Text

Public Stakeholder Meeting on Possible Regulatory Process Improvements for Non-Light Water Reactors, Slide Presentations
	ML18348B091
Person / Time
Issue date:	12/13/2018
From:	William Reckley; NRC/NRO/DSRA/ARPB
To:	;
	Reckley W, NRO/DSRA/ARPB, 415-7490
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	Download: ML18348B091 (106)
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Public Meeting on Possible Regulatory Process Improvements for Advanced Reactor Designs December 13, 2018 1

Telephone Bridge (888) 793-9929 Passcode: 1770692

Public Meeting

Telephone Bridge (888) 793-9929 Passcode: 1770692

Opportunities for public comments and questions at designated times

Meeting on Regulatory Basis for Possible Changes to Physical Security Requirements at 2:30 2

Introductions

Modeling & Simulation (NRC)

Interface Requirements for Staged Licensing (NIA)

Developer Priorities & HALEU (NIC)

Policy Issues, Industry Needs Assessment TRISO topical report Future Meetings Regulatory Basis Development for Possible Changes to Physical Security Requirements 3

Outline

DBE Confirmatory Analysis Code Suite for Non-LWRs (S. Bajorek)

MELCOR non-LWR ACTIVITIES (H. Esmaili)

Consequence Analysis (MACCS) Code Development Plan for Non-LWRs (J. Barr) 4 Modeling & Simulation

5 Break Meeting/Webinar will begin shortly Telephone Bridge (888) 793-9929 Passcode: 1770692

6

Nuclear Innovation Alliance

- Ashley Finan

- Establishing Interface Requirements in Support of Staged Licensing

7

Nuclear Industry Council

- David Blee, NIC

Developer Priorities

- Stephen Crowne, URENCO

Next Generation Nuclear Fuels

8 Lunch Meeting/Webinar will begin at 1:00pm Telephone Bridge (888) 793-9929 Passcode: 1770692

9 Implementation Action Plans Strategy 1 Knowledge, Skills and Capability Strategy 2 Computer Codes

& Review Tools Strategy 3 Flexible Review Processes Strategy 5 Policy and Key Technical Issues Strategy 6 Communication Strategy 4 Consensus Codes and Standards ONRL Molten Salt Reactor Training Knowledge Management Competency Modeling Regulatory Roadmap Prototype Guidance Non-LWR Design Criteria ASME BPVC Section III Division 5 ANS Standards 20.1, 20.2 30.2, 54.1 Non-LWR PRA Standard Siting near densely populated areas Insurance and Liability Consequence Based Security (SECY-18-0076)

NRC DOE Workshops Periodic Stakeholder Meetings NRC DOE GAIN MOU Identification &

Assessment of Available Codes International Coordination Licensing Modernization Project Functional Containment (SECY-18-0096)

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

Environmental Reviews Potential First Movers Micro-Reactors Updated HTGR and Fast Reactor Training

- Completed

10 NRC Status 1.

Staff Training 2.

Computer Code Assessments 3.

Interactions with Licensing Modernization Project (DG 1353)

Environmental Review Working Group Update Roadmap 4.

ASME Div 5, ANS Design Standards, non-LWR PRA Standard 5.

Policy Issues Siting, PAA, Security, EP, Functional Containment 6.

Communications 7.

Micro-Reactors

11 Policy Table Ongoing Activities 1

Prototype Guidance Staged Licensing Roadmap (plan to update) 2a Source Term Prepare MST Guidance Dose Calcs Siting Prepare Siting Guidance 2b SSC Design Issues NEI 18-04, DG-1353 3

Offsite EP SECY-18-103 4

Insurance/Liability Future (2021) Report to Congress (no change acceptable) 5 PRA in licensing NEI 18-04, DG-1353 6

Defense in Depth NEI 18-04, DG-1353 7

Physical Security SECY-18-0076 (limited to sabotage)

12 Policy Table Ongoing Activities 8

LBEs NEI 18-04, DG-1353 9a Fuel Qualification technology specific 9b Materials Qualification technology specific 10a MC&A Cat II facilities ML18267A184 10b Security Cat II facilities ML18267A184 10c Collaboration criticality benchmark HALEU shipping 11 Functional Containment Performance Criteria SECY-18-0096 & SRM

?

