ML19228A263

From kanterella
Revision as of 02:02, 5 September 2019 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
August 15, 2019 Advanced Reactors Stakeholders Meeting Presentation Slides
ML19228A263
Person / Time
Issue date: 08/15/2019
From: Vechioli-Feliciano L
Office of New Reactors
To:
Vechioli L, NRO/DAR/ARLB, 415-6035
References
Download: ML19228A263 (120)
v • d • e


Text

Public Meeting on Possible Regulatory Process Improvements for Advanced Reactor DesignsAugust 15, 2019 1Telephone Bridge: (800) 593

-0695 Passcode: 3798624 Public Meeting

  • Telephone Bridge 800-593-0695 Passcode: 3798624
  • Opportunities for public comments and questions at designated times 2

3Outline9:00-9:10 amOpening Remarks9:10-9:40 amStatus Update9:40-10:25 amTechnology Inclusive Content Applications Project (TICAP)10:25-10:35 amBreak10:35-11:00 amMicro-Reactors Regulatory Topics11:00-11:30 amConcrete and Elevated Temperature11:30-12:15 pmEnvironmental Interim Staff Guidance and Generic EIS Developments12:15-1:15 pmLunch1:15-1:45 pm10 CFR Part 53: Risk

-informed, Technology Inclusive Regulatory Framework for Advanced Reactors Rulemaking1:45-2:15 pmNuclear Energy Innovation and Modernization Act (NEIMA)2:15-2:45 pm Non-LWR Vision and Strategy Computer Code Reports: Modeling and Simulation of non

-LWRs 2:45-3:15 pm NRC Non-LWR Computer Code Development Plans 3:15-3:30 pmBreak3:30-4:00 pmProtecting Sensitive Information 4:00-4:30 pm Open Discussion

  • NRC Updates
  • DOE Updates 4
  • Technology Inclusive Content of Applications Project (TICAP)

-Amir Afzali, Southern Company Services 5

6BreakMeeting/Webinar will begin shortlyTelephone Bridge 800-593-0695 Passcode: 3798624

  • Micro-Reactors Regulatory Topics

-Marc Nichol, NEI 7

  • Concrete and Elevated Temperature

-Madhumita Sircar, NRC 8

  • Environmental Interim Staff Guidance and Generic EIS Developments

-Donald Palmrose/Mallecia Sutton, NRC 9

10LunchMeeting/Webinar will begin shortlyTelephone Bridge 800-593-0695 Passcode: 3798624

  • 10 CFR Part 53: Risk

-Informed, Technology Inclusive Regulatory Framework for Advanced Reactors Rulemaking

-William Reckley, NRC 11

  • Nuclear Energy Innovation and Modernization Act (NEIMA)

-NRC Section 103 Activities

  • John Segala, NRC

-Establishing Metrics and Milestones

  • Steven Lynch, NRC12
  • Non-LWR Vision and Strategy Computer Code Reports

-Stephen Bajorek, NRC

  • NRC Non-LWR Computer Code Development plans for severe accident progression, source term, and consequence analysis

-Hossein Esmaili, NRC13 14BreakMeeting/Webinar will begin shortlyTelephone Bridge 800-593-0695 Passcode: 3798624

  • Protecting Sensitive Information

-Stewart Magruder, NRC15 162019 Tentative Schedule for Periodic Stakeholder MeetingsOctober 10December 12 17Open Discussion and Closing Technology Inclusive Content of Application ProjectAmir AfzaliSouthern CompanyNRC Advanced Reactor Public Meeting Briefing August 15, 2019TICAP Introduction

  • Project Purpose: To collaboratively work with the ARRTF and NRC to develop a technology inclusive formulation for preparing content of application that will have the following attributes:
  • Versatile

-The variance in technologies and designs requires robust application guidance that is versatile enough to be used by most if not allpotential applicants

  • Systematic

-It facilitates thorough and consistent safety assessments for different designs across and within different technologies

  • Compatible

-It correlates to the underlying safety basis/objective of the current light

-water centric content of application, thereby demonstrating consistency with the Commission's mission of protecting public safety 2

Status and Important Inputs

  • Status*Team formed
  • Kickoff meeting July 18, 2019
  • Second meeting August 8, 2019
  • Focus on developing Project Plan (due September 30)
  • Plan to Propose a set of fundamental safety functions by 12/20/2019
  • Product or process/method for NRC's interaction (e.g., WP for NRC's official review) not yet defined.
  • Necessary Event
  • Regulatory Guide endorsing NEI 18 Expected Date Dec. 2019 3

