-informed, performance-based, technology-inclusive cybersecurity framework that would require licensees to demonstrate protection against cyberattacks in a manner that is commensurate with the potential consequences from those attacks 122

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Transportation of fueled reactors

• Factory-fabricated transportable micro-reactor developers (and potentially developers of floating nuclear power plants that use reactors with higher power levels) envision transporting fueled reactors from a fabrication site or a refurbishment and refueling facility to the deployment site for operation and later removing fueled reactors from the deployment site at the end of their useful lives or fuel cycles

• Transportation packages for factory-fabricated transportable micro-reactors may consist of the reactor itself or the reactor plus additional overpack, as needed. Packages for transporting a micro-reactor from the factory to the deployment site could be either a Type A fissile (Type AF) or Type B fissile (Type BF) package, as defined in 10 CFR Part 71

• The NRC staff intends to use the existing regulatory framework (primarily 10 CFR Part 71) to review transportation of fueled commercial micro-reactors in the near term, which may include the use of the alternate test criteria in 10 CFR 71.41(c), the special package authorization option in 10 CFR 71.41(d), or exemptions, as appropriate 123

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Storage of fuel after irradiation in a power reactor

• Depending on the duration between withdrawal of the fuel from the reactor (or the final reactor shutdown) and placement into a dry storage facility, different regulations may apply to the storage of the reactor fuel

• The definition of spent fuel in 10 CFR 72.3 includes criteria that the fuel has been withdrawn from a nuclear reactor following irradiation and has undergone at least one year's decay since being used as a source of energy in a power reactor

• In order to store irradiated power reactor fuel that had been withdrawn from a reactor for less than a year in an independent spent fuel storage installation, the licensee would be required to apply for a specific license under 10 CFR Part 72 and request and justify exemptions addressing the one-year decay time requirement in the regulations

• The NRC staff intends to engage with stakeholders as they further develop their strategies for handling and storage of irradiated and spent fuel generated in factory-fabricated transportable micro-reactors 124

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Decommissioning process and decommissioning funding assurance

• Factory-fabricated transportable micro-reactor deployment models might involve transporting a reactor away from the deployment site to a facility at a different location for decommissioning at the end of its life or for refurbishment and refueling before re-deployment

• Depending on the activities to be conducted at a decommissioning facility or a refurbishment and refueling facility, the facility may need to be licensed under a combination of the regulations in 10 CFR Part 30 for byproduct material, Part 50 or 52 for a facility operating license or combined license, Part 70 for special nuclear material, and Part 72 for spent fuel storage

• The deployment site licensee would need to establish decommissioning funding assurance that considers the cost of removing the reactor from the site and decommissioning it elsewhere in addition to the cost of decommissioning activities at the deployment site. The NRC staff may consider site-specific decommissioning cost estimates that appropriately account for all activities at both locations and all waste disposal costs 125

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Siting in densely populated areas

• Some micro-reactor license applicants might seek to site reactors at locations that are inconsistent with the current Commission policy and the regulations in 10 CFR 100.21(b), i.e., a location within a population center of 25,000 residents or more

• The NRC staff is currently revising the population-related siting guidance in Regulatory Guide (RG) 4.7, General Site Suitability Criteria for Nuclear Power Stations, Revision 3, issued March 2014 (ADAMS Accession No. ML12188A053) to provide technology-inclusive, risk-informed, and performance-based criteria to assess certain population-related issues in siting advanced reactors

• In the near term, the staff will continue its effort to revise RG 4.7 and will review license applications in accordance with current Commission policy that allows alternative population-related criteria but precludes siting a commercial power reactor, no matter the size or type of reactor, within a population center of 25,000 residents or more 126

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Commercial maritime applications

• The NRC staff is aware of growing interest in commercial maritime applications of factory-fabricated transportable micro-reactors and other reactor technologies for stationary power production, marine vessel propulsion, production of decarbonized fuels, and other uses

• Depending on the particular application, deployment of commercial maritime reactors could introduce a host of policy issues and legal matters, especially for nuclear propulsion in the international shipping industry

• The NRC staff will continue to engage with stakeholders and monitor developments related to commercial maritime applications and assess the need for future Commission direction 127

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Commercial space applications

• The NRC staff is aware that developers are considering space applications of factory-fabricated transportable micro-reactors. However, the NRC staff is not aware of any plans for fully commercial space applications

• In the case of a fully commercial space application of a factory-fabricated transportable micro-reactor, the NRCs established regulatory jurisdiction and licensing authority would cover the related terrestrial activities prior to launch activities, which would be under the authority of the Federal Aviation Administrations Office of Commercial Space Transportation (a part of the Department of Transportation)

• If developers engage the NRC staff on terrestrial activities related to commercial space applications of factory-fabricated transportable micro-reactors, the NRC staff intends to apply the established regulatory framework, as informed by the potential licensing approaches and strategies outlined in this presentation 128

Other Licensing and Deployment Topics and Potential Near-Term Strategies and Next Steps Commercial mobile applications

• Factory-fabricated transportable micro-reactor deployment models might include scenarios where the reactor would be operated on an as-needed, where-needed basis, such as for disaster relief or to meet temporary increases in demand