Advanced Manufacturing

13 Policy Table Open - Not Working 1

Annual Fees 2

Manufacturing License 3

Process Heat 4

Waste Issues 5

Operator Staffing*

Remote/Autonomous

14 Policy Table No Plans (Resolved or Need Feedback) 1 Multi-module License 2

Operator Staffing*

3 Operational Programs 4

Module Installation 5

Decommissioning Funding 6

Aircraft Impact Assessments

15 NEI / ARRTF Updates

16 TRISO Topical Update

17 Future Meetings 2019 Tentative Schedule; Periodic Stakeholder Meetings February 7 Civil/Structural Design/Licensing Issues (e.g., seismic isolation)

March 28 May 9 June 27 August 15 October 10 December 11

18 Break Meeting/Webinar on Regulatory Basis for Possible Rulemaking on Physical Security will begin shortly Telephone Bridge (888) 793-9929 Passcode: 1770692

IAP Strategy 2: DBE Confirmatory Analysis Code Suite for Non-LWRs Stephen M. Bajorek, Ph.D.

Office of Nuclear Regulatory Research United States Nuclear Regulatory Commission Ph.: (301) 415-2345 / Stephen.Bajorek@nrc.gov Advanced Reactor Stakeholder Meeting December 13, 2018 RES Implementation Action Plan for Advanced Non-LWR ; Codes and Tools

Slide 2 Strategy 2 Codes for Design Basis Events 2

Numerous options available for thermal-hydraulics, neutronics, and fuel performance analysis for non-LWRs.

Evaluation of codes for NRC use began with gaining a better understanding of the technologies. Existing PIRTs were augmented by new PIRTs developed for molten-salt reactors.

Hands-on training and experience in DOE codes by NRC staff.

Slide 3 Strategy 2 Codes for Design Basis Events 3

Codes considered:

- NRC legacy codes (TRACE, PARCS, FRAPCON, FAST)

- DOE NEAMS codes (MAMMOTH, PRONGHORN, RELAP7)

- ANL codes (SAS4A/SASSY, SAM, PROTEUS, MC2, Nek5000)

- DOE CASL codes (MPACT, CTF, BISON, MAMBA)

- Commercial codes (FLUENT, COMSOL)

Recommended approach is to use a system of coupled codes, Comprehensive Reactor Analysis Bundle (CRAB). This includes codes from the NRC and DOE.

Slide 4 TRACE System T/H MOOSE PARCS Neutronics SCALE Cross-sections FAST Fuel Performance BISON Fuel Performance PRONGHORN Core T/H SAM System and Core T/H Nek5000 CFD MELCOR Containment / FP DOE Code NRC Code MAMMOTH Neutronics Comprehensive Reactor Analysis Bundle (CRAB) Current View; Oct.2018 SERPENT Cross-sections SERPENT Cross-sections MAMMOTH Neutronics Intl Code FLUENT CFD Commercial

Slide 5 5

Code Selection Considerations Physics. Code suite must now or with development capture the correct physics to simulate non-LWRs. Selection of codes based on results of PIRTs. Code coupling necessary for multi-physics.

Flexibility. Multiple reactor design concepts require flexibility within code suite. A goal has been to limit the number of new codes and need for staff training.

Code V&V. Code assessment is critical, especially assessment relative to non-LWRs.

Computation Requirements. Must be able to run simulations on HPC platforms available to NRC.

Cost avoidance. An objective is to minimize cost to the NRC by leveraging DOE tools and influencing development plans.

Codes selected for CRAB satisfy these criteria.

Slide 6 DBE Analysis Codes

Code Suite Report (draft) describes analysis approach for each of 10 distinct design types.

- Gaps

- Assessment

- Tasks

Reference plant models being developed.

Slide 7 TRACE System T/H MOOSE PARCS Neutronics SCALE Cross-sections FAST Fuel Performance BISON Fuel Performance PRONGHORN Core T/H SAM System and Core T/H Nek5000 CFD MELCOR Containment / FP DOE Code NRC Code MAMMOTH Neutronics Comprehensive Reactor Analysis Bundle (CRAB for LWRs)

SERPENT Cross-sections SERPENT Cross-sections MAMMOTH Neutronics Intl Code FLUENT CFD Commercial

Slide 8 TRACE System T/H MOOSE PARCS Neutronics SCALE Cross-sections FAST Fuel Performance BISON Fuel Performance PRONGHORN Core T/H SAM System and Core T/H Nek5000 CFD MELCOR Containment / FP DOE Code NRC Code MAMMOTH Neutronics Comprehensive Reactor Analysis Bundle (CRAB for LWRs w/ATF)

SERPENT Cross-sections SERPENT Cross-sections MAMMOTH Neutronics Intl Code FLUENT CFD Commercial

Slide 9 TRACE System T/H MOOSE PARCS Neutronics SCALE Cross-sections FAST Fuel Performance BISON Fuel Performance PRONGHORN Core T/H SAM System and Core T/H Nek5000 CFD MELCOR Containment / FP DOE Code NRC Code MAMMOTH Neutronics Comprehensive Reactor Analysis Bundle (CRAB for GCRs)