TICAP Success Criteria and Guiding PrinciplesGuiding Principles

  • Right-size content of application to be commensurate with the complexity of the applicant's safety case.
  • Improve the overall efficiency of review and optimize generation of application's information in terms of scope and level of details.
  • Is limited to utilizing the outputs developed through application of theLicensing Modernization Project process defined in NEI 18

-04*Will be technology inclusive

  • Will include a formulation that facilitates meeting the underlying safety basis/objective of the current regulations(provide an alternative to show compliance with the underlying safety basis/objective vs. creating equivalency)

-*For certain designs this mean certain information required as part of application for a LWR may not be included

  • May result in regulatory questions being raised (expect conversations around 2020, 3 rdquarter).*Should contribute beneficially to the development of the NRC's new regulatory framework for advanced reactors (10 CFR Part 53) 4Success CriteriaProvide a document that outlines the content of an application in a manner that is technology inclusive, uses LMP methodology and can be submitted to NRC for endorsement within two calendar years from initiation of Phase 2.

5Consequence Based Security(SECY-18-0076)EP for SMRs and ONTs(SECY-18-0103)Functional Containment (SECY-18-0096)Insurance and LiabilitySiting near densely populated areasEnvironmentalReviewsLicensing ModernizationProjectSimplified representation role of LMP within regulatory framework Fundamental Consideration Based on Use of LMP ProcessTICAP will propose what level of information is needed for SSCs and their associated programs for each class of SSC

?7 Action Plan for Reviewing and Endorsing Non

-LWR PRA StandardAdvanced Reactor Stakeholder MeetingAugust 15, 2019 August 15, 2019 Page 2Outline of Plan

  • Objectives and Scope
  • TasksTask 1: Supporting development of the standardTask 2: Preparation for review of the standardTask 3: Reviewing the standardTask 4: Maintaining PRA standardTask 5: Development of scheduleTask 6: Identification of resourcesTask 7: Development of communication plan Task 2: Preparation for Review of the Standard

-Involves 6 Subtasks*Subtask 2-4: Develop staff position for an acceptable non

-LWR PRA (task initiated)Define the objectives for each technical element, considering the different applicationsDefine the technical attributes and characteristics needed to accomplish the objective Develop the staff position on an acceptable peer review process, addressing an acceptable peer review process, team qualifications, and documentation

  • Subtask 2-5: Identification and resolution of technical and policy issuesReview each technical element associated with each risk level and hazard, for each application type, and identify possible technical or policy issuesDescribe the significance of the issuesIdentify whether there is ongoing research to address the issues and what research are needed
  • Subtask 2-6: Guidance for staff review of non

-LWR PRA standard for endorsement (draft completed)Guidance on how to approach the reviewCriteria for determining acceptance (no objection)August 15, 2019Page 3 Status of NRC RG, ASME/ANS Standard, and NEI GuidanceUnder development

  • Initiated staff positionUnder development
  • Trial Use issued
  • Final, publication in 2020/2021Under development ?????PRA AcceptabilityConsensus Standardto demonstrate conformance with RPNRCRegulatory Position (RP) (RG x.xxx)Peer Reviewto demonstrate conformance with StandardAugust 15, 2019Page 4

©2019 Nuclear Energy InstituteRegulatory TopicsMarc Nichol, Director New Reactor DeploymentMicro-ReactorsAugust 15, 2019

©2019 Nuclear Energy Institute 2Micro-Reactor Regulatory IssuesPriorityIssuesAddressed in BroaderEffortsNon-Urgent 1.Review Scope,Duration, Level of Effort 2.Aircraft Impact 3.Operations (auto/remote) 4.Resident Inspector 5.Physical Security 6.Emergency Preparedness

  • Siting*Environmental Reviews
  • Transportation
  • Annual Licensee Fees
  • Fuel*Generic License
  • PRA*QANo issues identified to

-date*Liability Insurance

  • Decommissioning Funding

©2019 Nuclear Energy Institute 3Typically 1 MWeto 10 MweVery small size

  • Site <0.1 acres, building ~size of a house, reactor fits in shipping containerVery small potential consequences
  • Source terms as low as 1% of today's reactors
  • Fail-safe: shuts itself off, cannot meltdown
  • Proliferation resistant fuelOperational simplicity
  • Few to zero moving parts
  • Automatic operations
  • Minimal maintenanceUnique Micro

-Reactor Considerations**General description, all features may not be applicable to all designs 1International Conference on Radioecology and Environmental Radioactivity, Bergen, NorwayEnvironmental Interim Staff Guidance and Generic EIS DevelopmentsOffice of New ReactorsU.S. Nuclear Regulatory CommissionAdvanced Reactor Stakeholders MeetingAugust 15, 2019Donald Palmrose, PhD Mallecia SuttonSr. Reactor Engineer Sr. Project Manager Agenda*Staff seeking input on:

