ML18348B091

12/13/18 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
References
Download: ML18348B091 (106)

Text

Public Meeting on Possible Regulatory Process Improvements for Advanced Reactor DesignsDecember 13, 2018 1Telephone 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

IntroductionsModeling & Simulation (NRC)Interface Requirements for Staged Licensing (NIA)Developer Priorities & HALEU (NIC)Policy Issues, Industry Needs AssessmentTRISO topical reportFuture MeetingsRegulatory Basis Development for Possible Changes to Physical Security Requirements 3Outline 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)4Modeling & Simulation 5BreakMeeting/Webinar will begin shortlyTelephone 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 8LunchMeeting/Webinar will begin at 1:00pm Telephone Bridge(888) 793-9929 Passcode:

1770692 9Implementation Action PlansStrategy 1Knowledge, Skills and CapabilityStrategy 2Computer Codes & Review ToolsStrategy 3Flexible Review ProcessesStrategy 5Policy and Key Technical IssuesStrategy 6CommunicationStrategy 4Consensus Codes and StandardsONRL Molten Salt Reactor TrainingKnowledge ManagementCompetency ModelingRegulatory RoadmapPrototype Guidance Non-LWR Design CriteriaASME BPVC Section III Division 5ANS Standards20.1, 20.230.2, 54.1Non-LWRPRA StandardSiting near densely populated areasInsurance and LiabilityConsequence Based Security(SECY-18-0076)NRC DOE WorkshopsPeriodic Stakeholder MeetingsNRC DOE GAIN MOUIdentification & Assessment of Available CodesInternational CoordinationLicensing ModernizationProjectFunctional Containment (SECY-18-0096)EP for SMRs and ONTs(SECY-18-0103)EnvironmentalReviewsPotential First MoversMicro-ReactorsUpdated HTGR and Fast Reactor Training-Completed 10NRC Status 1.Staff Training 2.Computer Code Assessments 3.Interactions with Licensing Modernization Project (DG 1353)Environmental Review Working GroupUpdate Roadmap 4.ASME Div5, ANS Design Standards, non

-LWR PRA Standard 5.Policy IssuesSiting, PAA, Security, EP, Functional Containment 6.Communications 7."Micro-Reactors" 11Policy TableOngoing Activities 1Prototype GuidanceStaged LicensingRoadmap(plan to update) 2aSource TermPrepare MST GuidanceDose CalcsSitingPrepare Siting Guidance 2bSSC Design IssuesNEI 18-04, DG-1353 3Offsite EPSECY-18-103 4Insurance/LiabilityFuture (2021) Report to Congress (no change acceptable

)5PRA in licensingNEI 18-04, DG-1353 6Defense in DepthNEI 18-04, DG-1353 7PhysicalSecuritySECY-18-0076 (limitedto sabotage

)

12Policy TableOngoing Activities 8LBEsNEI 18-04, DG-1353 9aFuel Qualificationtechnology specific 9bMaterials Qualificationtechnology specific 10aMC&A Cat II facilitiesML18267A184 10bSecurity Cat II facilitiesML18267A184 10cCollaboration

• criticality benchmark

• HALEU shipping 11Functional Containment PerformanceCriteriaSECY-18-0096 & SRM

?Advanced Manufacturing 13Policy TableOpen -Not Working 1Annual Fees 2Manufacturing License 3Process Heat 4Waste Issues 5Operator Staffing*Remote/Autonomous 14Policy TableNo Plans(Resolved or Need Feedback) 1Multi-moduleLicense 2Operator Staffing*

3Operational Programs 4ModuleInstallation 5Decommissioning Funding 6Aircraft Impact Assessments 15NEI / ARRTF Updates 16TRISO Topical Update 17Future Meetings2019 Tentative Schedule; Periodic Stakeholder MeetingsFebruary 7Civil/StructuralDesign/Licensing Issues(e.g., seismic isolation)March 28May 9June 27August 15October 10December 11 18BreakMeeting/Webinar on Regulatory Basis for Possible Rulemaking on Physical Security will begin shortlyTelephone Bridge(888) 793-9929 Passcode: 1770692 IAP Strategy 2: DBE Confirmatory Analysis Code Suite for Non

-LWRsStephen M. Bajorek, Ph.D.Office of Nuclear Regulatory ResearchUnited States Nuclear Regulatory CommissionPh.: (301) 415