SERPENT Cross-sections SERPENT Cross-sections MAMMOTH Neutronics Intl Code FLUENT CFD Commercial

Slide 10 TRACE System T/H MOOSE PARCS Neutronics SCALE Cross-sections FAST Fuel Performance BISON Fuel Performance PRONGHORN Core T/H SAM System and Core T/H Nek5000 CFD MELCOR Containment / FP DOE Code NRC Code MAMMOTH Neutronics Comprehensive Reactor Analysis Bundle (CRAB for Heat Pipe Reactors)

SERPENT Cross-sections SERPENT Cross-sections MAMMOTH Neutronics Intl Code FLUENT CFD Commercial

Slide 11 11 Code Readiness

Using PCMM (Predictive Capability Maturity Model) to characterize code readiness.

- Geometric Fidelity

- Physics and Model Fidelity

- Code Verification

- Solution Verification

- Code Validation

- Uncertainty Quantification

Rating scale 0 to 3 D A

Slide 12 Summary & Conclusions Code Suite Report recommends the codes in CRAB as the approach for non-LWR analysis.

Using the combination of NRC and DOE codes will provide a technically superior product than can be attained with further development of the NRCs legacy LWR codes only.

Using the DOE codes provides a significant benefit in resources & schedule to the NRC. DOE has been cooperative in revising their plans to fit our needs and schedule.

MELCOR non-LWR ACTIVITIES Hossein Esmaili Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission December 13, 2018

2 MELCOR Overview State-of-the-art tool for severe accident progression and source term analysis. Ongoing development of new capabilities Replace collection of simple, special purpose codes, i.e., Source Term Code Package (STCP)

Eliminate tedious hand-coupling between modules Capture feedback effects (i.e., coupling of temperatures, release rates, and decay heating)

MELCOR developed at Sandia National Laboratories for the U.S. NRC

MELCOR Code Development 3

Fully Integrated, engineering-level code Thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; Core heat-up, degradation, and relocation; Core-concrete attack; Hydrogen production, transport, and combustion; Fission product release and transport behavior Traditional Application User constructs models from basic constructs Control volumes, flow paths, heat structures, Multiple CORE designs PWR, BWR, HTGR (Pebble Bed & PMR), PWR-SFP, BWR-SFP, SMR, Sodium (Containment)

Adaptability to new reactor designs Validated physical models ISPs, benchmarks, experiments, accidents Uncertainty Analysis Relatively fast-running Characterized numerical variance User Convenience Windows/Linux versions Utilities for constructing input decks (GUI)

Capabilities for post-processing, visualization Extensive documentation Non-LWR Reactors HTGR/SFR/MSR Code Development & Regulatory Applications International Collaboration (CSARP/MCAP/EMUG/AMUG)

Integrated models required for self-consistent analysis

Development of evaluation models (example HTGR)

ACRS Future Plant Designs Subcommittee, April 5, 2011 4

Non-LWR Beyond Design Basis Events

5 SCALE Code & Application to MELCOR/MACCS Oak Ridge Isotope Generation code (ORIGEN)

Irradiation and decay simulation code Fuel depletion and used fuel characterization Source terms for accident analyses (operating reactors, spent fuel handling, storage, etc.)

Structural material activation (in-core, ex-core)

Material feed and removal for fuel cycle and liquid fuel ORIGEN data enable comprehensive isotopic characterization of fuel over a large time scale, including repository analysis ORIGEN / ORIGAMI Depletion, activation and decay Reactor-specific radioactive source term characterization AMPX Validated cross section libraries; depletion and decay data TRITON / Polaris Transport and depletion in 1D, 2D, and 3D for LWR, ATF, and nonLWR ENDF/B Physics data Thermal scattering law, resonance data, energy distributions, fission yields, decay constants, etc.

High Temperature Gas Cooled Reactors 6

Helium Properties

Accelerated steady-state initialization

Two-sided reflector (RF) component

Modified clad (CL) component (PMR/PBR)

Core conduction

Point kinetics

Fission product diffusion, transport, and release

TRISO fuel failure

Graphite dust transport

Turbulent deposition, Resuspension

Basic balance-of-plant models (Turbomachinery, Heat exchangers)

Momentum exchange between adjacent flow paths (lock-exchange air ingress)

Graphite oxidation Modeling Gaps Existing Modeling Capabilities

Current modeling uses UO2 material properties, needs to be extended to UCO

Molten Salt Reactors 8

Properties for LiF-BeF2 have been added Equation of State Current capability Thermal-mechanical properties Current capability EOS for other molten salt fluids would need to be developed Minor modeling gap Fission product modeling Fission product interaction with coolant, speciation, vaporization, and chemistry Moderate modeling gap Two reactor types envisioned Fixed fuel geometry TRISO fuel models Current capability Liquid fuel geometry MELCOR CVH/RN package can model flow of coolant and advection of internal heat source with minimal changes.