-Draft ISG-GEIS-Possible EA for some Advanced Reactors

  • Open discussion 2Advanced Reactor Stakeholders Meeting on August 15, 2019 3Advanced Reactor Stakeholders Meeting on August 15, 2019Interim Staff GuidanceEnvironmental Considerations Associated with Micro

-reactors*Provides supplemental staff guidance for the environmental review to address differences with large LWRs:

-Smaller footprint affects fewer resources

-May not use cooling water

-Smaller rad and non

-rad waste streams

-Reduced socioeconomic impacts

-Smaller size generally translates to fewer impacts

  • Provides guidance for how to "incorporate by reference," or IBR, for environmental reviews
  • Reducing duplication of effort, size of documents while maintaining quality to meet NRC's NEPA obligations
  • Issuance of a draft ISG for comment and use by December 2019 4Advanced Reactor Stakeholders Meeting on August 15, 2019Interim Staff Guidance (cont.)

Considerations forAdvanced Reactor Reviews

  • GEIS feedback

-Review of previous GEISs for benefits, costs, and limitations

-Appropriate scope acceptance criteria

-Enough publicly available data

-Staff resource assessment

-Need for new supporting studies

-Need for rulemaking

-What would be a realistic schedule 5Advanced Reactor Stakeholders Meeting on August 15, 2019 Considerations forAdvanced Reactor Reviews

  • EIS versus EA feedback
  • Can an EA address some advanced reactor reviews-10 CFR 51.21

-Rulemaking needed?

-Considering staff guidance 6Advanced Reactor Stakeholders Meeting on August 15, 2019 Open Discussion 7Advanced Reactor Stakeholders Meeting on August 15, 2019 Concrete and Elevated TemperatureMadhumita Sircar

    • U.S. Nuclear Regulatory Commission
  • Office of Nuclear Regulatory Research
    • Office of New ReactorsAdvanced Reactors Stakeholders Meeting Nuclear Energy Institute Washington, DC. August 15, 2019 Scope of Presentation:
  • Effects of High Temperature on ConcreteCurrent Code Requirements by ACI 349:
  • Provision E.4, "Concrete Temperature,"

150 oF for Normal Operations 350 oF for Accident or Short

-Term Period 2

Key References

  • NUREG/CR-6900,"The Effect of Elevated Temperature on Concrete Materials and Structures

-A Literature Review. US NRC, March 2006."

  • NUREG/CR-7031, "A Compilation of Elevated Temperature Concrete Material Property Data and Information for Use in Assessments of Nuclear Power Plant Reinforced Concrete Structures. US NRC, December 2010."

Advanced Reactor Concrete Structures under High Temperature

  • Concrete Reactor Building which may be one of the Functional Containments

-Independent barrier provide defense

-in-depth*Other Concrete Structures 4

Example of a Passive Cooling System for Concrete Structures under High Temperature Independent barriers provide defense

-in-depth[Source: NRC HTGR training prepared by INL]

5 Concrete under High TemperatureUltimate Compressive Strength, psiTemperature, °CTemperature, °FUltimate Compressive Strength, MPaTemperature, °FTemperature, °CModulus of Elasticity, MPaModulus of Elasticity, 10 6psi 6Ultimate compressive strength and modulus of elasticity of Type I Portland cement paste (w/c:0.33)

[Source:Harmathyet al. Fig. 1 and 2 as referenced in NUREG/CR

-6900]

Concrete under High Temperature, cont.Comparison of Effect of Elevated Temperature on the Compressive Strength of Concretes Fabricated using Different Types of Conventional Aggregate Materials [Source: Blundell et al. Fig. 2.121 as referenced in NUREG/CR

-7031]

Concrete under High Temperature, cont.Comparison of the Effect of Elevated Temperature on the Tensile Strength of Concretes Fabricated using Different Types of Conventional Aggregate Materials [Source: Blundell et al. Fig. 2.121 as referenced in NUREG/CR

-7031]

Comparison of Effect of Elevated Temperature on the Relative Bond Strengths of Mild Steel to Concretes Fabricated using Different Types of Conventional Aggregate Materials[Source: Sullivan Fig. 2.174 as referenced in NUREG/CR

-7031]Concrete Under High Temperature, cont.

Concrete under High Temperature, cont.The Effect of Temperature on the Compressive Strength of Portland Cement Concrete

[Source: Fig. 2.86 as referenced in NUREG/CR

-7031]

Concrete under High Temperature, cont.