-2345 / Stephen.Bajorek@nrc.govAdvanced Reactor Stakeholder MeetingDecember 13, 2018RES 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 4TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/H SAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB) Current View; Oct.2018 SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 5 5Code 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 6DBE 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 7TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/H SAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB for LWRs) SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 8TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/H SAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB for LWRs w/ATF) SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 9TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/H SAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB for GCRs) SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 10TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/H SAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB for Heat Pipe Reactors)SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 11 11Code 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 12Summary & 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 productthan can be attained with further development of the NRC's legacy LWR codes only. Using the DOE codes provides a significant benefit in resources & scheduleto the NRC. DOE has been cooperative in revising their plans to fit our needs and schedule.

MELCOR non

-LWR ACTIVITIESHossein EsmailiOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionDecember 13, 2018 2MELCOR 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/MSRCode Development & Regulatory ApplicationsInternational 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 4Non-LWR Beyond Design Basis Events 5SCALE 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 analysisORIGEN / ORIGAMIDepletion, activation and decayReactor-specific radioactive source term characterizationAMPXValidated cross section libraries; depletion and decay dataTRITON / PolarisTransport and depletion in 1D, 2D, and 3D for LWR, ATF, and nonLWRENDF/BPhysics dataThermal scattering law, resonance data, energy distributions, fission yields, decay constants, etc.

High Temperature Gas Cooled Reactors 6Helium PropertiesAccelerated steady-state initializationTwo-sided reflector (RF) component Modified clad (CL) component (PMR/PBR)Core conductionPoint kinetics Fission product diffusion, transport, and releaseTRISO fuel failureGraphite dust transportTurbulent deposition, ResuspensionBasic balance-of-plant models (Turbomachinery, Heat exchangers)Momentum exchange between adjacent flow paths (lock

-exchange air ingress) Graphite oxidationModeling GapsExisting Modeling CapabilitiesCurrent 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 neutronicskinetics 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 reactionsModeling GapsExisting Modeling Capabilities 9Design Basis Source Term Development Process (example: MOX & High Burnup Fuel) 9Fission Product TransportMELCOROxidation/Gas Generation Experimental BasisMelt ProgressionFission Product ReleasePIRT processAccident AnalysisDesign BasisSource TermScenario # 1Scenario # 2------.Synthesize timings and release fractionsCs Diffusivity

• Similar RFs to NUREG

-1465 but prolonged release

• Differences not from change of fuel but from code advancesScenario # n

-1Scenario # 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

-LWRsJonathan BarrOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionDecember 13, 2018 2MACCS 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, uncertaintyMACCS Gaussian plume segment ATD model animation for a single weather trial 3MACCS 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 4MACCS 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 5MACCS 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 6Near-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 phenomenaLloyd 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

-390Wind tunnel simulation of streamlines near a cubic building 7Near-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 applicationQUIC Factsheet, Los Alamos National LaboratoryExample QUIC

-URB simulation of wind vectorsExample QUIC

-PLUME simulation of urban transport and dispersion Establishing Interface Requirements in Support of Staged LicensingDecember 13, 2018Ashley Finanashley@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" 3Table of Contents:Executive SummaryIntroductionPurpose and ScopeStandard Design ApprovalMethods to Develop Interface RequirementsExample CasesCore DesignReactor Vessel Auxiliary Cooling System DesignReactor Coolant System Piping DesignReactor Building Structural DesignConclusions 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 5Figure 1: Current Project Risk/Investment Profile Relative to Detailed Design & LicensingFigure 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 13ARDCTitleSampleInterfaceRequirementsforRVACSystem 2Design basis for protection against natural phenomenaInterface RequirementThe 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.

3Fire protectionInterface RequirementThe 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 FeedbackThank you!

14 Thank youFeedback & QuestionsPlease feel welcome to send additional input at any time to Ashley Finan (ashley@nuclearinnovationalliance.org).