Current capability COR package representation no longer applicable but structures can be represented by HS package Calculation of neutronics kinetics for flowing fuel Significant modeling gap.

Sodium Fast Reactors 7

Sodium Properties Sodium Equation of State Sodium Thermo-mechanical properties Containment Modeling Sodium pool fire model Sodium spray fire model Atmospheric chemistry model Sodium-concrete interaction model SFR Core modeling Fuel thermal-mechanical properties Fuel fission product release and transport FP speciation & chemistry Bubble transport through a sodium pool Core degradation models SASS4A surrogate model Heat pipe specific models Containment Modeling Capability for having more than one working fluid Vaporization rates of RNs from sodium pool surface Radionuclide entrainment near pool surface during fires Transport of FP in sodium drops Hot gas layer formation during sodium fires.

Oxygen entrainment into a pool fire Sodium water reactions Modeling Gaps Existing Modeling Capabilities

9 Design Basis Source Term Development Process (example: MOX & High Burnup Fuel) 9 Fission Product Transport MELCOR Oxidation/Gas Generation Experimental Basis Melt Progression Fission Product Release PIRT process Accident Analysis Design Basis Source Term Scenario # 1 Scenario # 2 Synthesize timings and release fractions Cs Diffusivity Similar RFs to NUREG-1465 but prolonged release Differences not from change of fuel but from code advances Scenario # n-1 Scenario # n Powers, et al. Accident Source Terms for Light Water Nuclear Power Plants Using High-Burnup or MOX Fuel, SAND2011-0128 January 2011

Consequence Analysis (MACCS)

Code Development Plan for Non-LWRs Jonathan Barr Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission December 13, 2018

2 MACCS Overview MACCS is the only code used in U.S. for probabilistic offsite consequence analysis Treats all technical elements of Level 3 PRA standard: radionuclide release, atmospheric transport, meteorology, protective actions, site data, dosimetry, health effects, economic factors, uncertainty MACCS Gaussian plume segment ATD model animation for a single weather trial

3 MACCS Overview Highly flexible code enabling applicability to different types of sources and accidents Variety of associated risk measures

- Dose

- Radiological health effects and fatality risk

- Economic impact

- Land contamination

- Population affected by protective actions Developed by NRC over 3+ decades MACCS recently has been used in major studies including State-of-the-Art Reactor Consequence Analyses (SOARCA), Level 3 PRA project, and various Fukushima-related applications Part of Cooperative Severe Accident Research Program (CSARP) with 28 member countries

4 MACCS Applications Regulatory cost-benefit analysis Environmental report analyses of Severe Accident Mitigation Alternatives (SAMA) and Design Alternatives (SAMDA)

Level 3 PRA Research studies of accident consequences Support for emergency preparedness Dose-distance evaluations for emergency planning

5 MACCS for Non-LWRs

Code development plans for design-specific issues

- Radionuclide screening

- Radionuclide size

- Radionuclide chemical form

- Radionuclide shape factor

- Tritium

Code development plans for site-related issues

- Near-field atmospheric transport

- Decontamination modeling

6 Near-Field Atmospheric Transport MACCS currently has a simple model for building wake effects; user guide cautions against use closer than 500m Non-LWRs (and SMRs) desire smaller EPZ and site boundary than large LWRs; therefore desire better modeling of near-field phenomena Lloyd L. Schulman, David G. Strimaitis & Joseph S. Scire (2000) Development and Evaluation of the PRIME Plume Rise and Building Downwash Model, Journal of the Air & Waste Management Association, 50:3, 378-390 Wind tunnel simulation of streamlines near a cubic building

7 Near-Field Atmospheric Transport Various options for addressing near-field ATD Modifications to Gaussian plume segment ATD model CFD modeling of 3-d wind field with Lagrangian particle tracking ATD model Empirical models of 3-d wind fields with Lagrangian particle tracking ATD model Considerations for evaluating options Extent of practical acceptance in the user community Simplicity of use Computational efficiency Cost and time efficiency Accuracy Feasibility for probabilistic application QUIC Factsheet, Los Alamos National Laboratory Example QUIC-URB simulation of wind vectors Example QUIC-PLUME simulation of urban transport and dispersion

Establishing Interface Requirements in Support of Staged Licensing December 13, 2018 Ashley Finan ashley@nuclearinnovationalliance.org

Background Documents 10 CFR Part 52, Subpart E allows an applicant to seek standard design approval for either an entire plant or major portions thereof NRC document: A Regulatory Review Roadmap for Non-Light Water Reactors (ML17312B567)