11[Source: Fig. 2.6 as referenced in NUREG/CR

-7031]

Summary*Prevention of elevated temperature in concrete

  • Preferably meeting ACI 349 code limits
  • Reliability of preventive systems
  • Condition monitoring
  • Other considerations

-Thermal cycling

-Normal and accident conditions 12 THANKS 13 10 CFR Part 53: Risk

-informed, TechnologyInclusive Regulatory Framework forAdvanced Reactors RulemakingAdvanced ReactorStakeholder MeetingAugust 15, 2019 Nuclear Energy Innovation and Modernization Act (NEIMA)

  • NEIMA Section 103 requires that the NRC "complete a rulemaking to establish a technology

-inclusive, regulatory framework for optional use by commercial advanced nuclear reactor applicants for new reactor license applications."

  • Rulemaking is to be completed no later than December 31, 2027.

NEIMA*NEIMA defines "advanced nuclear reactor" as "a nuclear fission or fusion reactor, including a prototype plant . . . with significant improvements compared to commercial nuclear reactors under construction" as of January 14, 2019, including improvements such as additional inherent safety features; significantly lower levelized cost of electricity; lower waste yields; greater fuel utilization; enhanced reliability; increased proliferation resistance; increased thermal efficiency; or ability to integrate into electric and nonelectric applications.

NRC Staff Activities

  • NRC has formed a staff working group
  • Working group is drafting a rulemaking plan and formulating initial thoughts on scope of rule*Today's meeting marks first staff outreach to external stakeholders
  • A focused public workshop is being planned for October 2019 Rule Applicability
  • NEIMA's definition of "advanced nuclear reactor" covers:

-Light-water small modular reactors

-Non-light-water reactors (non

-LWRs)-Fusion reactors

  • The staff interprets NEIMA as not requiring the rulemaking to cover reactor technologies similar to current operating reactors or Generation III+ large LWRs Questions for Discussion
  • Question: Which types of requirements should be included?

-Technical requirements equivalent to 10 CFR Part 50?

-Licensing processes equivalent to 10 CFR Parts 50 & 52?

-Complete plant/license life cycle or initial license applications?

-Level of detail for technical requirements

-All technical requirements, including security and emergency preparedness?

Next Generation Nuclear Plant (NGNP) ConceptsDeveloping Functional Requirements 8Licensing Modernization ProjectFigure 4.2. Definition of Risk

-Significant and Safety

-Significant SSCs 9Integrated ApproachConsequence Based SecurityEP for SMRs and ONTsFunctional Containment Insurance and LiabilitySiting near densely populated areasEnvironmentalReviewsLicensing ModernizationProject NRC Path Forward

  • Draft a rulemaking plan, taking into consideration today's feedback
  • Hold more focused public meeting(s) in the months ahead
  • Finalize rulemaking plan and associated Commission paper
  • Send rulemaking plan to Commission in April 2020.
  • Documentation related to the Part 53 rulemaking can be found on the regulations.govwebsite by searching for the NRC Docket ID "NRC-2019-0062" Nuclear Energy Innovation and Modernization Act:Establishing Metrics and MilestonesSteven LynchActing Chief, Advanced Reactor Licensing BranchDivision of Advanced Reactors 1

Background

  • Nuclear Energy Innovation and Modernization Act (NEIMA) signed into law on January 14, 2019
  • Section 102(c) of NEIMA requires:

1.The development of performance metrics and milestone schedules for "requested activities of the Commission" 2.Reports for certain delays associated with these activities 2

Definition of "Requested Activities"

  • Section 3(10) of NEIMA defines "requested activity of the Commission" as A.The processing of applications for i.Design certifications or approvals ii.Licenses iii.Permits iv.License amendments v.License renewals vi.Certificates of compliance vii.Power uprates B.Any other activity requested by a licensee or applicant
  • In general, for the purposes of NEIMA, requested activities of the Commission involve the preparation and issuance of a final safety evaluation by the NRC 3

Establishing Milestone Schedules

-nrc/generic

-schedules.html

-Generic schedules based on historical data, ongoing reviews, and modernization efforts

  • For the initial application reviews for non

-light water reactors, a generic milestone schedule of 36 months (30 months if referencing a certified design) has been established

  • Application

-specific schedules, which may be shorter or longer, will be established for each requested review 4 Impacts on Specific Review Schedules

  • Quality of application

-Adherence to regulatory requirements

-Technical completeness

-Attention to detail (i.e., organization, format, etc.)