Priorities for Advanced Reactor Developers:USNIC Survey of Developer PrioritiesDavid BleePresident & CEOU.S. Nuclear Industry Council Hon. Jeffrey S. MerrifieldFormer Commissioner, USNRC;Chairman, USNIC Advanced Reactors Task Force; Partner, Pillsbury Winthrop Shaw PittmanDecember 13, 2018 USNIC AR Developers SurveyUSNIC conducted a third in a series survey of 16 leading U.S. Advanced Reactor technology developers with regard to DOE Initiatives15 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 LicensingFeedback is intended to give constructive input to the Commission and StaffSurvey provides a snapshot of the current policy priorities of the Advanced Reactor CommunityAssessment goes beyond the efforts of the Office of New Reactors to include the preparations of other NRC officesProvides feedback on the perceived technical readiness of the NRC staff 3

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

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

5NRC Chairman & CommissionersOffice of New ReactorsOffice of Nuclear Material Safety and SafeguardsOffice of Nuclear Security and Incident Response Q3: Planning Timeframe for Licensing Application Submittals: What should the NRC and DOE's Planning Timeframe be for new Advanced Reactor License Applications?

6 Q4: Focus for NRC Advanced Reactors Licensing Transformation in 2019: What should the NRC's 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 NRC's pre

-licensing engagement (rank in orderof priority)?

9Cost-share for pre

-licensingFixed price and schedule certainty for pre-licensingEnhanced NRC Advanced Reactor Technology capabilityMore robust stakeholder engagementAdditional involvement by the Office of New ReactorsAdditional involvement by the Office of Nuclear Security & Incident ResponseAdditional involvement by the Office of Nuclear Material, Safety & Safeguards Q7: NRC Advanced Reactors Technical Capability: Please rate the NRC's 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 NRC's Stakeholder Meetings (held every 6

-8 weeks)?13 Q11: Do you believe the NRC Office of Research is putting sufficienttime and resources towards Advanced Reactordevelopment?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 ResultsCommission and staff of Office of New Reactors are perceived as making progress on Advanced Reactor policy decisions and licensing readinessOffice 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 issuesAgency readiness for High Temperature Reactors is very goodHigher level of questioning about NRC readiness to license Molten Salt, Fast and Liquid Metal ReactorsThere is a lack of understanding of what the Office of Research is doing to assist in preparing the NRC for Advanced ReactorsThere was an overwhelming view that the Commission needs to do more to assist in lifting the burden of Fee Based programs on Advanced Reactors16 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 visitwww.usnic.orgU.S. Nuclear Industry Council1317 F Street, NW

-Washington, DC 20004(202) 332-8155 www.usnic.org17 Copyright © 2018 URENCO LimitedStephen Cowne, Chief Nuclear Officer, UUSAMeeting on Possible Regulatory ProcessImprovements for Advanced ReactorsDecember 13, 2018Next Generation Nuclear Fuels The Nuclear Institute: Advance Nuclear Technologies 1Copyright © 2018 URENCO LimitedToday's Front

-End Nuclear Fuel CycleLWR Fuels LEU-UO 2-ZircAlloy LWR FuelsLEU-UO 2-ZircAlloy 2DOE ProgrammeAccident Tolerant FuelSite LicensingCat-II FacilityOperational Criticality & SafetyIntrinsicallySafe FuelsNationalRegulator(s)Deconversion/H2M

• U-Metal*U-Oxides*U-SaltsStorage & Transport

• Cylinders*Overpacks*Class 7 Shipping

• InsuranceFuel FabricationHigher EnrichmentHA-LEU ~19.75%EnrichmentLEU+Plus (5~10%)Test & ResearchReactorsMolten SaltReactorsLead CooledReactorsFast BreederReactorsSodium FastReactorsHTGRGen-IIIReactor UpratesSMRsMicro-SMRsTRISO Fuel

• UCO*U02*Uranium Nitride

• Uranium SilicideATF High Density Fuel Pellets

• U-Silicide*U-Nitride*Chromium doped U02

• FCM Ceramic FuelFabricated TRISO

• Prismatic Block

• Pebble BedATF Cladding Systems

• Chromium coating

• Silicon-carbide claddingMetallic Fuel

• Lightbridge Zr

-U Alloy*U-MolybdenumLiquid Fuels

• Molten salts

• Aqueous uranyl salt solutionsRepUExisting UO2 Fuel Pellets

• ~5.95% EnrichmentNext Generation Fuel Pathways: Range of optionsCopyright © 2018URENCO Limited The Nuclear Institute: Advance Nuclear Technologies 3Copyright © 2018 URENCO LimitedThe Future Nuclear Fuel Supply ChainExisting Nuclear Fuel Supply ChainMiningConversionEnrichmentFabricationBack End U 3 O 80.711%UF 6<5%LEU UO 2SpentFuelLWR Reactors UO 2/ ZircAlloyFuelsFabrication0.711%UF 6<5%LEUNext Generation FuelsTRISO (UCO),Uranium Nitride,Uranium Silicide