NIA report: Clarifying Major Portions of a Reactor Design in Support of a Standard Design Approval (ML17128A507)

NRC staff provided feedback on this report on July 20, 2017 (ML17201Q109) 2

NIA Draft Report: Establishing Interface Requirements in Support of Staged Licensing 3

Table of Contents:

Executive Summary Introduction Purpose and Scope Standard Design Approval Methods to Develop Interface Requirements Example Cases Core Design Reactor Vessel Auxiliary Cooling System Design Reactor Coolant System Piping Design Reactor Building Structural Design Conclusions

Introduction

Many companies are developing new designs with new safety approaches

Some companies are using predominantly private funding, and thus confront different investment requirements from historic projects

Companies will take a variety of licensing approaches appropriate to their business plan 4

5 Figure 1: Current Project Risk/Investment Profile Relative to Detailed Design & Licensing Figure 2: Desirable Project Risk/Investment Profile Relative to Detailed Design &

Licensing

Staged Licensing Review Approach

Some companies may opt for a staged review approach using any of:

- Licensing project plan or regulatory engagement plan

- Preliminary design reviews

- Topical and/or technical reports

- Standard design approval

- Construction permit or design certification 6

Purpose and Scope

Provide guidance to vendors using the SDA on the establishment of interface requirements between portions of a design in the SDA with those that will be submitted at a later date

Any reactor type 7

Standard Design Approval 10 CFR Part 52 Subpart E

- Documents staff findings, involves ACRS reviews, provides reference for subsequent applications

- Incremental progress towards licensing or certification as part of staged licensing Potential value:

- Licensing risk reduction (via approval of limited portion of design)

- Reduce initial development cost (defer portions to subsequent licensing steps)

- Approval for portion as part of commercial strategy, e.g.:

Optional design features such as power uprate or non-electric application

Deployment outside US May result in greater overall cost/timeline compared with single successful application 8

Methods to Develop Interface Requirements

Have approved QA program

Clearly define scope of SDA

- SSCs, engineering disciplines, technical bases for satisfying principal design criteria (PDC)

PDC could be derived from Reg Guide 1.232, for example, or the LMP guidelines.

Set boundary conditions with functional and operational characteristics of SSCs that are not within scope

- These will have to be satisfied in subsequent submittals, if full design approval is sought

- Margins are required; size of margins may impact economics 9

Process for Developing Interface Requirements in Support of an SDA 10

Example Cases Core Design Reactor Vessel Auxiliary Cooling System Design Reactor Coolant System Piping Design Reactor Building Structural Design

Tables delineate interface requirements of the SDA example and are organized by ARDC 11

Example: RVAC System Interface Requirements Quality standards and records Design basis for protection against natural phenomena Fire protection Environmental and dynamic effects design bases Instrumentation and control Containment design Protection system functions Residual heat removal Emergency core cooling Containment heat removal Inspection of containment heat removal system Testing of containment heat removal system Containment design basis 12

13 ARDC Title Sample Interface Requirements for RVAC System 2

Design basis for protection against natural phenomena Interface Requirement The ability of the SSCs of the RVAC to withstand the design basis natural phenomena will be addressed in the FSAR. The comparison of the FSAR design assumptions to those relating to an actual site will be addressed in a future submittal. Adequate margin should be included in the assumed values for the natural phenomena to provide flexibility in siting the design.

The FSAR will specify seismic, hurricane, and tornado design parameters (e.g., earthquake design response spectra, soil conditions, tornado and hurricane wind speeds, etc.). These parameters will be compared to those evaluated for a future site.

3 Fire protection Interface Requirement The RVAC is required to have a fire protection program. The fire protection program will be addressed in a future submittal.

The FSAR will include a commitment that the materials used in the RVAC structure will use noncombustible and fire-resistant materials wherever practical, particularly in locations with SSCs important to safety.

Next Steps

Q&A today

Feedback factored into revised report

NRC Feedback Thank you!

14

Thank you Feedback & Questions Please feel welcome to send additional input at any time to Ashley Finan (ashley@nuclearinnovationalliance.org).

Priorities for Advanced Reactor Developers:

USNIC Survey of Developer Priorities David Blee President & CEO U.S. Nuclear Industry Council Hon. Jeffrey S. Merrifield Former Commissioner, USNRC; Chairman, USNIC Advanced Reactors Task Force; Partner, Pillsbury Winthrop Shaw Pittman December 13, 2018

USNIC AR Developers Survey USNIC conducted a third in a series survey of 16 leading U.S. Advanced Reactor technology developers with regard to DOE Initiatives 15 Developers responded, one respondent per company This was a blind survey so individual results were not identified 2

Survey Goals Intended to provide stakeholder feedback on NRC preparations for Advanced Reactor Licensing Feedback is intended to give constructive input to the Commission and Staff Survey provides a snapshot of the current policy priorities of the Advanced Reactor Community Assessment goes beyond the efforts of the Office of New Reactors to include the preparations of other NRC offices Provides feedback on the perceived technical readiness of the NRC staff 3

Q1: Pace of the NRCs Advanced Reactor Licensing Transformation: Rate the pace of the NRCs Preparation Activities for Advanced Reactor licensing?