  • Requests for additional information (RAIs)

-Complexity and novelty of technology

-Completeness, timeliness, and responsiveness to requests

-Number of RAIs and need for follow

-up-Evaluation of new information

  • Policy questions

-Commission involvement to resolve unique considerations

  • Advisory Committee on Reactor Safeguards

-Number of subcommittee meetings

-Follow-up items 5 Other Scheduling Considerations

  • Potential for contested hearing
  • Mandatory hearing for certain applications

-Cannot hold mandatory hearing until completion of Safety Evaluation Report, Environmental Impact Statement, ACRS Review, and any contested hearing

  • Commission decision to issue or deny permit or license

-Decisions typically made 2

-4 months following mandatory hearing 6

Metrics and Reporting

  • The performance indicator is 100 percent timely completion of final safety evaluations within the established generic milestone schedules
  • The NRC staff to notify the Commission within 30 days of missing a generic milestone schedule
  • The NRC staff to prepare a report to Congress if a requested activity is not completed within 180 days after the established generic milestone schedule 7 Performing Effective Reviews
  • Meeting the performance metrics and milestone schedules established in NEIMA supports NRC staff commitment to performing effective reviews
  • Prospective applicants should engage with NRC staff early on anticipated licensing actions to develop specific review schedules 8 Nuclear Energy Innovation and Modernization Act (NEIMA): NRC Section 103 ActivitiesJohn SegalaChief, Advanced Reactor Policy BranchDivision of Advanced ReactorsAugust 15, 2019 Sec. 103. Advanced Nuclear Reactor Program a)Licensing 1)Staged Licensing 2)Risk Informed Licensing 3)Research and Test Reactor Licensing 4)Technology

-Inclusive Regulatory Framework 5)Training and Expertise 6)Authorization of Appropriations b)Report to Establish Stages in Licensing Process c)Report to Increase RIPB Techniques d)Report to Prepare RTR Licensing Process e)Report to Complete Rulemaking NRC Staff Activities

  • Issued a Letter to Congress on July 12, 2019 (ADAMS Accession # ML19128A289) enclosing two reports:

1.Establishing Stages in Advanced Reactor Licensing (Sec. 103(b))

  • Implementation of stages in licensing process within 2 years-NRC has completed staged licensing activities with issuing Regulatory Review Roadmap

-Topical Reports, Standard Design Approval, Preapplication Engagement, and Licensing Project Plans/Regulatory Engagement Plans

  • Required evaluations

-Fuel Qualification, Industry Codes and Standards, etc.

NRC Staff Activities (Cont.)

2.Increasing Use of RIPB Techniques and Guidance (Sec. 103(c))*Licensing Modernization Project (NEI 18

-04, DG-1353, and draft SECY paper)

  • Mechanistic Source Term (SECY 092, NRC contract with INL to develop guidance)
  • Emergency Preparedness (Proposed rule provided to Commission on October 12, 2018)
  • Other Policy Issues (Siting as it relates to population, Physical Security, and Micro Reactors)
  • Coordination and stakeholder input

-Public Stakeholder meeting on March 28 th*Cost and schedule estimates

-Non-fee recoverable advanced reactor appropriations NRC Staff Activities (Cont.)

  • Issued Internal Memo on August 8, 2019

-Staff training or hiring of experts to support staged licensing, risk

-informed licensing, research and test reactor licensing, and technology

-inclusive regulatory framework (Sec. 103(a)(5))

  • Training (IAP Strategy 1)

-Technology Training Courses (MSRs, SFRs, Micro, HTGRs)

-Computer Code training (MOOSE and BISON)

-Research and Test Reactor Training

  • Knowledge Management

-Contractor Reports, MOUs with DOE to share expertise and knowledge, etc*Hiring-Competency Modeling, New Division, Core Review Team, and Merger of NRO and NRR Non-LWR Vision and Strategy Computer Code ReportsModeling and Simulation of non

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

-2345 / Stephen.Bajorek@nrc.govAdvanced Reactor Stakeholder's MeetingAugust 15, 2019RES Implementation Action Plan for Advanced Non

-LWR ; Codes and Tools Slide 2NRC's Integrated Action Plan (IAP) for Advanced Reactors 2Near-Term Implementation Action PlanStrategy 1Knowledge, Skills, and CapacityStrategy 2Computer CodesStrategy 3Flexible Review ProcessStrategy 4Industry Codes and StandardsStrategy 5Technology Inclusive IssuesStrategy 6Communication NRC's Implementation Action Plan, Strategy 2

-Computer Codes IntroductionVolume 1Volume 3Volume 2 = Fuel PerformanceVolume 4 = Radiation ProtectionUnder DevelopmentML19093B266ML19093B322ML19093B404 4

Slide 4Event Selection 4*"Chapter 15" vs "Chapter 19" deterministic approach to be replaced with LMP.

  • "Design Basis" Code(s) = those to be used for confirmatory analysis of events that little/no core (geometric) disruption or fission product release.

-Unprotected loss of flow

-Loss of heat sink(s)

-Events that may involve multiple failures

  • "Beyond Design Basis" Code(s) = for events involving core melt, fission product release & transport.

Slide 5Some Recent & Upcoming Events . . .