• The current regulatory framework for reactor licensing is not conducive to this deployment strategy because the regulations in 10 CFR Part 100 apply to every site at which a reactor may be operated, and NRCs implementation of the National Environmental Policy Act relies on performing an environmental review that contemplates a particular site

• The AEA and regulations in 10 CFR Parts 50 and 52 for licensing utilization facilities also require opportunities for public hearings before the Commission can issue a facility operating license or authorize operation under a combined license. These may take a minimum of several months to complete, limiting the ability to rapidly deploy a reactor to meet immediate, short-term needs

• The NRC staff will monitor developments in the commercial sector related to deployment models and the demand for commercial mobile micro-reactor licensing. The staff will assess the need for future Commission direction, rulemaking, and coordination with other Federal agencies in this area 129

Next Steps

• Publish a draft white paper to further stakeholder engagement in August 2023

• Develop a Commission paper on licensing and deployment considerations factory-fabricated transportable micro-reactors:

- Request Commission direction on regulatory approaches for loading fuel and operational testing at the factory

- Provide information on other topics, including the NRC staffs related near-term strategies and next steps 130

Discussion Items

• Are there other approaches that the NRC staff should consider for loading fuel and operational testing at the factory that would not involve rulemaking?

• Are there other near-term strategies the NRC staff should consider for the other identified topics?

• Other feedback or questions for the NRC staff.

131

Advanced Reactor Stakeholder Public Meeting Break Meeting will resume at 4:30 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 501 432 683#

Transportable Microreactor Package Approval Advanced Reactor Stakeholder Meeting July 20, 2023 Bernard White and Olivia Hunsberger Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission

Purpose

• Inform external stakeholders of the NRC activities related to transportable microreactors.

• Importance and benefit of timely pre-application engagements in the regulatory process.

134

Overview

• Package approval standards

• Package approval regulatory approaches

• Microreactor Package Approval

• Risk-Informed Methodology 135

Package Approval Standards

• Prescriptive performance-based regulations

• Tests and conditions

- Normal conditions of transport

- Hypothetical accident conditions

• Post-test criteria

- Criticality safety

- Maximum dose rates

- Containment criteria 136

• Normal conditions of transport (10 CFR 71.71 & 49 CFR 173.465)

- Hot and cold temperatures Package Tests

- Reduced and increased external pressure

- Vibration

- Water spray

- Free drop (1 foot)

- Penetration test

• Hypothetical accident conditions (10 CFR 71.73) foot drop test in most damaging orientation inch puncture test minute fire test

- Water immersion test 137

Post-Test Performance Criteria

• Criticality safety

- Single package (10 CFR 71.55)

- Array of packages (10 CFR 71.59)

• Maximum dose rates for normal transport (10 CFR 71.47 & 49 CFR 173.441)

• Additional criteria for Type B packages (10 CFR 71.51)

- Containment for normal conditions of transport and hypothetical accident conditions

- Maximum dose rates after hypothetical accident conditions 138

Package Approval Regulatory Approaches

• Final Draft: Micro-reactors Licensing Strategies (ML21328A819)

• 10 CFR 71.41(c)

- Limited to changes to tests for normal conditions of transport and hypothetical accident conditions

- No changes to acceptance criteria

- Shipper controls for equivalent level of safety

• 10 CFR 71.41(d)

- One-time shipment of large packages

- Special package authorization

- Equivalent level of safety

• Exemptions via 10 CFR 71.12 139

Microreactor Package Pre-Application Engagements

• Provide staff with knowledge on specific designs and technologies

• Enhances quality of applications

• Helps NRC to understand future needs and inform its budget

• Ensures applicants and regulator have shared understanding of

- the applicable requirements

- review approach and

- whether data gaps exist (e.g., testing) that need to be addressed, as these may be the critical path, impacting the overall schedule

• Allows NRC to plan for package reviews 140

Risk-Informed Methodology

• NRC is reviewing a risk-informed methodology (ML23066A201) for limited number of shipments for a single transportable microreactor.

• Staff can approve exemptions that meet criteria in §71.12, unless directed to send to Commission.

• Uncertainties on the use of risk-informed methodology:

- May only be used by one reactor vendor

- Number and type of exemptions requested for a transportable microreactor package approval

• Planning a Commission paper on risk-informed methodology.

• Significant number of transportable microreactor package approvals needing exemptions would likely warrant Commission direction.

141

Closing Remarks

• NRC is ready to review packages for transportable microreactors.

• NRC regulatory framework in 10 CFR Part 71 is adequate for approving transportable microreactors.

• Package approval method could be package/reactor dependent.

• NRC is aware of numerous transportable microreactor designs but has not had pre-application engagement on most of them.

142

Bernard White, Yoira Diaz-Sanabria, Chief Sr. Project Manager Storage and Transportation Licensing Branch Bernard.White@nrc.gov Yoira.Diaz-Sanabria@nrc.gov 301-415-6577 301-415-8064 143

Future Meeting Planning

• The next periodic stakeholder meetings in 2023 are scheduled for September 14, October 25, and December 7.

• If you have suggested topics, please reach out to Steve Lynch at Steven.Lynch@nrc.gov 144

How Did We Do?

• Click link to NRC public meeting information:

https://www.nrc.gov/pmns/mtg?do=details&Code=20230270

• Then, click link to NRC public feedback form:

145