4

Q2: NRC Support for Advanced Reactor Licensing Transformation: Please rank the NRC Offices' prioritization of Advanced Reactor transformation?

5 NRC Chairman & Commissioners Office of New Reactors Office of Nuclear Material Safety and Safeguards Office of Nuclear Security and Incident Response

Q3: Planning Timeframe for Licensing Application Submittals: What should the NRC and DOEs Planning Timeframe be for new Advanced Reactor License Applications?

6

Q4: Focus for NRC Advanced Reactors Licensing Transformation in 2019: What should the NRCs key Licensing Transformation Focus be in? (ranked) 7

Q5: Early Resolution of NRC Policy Issues (e.g. emergency preparedness, consequence-based physical security): How do you think the NRC is doing with respect to resolving Key Policy issues early?

8

Q6: Enhanced Pre-Licensing Engagement: What actions would most improve the NRCs pre-licensing engagement (rank in order of priority)?

9 Cost-share for pre-licensing Fixed price and schedule certainty for pre-licensing Enhanced NRC Advanced Reactor Technology capability More robust stakeholder engagement Additional involvement by the Office of New Reactors Additional involvement by the Office of Nuclear Security & Incident Response Additional involvement by the Office of Nuclear Material, Safety & Safeguards

Q7: NRC Advanced Reactors Technical Capability:

Please rate the NRCs Advanced Reactor technology technical capability?

10

Q8: Confidence in NRC Advanced Reactors Licensing Schedule and Cost: What is your confidence that the NRC can transform its licensing process to provide greater schedule and cost certainty?

11

Q9: Should the NRC be doing more to seek non-fee based funding?

12

Q10: Value of NRC Advanced Reactor Stakeholder Meetings:

Are the NRCs Stakeholder Meetings (held every 6-8 weeks)?

13

Q11: Do you believe the NRC Office of Research is putting sufficient time and resources towards Advanced Reactor development?

14

Q12: Versatile Advanced Test Reactor: How important is the deployment of a new U.S. Department of Energy advanced test reactor (Versatile Test Reactor) by 2026?

15

Summary Results Commission and staff of Office of New Reactors are perceived as making progress on Advanced Reactor policy decisions and licensing readiness Office of Nuclear Materials Safety and Safeguards and to a somewhat lesser extent the Office of Nuclear Security and Incident Response are not perceived as having the same level of engagement on Advanced Reactor issues Agency readiness for High Temperature Reactors is very good Higher level of questioning about NRC readiness to license Molten Salt, Fast and Liquid Metal Reactors There is a lack of understanding of what the Office of Research is doing to assist in preparing the NRC for Advanced Reactors There was an overwhelming view that the Commission needs to do more to assist in lifting the burden of Fee Based programs on Advanced Reactors 16

The United States Nuclear Industry Council (USNIC) is the leading U.S.

business consortium advocate for nuclear energy and promotion of the American supply chain globally. Composed of over 80 companies USNIC represents the "Who's Who" of the nuclear supply chain community, including key utility movers, technology developers, construction engineers, manufacturers and service providers. USNIC encompasses eight working groups and select task forces. For more information visit www.usnic.org U.S. Nuclear Industry Council 1317 F Street, NW - Washington, DC 20004 (202) 332-8155 www.usnic.org 17

The Nuclear Institute: Advance Nuclear Technologies 1

LWR Fuels LEU-UO2-ZircAlloy 2

DOE Programme Accident Tolerant Fuel Site Licensing Cat-II Facility Operational Criticality & Safety Intrinsically Safe Fuels National Regulator(s)

Deconversion/H2M U-Metal U-Oxides U-Salts Storage & Transport Cylinders Overpacks Class 7 Shipping Insurance Fuel Fabrication Higher Enrichment HA-LEU ~19.75%

Enrichment LEU+ Plus (5~10%)

Test & Research Reactors Molten Salt Reactors Lead Cooled Reactors Fast Breeder Reactors Sodium Fast Reactors HTGR Gen-III Reactor Uprates SMRs Micro-SMRs TRISO Fuel UCO U02 Uranium Nitride Uranium Silicide ATF High Density Fuel Pellets U-Silicide U-Nitride Chromium doped U02 FCM Ceramic Fuel Fabricated TRISO Prismatic Block Pebble Bed ATF Cladding Systems Chromium coating Silicon-carbide cladding Metallic Fuel Lightbridge Zr-U Alloy U-Molybdenum Liquid Fuels Molten salts Aqueous uranyl salt solutions RepU Existing UO2 Fuel Pellets