5*Technical approach in Volumes 1 and 3 discussed with ACRS "Future Plant Design" subcommittee on May 1, 2019.

>>Scenarios>>"Gaps"*September 5 "Data Needs" Meeting with DOE.

  • Volume 2 (Fuel Performance) to be discussed with ACRS "Future Plant Design" subcommittee on September 17, 2019.

Slide 6 6Volume 1 "Design Basis Event Analysis" Slide 7Introduction / Outline 7*Volume 1 "Design Basis Event Analysis" :

-Phenomena Identification and Ranking Tables (PIRTs)-Event Scenarios

-"New" Physical Phenomena for non-LWRs-Gaps -Tasks Slide 8 Slide 9Characterization of Design TypesPlant Type No.Description Fuel 1HTGR; prismatic core, thermal spectrumTRISO (rods or plates) 2PBMR; pebble bed core, thermal spectrumTRISO (pebbles) 3GCFR; prismatic core, fast spectrumSIC clad UC (plates)4SFR; sodium cooled, fast spectrumMetallic (U

-10Zr)5LMR; lead cooled, fast spectrumNot available. (Possibly nitride fuel.)6HPR; heat pipe cooled, fast spectrumMetallic (U

-10Zr)7MSR; prismatic core, thermal spectrumTRISO (plates) 8MSPR; pebble bed, thermal spectrumTRISO (pebbles) 9MFSR; fluoride fuel salt, thermal/epithermal spectrumFuel salt 10MCSR; chloride fuel salt, fast spectrumFuel salt Slide 10"Modeling Gaps" Identified by PIRTs 10*Phenomena that are significant and "new" with increased importance for non

-LWRs relative to conventional LWRs include but are not limited to:

  • -Thermal stratification and thermal striping

-Thermo-mechanical expansion and effect on reactivity

-Large neutron mean

-free path length in fast reactors

-Transport of neutron pre

-cursors (in fuel salt MSRs)

-Solidification and plate

-out (MSRs)

-3D conduction / radiation (passive decay heat removal)"Modeling Gaps in NRC Codes" Slide 11 11Code 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 validation is critical and represents the major gap in EM development. Database is weak for some designs.
  • Computation Requirements. Must be able to run simulations on NRC desktops or HPC platforms readily available to NRC. Codes selected for CRAB satisfy these criteria.

Slide 12TRACESystem T/HMOOSEPARCSNeutronicsSCALE Cross-sectionsFASTFuel PerformanceBISONFuel PerformancePRONGHORNCore T/HSAMSystem and Core T/HNek5000CFDMELCORContainment / FPDOE CodeNRC CodeMAMMOTHNeutronics Comprehensive Reactor Analysis Bundle (CRAB) SERPENT Cross-sectionsSERPENT Cross-sectionsMAMMOTHNeutronicsInt'l CodeFLUENTCFDCommercial Slide 13Approach to Validation 13 (1)Review PIRT phenomena & prior test programs for applicability to each of the new designs.

(2)Identify and prioritize validation tests (based on PIRT findings and NRO expected review schedule).

(3)Develop "reference plant" models to define nodalizationscheme and modeling options.

(4)Coordinate efforts with DOE and national labs to complete validation & improve code performance based on findings.

Slide 14Summary & Conclusions"Volume 1" recommends the codes in CRAB as the approach for non

-LWR DBE analysis. Flexibility to simulate multiple designs (including LWRs with ATF)."Gaps" in code capability, V&V are identified along with tasks for resolution. Using the combination of NRC and DOE codes will provide a technically superior productthan can be attained with further development of only the NRC's conventional LWR codes.

Slide 15Extra Stuff Slide 16Codes for Design Basis Event Analysis 16*Codes considered:

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

-DOE NEAMS codes (MAMMOTH, PRONGHORN, RELAP7)

-ANL codes (SAS4A/SASSY, SAM, PROTEUS, 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 both NRC and DOE.

Slide 17Unique Capabilities Available in CRAB 17*Examples-Multiphysics Coupling

-Geometric Fidelity

-Advanced Equivalence Methods

-Multi-Scheme Capability

-3D Reduced Order Flow ModelGoal: Enable analysis of advanced designs without over

-simplifying assumptions to provide intermediate fidelity model for modest computational resources.