, U-metal Alloys UF 4Saltsetc-Gen III+, ATFsSMRs, GenIV , Advanced ReactorsResearch & Test Reactors 5%-20%HA-LEU U-metal U-oxide U-saltsCompleting the Future Nuclear Fuel Supply ChainEnrichmentHigherEnrichmentDeconversion The Nuclear Institute: Advance Nuclear Technologies 4Copyright © 2018 URENCO Limited HA-LEU and the HA-LEU Community*High Assay

-Low Enriched Uranium (HA

-LEU) refers to enrichments above 5.0% U235 and below 20.0% U235.

• 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 5Copyright © 2018 URENCO Limited HA-LEU Fuel Cycle

• A complete and sustainable HA-LEU fuel cycle includes three fundamental capabilities:

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 6Copyright © 2018 URENCO LimitedTransport & Packaging Considerations

• Are HA-LEU UF6 shipments limited to use of a small packaging?

• Are moderator exclusion requirements met through the cylinder or through an overpack?*Criticality benchmarking data is needed for HA-LEU assays.CylinderModelDiameter (inches / mm)Maximum EnrichmentMaximum 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 30B30 / 762 5%5020 / 2277Existing UF6 Cylinders for Higher Assays (ANSI N14.1)

The Nuclear Institute: Advance Nuclear Technologies 7Copyright © 2018 URENCO Limited 2-Box Model:

Co-location of Enrichment & DeconversionProblem:*There is currently no available "transport package" for HA

-LEU.Possible Solution: "2-Box" Model: Co

-location of Higher Enrichment and Deconversion Facilities

.<5% UF 60.711% ENU<19.99% UF 6UF6 DeconversionFacility<19.99%U-metal U-oxide U-saltsNext Generation Fuel Manufacturing FacilityFabricated HA-LEU FuelsTRISO (UCO)

U0 2 U-metal Alloys UF 4SaltsUranium NitrideUranium Silicide(Cat 2 License)Higher EnrichmentFacility*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 8Copyright © 2018 URENCO Limited HA-LEU Fuel Cycle: Licensing Approach1a. Enrichments up to 5.5%

• 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 UF 6above 5%*Criticality benchmarking data is needed for HA

-LEU assays*For use with UO 2fuel pellets2b. UF 6deconversion

• For other fuel types

• If existing transport packages are not approved at higher enrichments The Nuclear Institute: Advance Nuclear Technologies 9Copyright © 2018 URENCO Limited HA-LEU Fuel Cycle: Licensing Challenges 1.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%.

The Nuclear Institute: Advance Nuclear Technologies 10Copyright © 2018 URENCO Limited HA-LEU Fuel Cycle: Initial Observations 1.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!"

The Nuclear Institute: Advance Nuclear Technologies 11Copyright © 2018 URENCO LimitedURENCO: An Integrated Supplier11Thank You SECY-18-0076OPTIONS AND RECOMMENDATION FOR PHYSICAL SECURITY FOR ADVANCED REACTORSDecember 13, 2018 1 2BackgroundNRC 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 3BackgroundNRC 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.

4Background

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

oThe staff's assessment determined that the current security regulatory framework is adequate to certify, approve, and license iPWRs-oThe 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 ConditionsExemptions

• "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."

5SECY-18-0076 OptionsIdentifies 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 6Option 3 -Limited Scope Rulemaking

• Revise specific regulations and guidance related to physical security for SMRs and non

-LWRs through rulemaking.

oExample -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.

7Staff Requirements Memorandum (SRM)SRM Dated November 19, 2018 The Commission approved the staff's 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 Commission's SRM

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

.

8Rulemaking Process 9Barrier 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 10Revisit First Principles 11NEI Proposed Logic for Applicability of Alternate Regulations(Armed Responders Not Required)Possible Performance (Consequence)Based Approach 12Security Design ConsiderationsPreliminary 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 13DiscussionPotential Scope of Alternative Requirements

• 10 CFR 73.55(k)

-armed responders

• 10 CFR 73.55(i) -secondary alarm stations

• ?*?*?

14Stakeholder Presentation/Discussion NEI 15DiscussionStakeholder Presentation/Discussion USUCS 16General DiscussionPublic Questions/Feedback