The Nuclear Institute: Advance Nuclear Technologies 3

UF6

<5%

LEU UO2 Spent Fuel LWR Reactors UO2 / ZircAlloy Fuels Fabrication 0.711%

UF6

<5%

LEU Next Generation Fuels TRISO (UCO),

Uranium Nitride, Uranium Silicide, U-metal Alloys UF4 Salts etc Gen III+, ATFs SMRs, GenIV, Advanced Reactors Research & Test Reactors 5%-20%

HA-LEU U-metal U-oxide U-salts Completing the Future Nuclear Fuel Supply Chain Enrichment Higher Enrichment Deconversion

The Nuclear Institute: Advance Nuclear Technologies 4

A broad community of users may benefit from HA-LEU:

Research & Test Reactors Operators of existing LWRs seeking improvements in fuel reliability and economics through higher burnup and extended operating cycles Accident Tolerant Fuels Gen IV and other Advanced reactor designs Advanced fuel designs Producers of targets for medical isotope production Fuel solutions are needed across the full span of HALEU enrichments some clumping may develop in the ranges of 6.0%-8.0% U235 and 13.0-16.0%

U235 and at 19.75% U235.

The Nuclear Institute: Advance Nuclear Technologies 5

1. A Higher Enrichment Facility to produce HA-LEU enrichments:

- the material will be in the form of uranium hexafluoride (UF6)

2. A conversion facility to (de)convert HA-LEU UF6 into metal, oxide and/or salts

3. One or more fabrication facilities that can manufacture the specific fuel types required by the various reactor and fuel designs Packaging and transportation solutions are needed between each of these processing steps and to the ultimate user Spent fuel packaging will also need to be considered at the back-end of the fuel cycle

The Nuclear Institute: Advance Nuclear Technologies 6

Are moderator exclusion requirements met through the cylinder or through an overpack?

Criticality benchmarking data is needed for HA-LEU assays.

Cylinder Model Diameter (inches / mm)

Maximum Enrichment Maximum UF6 (lbs / kgs) 1S 1.5 / 38.1 100.00%

1.0 / 0.5 2S 3.5 / 88.9 100.00%

4.9 / 2.2 5B 5.0 / 127 100.00%

54.9 / 24.9 8A 8.0 / 203.2 12.5%

255 / 115.7 30B 30 / 762 5%

5020 / 2277 Existing UF6 Cylinders for Higher Assays (ANSI N14.1)

The Nuclear Institute: Advance Nuclear Technologies 7

Co-location of Enrichment & Deconversion Problem:

There is currently no available transport package for HA-LEU.

Possible Solution: 2-Box Model: Co-location of Higher Enrichment and Deconversion Facilities.

<5% UF6 0.711% ENU

<19.99% UF6 UF6 Deconversion Facility

<19.99%

U-metal U-oxide U-salts Next Generation Fuel Manufacturing Facility Fabricated HA-LEU Fuels TRISO (UCO)

U02 U-metal Alloys UF4 Salts Uranium Nitride Uranium Silicide (Cat 2 License)

Higher Enrichment Facility

Reduces expense and time required to develop packaging and transport solutions

Can be expanded to include fabrication facilities

Satisfying the requirements of a number next generation fuel types for HA-LEU.

Leverages existing site characterization data, site infrastructure, and regulator familiarity

The Nuclear Institute: Advance Nuclear Technologies 8

UUSA safety basis is analyzed at 6%, UUSA would need to demonstrate the reduction in the margin of safety to increase enrichment level limit.

- Could be done quickly 1b. Enrichments above 5.5%

UUSA would need to reanalyze the design safety basis at higher enrichments

- Analysis would require additional resources and will take more time.

CAT 2 - Changes to FNMCP and Security Plan Level of effort required to achieve 19.75% limit vs. 7.0% limit is not that great.

2a. Utilizing existing transport packages for UF6 above 5%

Criticality benchmarking data is needed for HA-LEU assays For use with UO2 fuel pellets 2b. UF6 deconversion For other fuel types If existing transport packages are not approved at higher enrichments

The Nuclear Institute: Advance Nuclear Technologies 9

NRC resources and priorities-due to the reductions in licensing staff at the NRC, the ability to review a license amendment in a timely manner is a concern. NRC should prioritize appropriately.