Slide 18 18Multiphysics CouplingSAM: System Level Thermo

-FluidsTensor Mechanics ModuleMAMMOTH: Rx KineticsTemperatures & DensitiesPowerTemperaturesDisplacements Slide 19 19Geometric FidelityHTR-10Bottom Reflector & ConusPRONGHORN Mesh D Channels for upcomer & CR's D Porous body for conus Slide 20 20Advanced Equivalence MethodsHTR-PMPebble bed HTR (420k pebbles)Diffusion problem size

-54,656 cellsHTTRPrismatic HTRDiffusion problem size

-15,552 cells HTR-PM (Tfuel= 1100 K)keffpcmRMS % Err fMax %Err fSerpent1.01159+/-1.8--Diffusion1.03653 24356.040.6 SPH-Diffusion1.01159 01.55E-032.94E-03HTTR ( Tfuel= 1300 K)keffpcmRMS % Err PowerMax %ErrPowerSerpent1.00259+/-2.7--Diffusion1.01978 17153.126.20 SPH-Diffusion1.00259 07.0E-022.0E-01Transport level accuracy for the price of adiffusion calculation Slide 21 21Multi-Scheme Capability

  • MAMMOTH-Allows usage of transport where more detail is needed with efficiency of diffusion for remainder of domain Slide 22 22SAM: 3-D Flow Model
  • Validation ExamplesLid-Driven Cavity FlowNatural ConvectionRa = 10 5 Slide 23Verification & Validation "Gaps" 23 Slide 24Code Assessment Issues 24*Code "Assessment" = Verification

& Validationrepresents the most significant "gap" in readiness for the DBE analysis codes.

  • Verification: Considered generally good

-however "coupling" may need additional cases to ensure conservation of mass, energy & momentum.

  • Validation: Completed and on

-going validation shows good agreement between predicted & measured results. More is needed, and should be done with a "frozen" code.

Slide 25Code Validation Matrix 25*Volume 1 identifies the most important validation cases for each of the 10 design types. Additional validation is being performed by DOE as part of developmental assessment.

  • An additional report is being developed to summarize all of the V&V needed for CRAB.

Slide 26Validation Status 26GCRs: HTR-10, PBMR-268,-400, SANA, HTTU, AVR, . . . SFRs: EBR-II , FFTF, CEFR , ZPPR , Monju, . . . LMRs: HeliosHPRs: KRUSTY , GodivaMSRs: MSRE, UCB-Ciet , UW-Loop, . . .RCCS: NSTF, UW

-Loop, . . .

  • There are significant "gaps" : Validation is partial, with numerous tests in

-progress or planned.

  • More importantly, there is a lack of experimental data for some designs. CompletedIn-progressPlanned Slide 27Validation / Experimental "Gaps" 27GCRs:Prismatic gas

-cooled IET (i.e. HTTR, OSU

-HTTF)SFRs:Pool type IET data, International dataLMRs:Additional IET data, SET data for T/H, fuel, kinetics HPRs:Monolith conduction and heat release SET dataMSRs:Pool type IET data, natural circulation loop dataScaling of IETs and Range of Conditions of existing data to full

-scale prototypes remains to be established.

Slide 28Molten Salt Reactor (Inventory Control "Gap")Chemical ReactionPrimary FlowFission Product DecayCover Gas Gaseous Fission ProductsFission Product GenerationCoreFissile Material DepletionFission Product FilteringFuel Cycle FacilityFissile Material AdditionSolid MaterialPlateout, SedimentSystem FilteringCorrosion Product, Particulate Removal NRC non-LWR COMPUTER CODE DEVELOPMENT PLANS FOR SEVERE ACCIDENT PROGRESSION, SOURCE TERM, AND CONSEQUENCE ANALYSISOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionAugust 15, 2019ML19093B404https://www.nrc.gov/docs/ML1914/ML19143A120.pdf 2Design Basis Source Term Development Process(example: MOX & High Burnup Fuel)Fission Product TransportMELCOROxidation/Gas Generation Experimental BasisMelt ProgressionFission Product ReleasePIRT processAccident AnalysisDesign BasisSource TermScenario # 1Scenario # 2------.Synthesize timings and release fractionsCs Diffusivity

-1465 but prolonged release

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

-1Scenario # n------.D. Powers, et al. "Accident Source Terms for Light Water Nuclear Power Plants Using High