2. Key rulemaking activities Part 50.68 change to support power industry Part 171 Fees - new category for combined fuel cycle facility Part 171 Fees - new category for moderate strategic SNM facility Part 73 - highly diluted category 3.

NRC must resist the temptation to revisit issues they want to change but are not required to raise enrichment limits. If analytical models are approved for licensees, there is no need to change.

4.

Analytical codes are well validated up to 6%. Would need additional validation beyond 6%.

It is imperative that the enrichment, conversion and fabrication facilities - and the concordant packaging solutions - be developed on concurrent schedules.

2.

The licensing framework needs to support development of a HA-LEU fuel cycle and regulator resources are needed.

3.

Companies making investments in HA-LEU facilities need to be sufficiently assured of an economic return.

4.

URENCO USA could submit a License Amendment Request (LAR) for 5.5% enrichment limit by April 30, 2019. A 6% LAR could be ready by June 30, 2019.

5.

We all must hold hands and jump together!

SECY-18-0076 OPTIONS AND RECOMMENDATION FOR PHYSICAL SECURITY FOR ADVANCED REACTORS December 13, 2018 1

2 Background

NRC Advanced Reactor Policy Statement - Attributes:

Highly reliable and less complex decay heat removal systems; Longer time constants to reaching safety system challenges; Simplified safety systems that reduce required operator actions; Designs that minimize the potential for severe accidents and their consequences; and Designs that incorporate the defense-in-depth philosophy by maintaining multiple barriers against radiation release

3 Background

NRC Advanced Reactor Policy Statement Designs that include considerations for safety and security requirements together in the design process such that security issues (e.g., newly identified threats of terrorist attacks) can be effectively resolved through facility design and engineered security features, and formulation of mitigation measures, with reduced reliance on human actions.

Challenge is to address policy issues related to how safety and security requirements for advanced reactors should reflect inherent design characteristics such as longer time constants before degradation of barriers and release of radioactive material given a loss of safety functions.

4 Background

SECY-11-0184, Security Regulatory Framework for Certifying, Approving, and Licensing Small Modular Reactors.

o The staffs assessment determined that the current security regulatory framework is adequate to certify, approve, and license iPWRs o

The current regulations allow SMR designers and potential applicants to propose alternative methods or approaches to meet the performance-based and prescriptive security and MC&A requirements.

Alternate Measures (10 CFR 73.55(r))

License Conditions

Exemptions The question at hand is whether some type of generic regulatory action would be preferable to the case-by-case approach described in SECY-11-0184.

5 SECY-18-0076 Options Identifies 4 Options:

1)

No change / Status quo 2)

Address possible requests for alternatives via guidance 3)

Limited scope rulemaking to address what would otherwise be likely requests for alternatives 4)

Broader based rulemaking to more fully reflect attributes of advanced reactors

6 Option 3 - Limited Scope Rulemaking Revise specific regulations and guidance related to physical security for SMRs and non-LWRs through rulemaking.

o Example - NEI proposal for reductions in the number of armed responders (10 CFR 73.55(k)(5))

NRC staff would interact with stakeholders to identify specific requirements within existing regulations that may play a diminished role in providing physical security for SMRs and non-LWRs while contributing significantly to capital or operating costs.

NRC staff would develop guidance documents to support the implementation of the requirements defined through the rulemaking.

7 Staff Requirements Memorandum (SRM)

SRM Dated November 19, 2018 The Commission approved the staffs recommended Option 3, to initiate a limited-scope revision of regulations and guidance related to physical security for advanced reactors and approved the enclosed rulemaking plan, subject to the enclosed edits.

Complete regulatory basis 12 months following Commissions SRM

Another potential area is the prescriptive requirements in 10 CFR 73.55 for onsite secondary alarm stations.

8 Rulemaking Process

9 Barrier Assessment (Bow Tie Diagram)

Note that top level event generally aligns with security concerns for radiological sabotage; a rulemaking, if pursued, would also need to address threats related to theft/diversion

10 Revisit First Principles

11 NEI Proposed Logic for Applicability of Alternate Regulations (Armed Responders Not Required)

Possible Performance (Consequence)

Based Approach

12 Security Design Considerations Preliminary Draft Guidance (March 2017)

Intrusion Detection Systems Intrusion Assessment Systems Security Communication Systems Security Delay Systems Security Response Control Measures for land/waterborne vehicle bombs Access Control Portals Cyber Security

13 Discussion Potential Scope of Alternative Requirements 10 CFR 73.55(k) - armed responders 10 CFR 73.55(i) - secondary alarm stations

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14 Stakeholder Presentation/Discussion NEI

15 Discussion Stakeholder Presentation/Discussion USUCS

16 General Discussion Public Questions/Feedback

ML18348B091

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Introductions

2 Background

3 Background

4 Background

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