-Burnup or MOX Fuel", SAND2011-0128 January 2011 3FP ReleaseDeposition /CondensationResuspension/RevaporizationPrimary SystemSodiumConcreteInteractionSodium Fire & Aerosol GenerationFP ReleaseDeposition /CondensationPrimary SystemBubbleTransportCondensation & Dissolution of VaporsEntrainment & Dissolution of AerosolsResuspension/RevaporizationFP ReleaseDeposition /Condensation /ChemisorptionResuspension /RevaporizationPrimary SystemPhenomenology & Release Paths(common processes)Condensation /DepositionResuspension /EvaporationCondensation /Evaporation /AgglomerationContainmentLeak/FailureBubble TransportPool ScrubbingMolten CoreConcreteInteractionRelease of RNsand AerosolsHTGRLWRVessel LeakAir/Moisture IngressSFRVessel LeakVessel Leak FHRMSRVaporizationDeposition /CondensationPrimary SystemBubble Transport &Entrainment /RN VaporizationCondensation & Dissolution of VaporsEntrainment & Dissolution of AerosolsDeposition /CondensationFP ReleaseResuspension/RevaporizationVaporizationDeposition /CondensationPrimary SystemBubble Transport &Entrainment /RN VaporizationDeposition /CondensationResuspension/RevaporizationVessel LeakVessel Leak 4Development of Evaluation ModelLBE Transient AnalysisReactor PhysicsFHR/MSR onlyTritium production and sequestrationKineticsParametersPowerDistributionsIsotopic FPInventoryDecay HeatLibraryEvaluated Nuclear Data File(ENDF-B/VII+) Cross-Section Library Generation(AMPX)Reactor Physics Simulation(SCALE)Reactor-and State-specific Librariesfor Rapid AnalysisAccident Progression & Source TermConsequence Analysis(MACCS)Dose, Health Effects, Economic/Societal Consequences System Accident Analysis(MELCOR)HTGR/FHR MSRLWR/SFRLWR/SFR/HTGR/FHRLWR/SFR/HTGR/FHR/MSR 5MELCOR Input & Data RequirementsInput DataHTGRSFR MSRFHRFP InventorySCALESCALESCALESCALEFP diffusioncoefficients (D)and release Experiments (e.g., AGR) and analysis (e.g., DOE tools)ExperimentsExperiments (e.g., AGR) and analysis (e.g., DOE tools)Core powershapeRadial/Axial profiles (e.g., SCALE)Radial/Axial profiles (e.g., SCALE)Radial/Axial profiles (e.g., SCALE)Radial/Axial profiles (e.g., SCALE)Fuel failureExperiments/other codes (e.g., DOE tools)Experiments/other codes (e.g., DOE tools)Experiments/other codes (e.g., DOE tools)Dust generation & FP transportExperiments,historical data and other code (e.g., DOE tools)FP release under air/wateringress & interaction w/ graphiteExperimentsKinetics parameters and reactivity feedback coefficientsExperiments/other codes (e.g., SCALE)Experiments/other codes (e.g., SCALE)Experiments/other codes (e.g., SCALE)Experiments/other codes (e.g., SCALE)Equilibrium constants for release from pool and vapor pressuredataExperiments/other codes (e.g., DOE tools)Experiments/other codes (e.g., DOE tools)Experiments/other codes (e.g., DOE tools)Distribution of Cs-137 in different layers as a function of timeExperiments/AnalysisSCALE D S 6Power Distributions/ Limiting Operating ConditionsIsotopic Composition/Decay HeatSCALE (ORNL)PARCS (Univ. of Michigan)

  • Reactor Kinetics
  • Core Design/Follow
  • Flow Distribution
  • Peaking Factors
  • Reactivity CoefficientsCLAB(Sweden)HTC CRITICALS(France)FP CRITICALS(France)MALIBU/REBUSARIANECSNTAKAHAMADomestic ProgramsSt. LaurentMALIBU/REBUSIRPhEB/OECD:HTTR (Japan)HTR10 (China)PROTEUS (Switzerland)PBMR400Spent fuel calorimeter measurementsActinides reactivity worth in spent fuelFission products reactivity worth in spent fuelRadiochemical isotopic assays of HI BU SNFMOX cycle exposure dataNMSS/SFSTTechnical Basis for Decay Heat R.G. 3.54 NRRCatawba MOX Lead Test AssemblyNMSS/FCSSFuel Cycle AnalysisNMSS/SFSTHBU Fuel for Storage and Transportation Casks Criticality and Decay HeatHTGR criticality

-neutronics operating and safety parametersNRR/NROTechnical Basis for BUC for Spent Fuel Pool Criticality Safety NRONon-LWR reactors NRONew Reactor Design (ESBWR, AP1000, APR1400, NuScale, -)10CFR7210CFR7110CFR7210CFR50.68MELCORTRACEFASTNMSS/SFSTTechnical Basis for PWR BUC Interim Staff guidance ISG-8 & BWR BUC NRRLicensing Amendments(i.e., MELLLA+)

SCALE Development

  • Leveraging decades of physics models, nuclear data, and validation that can be extended to non

-LWRs-Most efficient approach to support accident progression and source term analysis

-For some technologies, the models are ready to be tested

  • Experimental Needs

-Decay heat, isotopic, validation data consistent with design and expected operating envelope-Criticality benchmarks

-Destructive assay data for new fuel forms (e.g.: TRISO)

  • Capabilities will be enhanced as more experience is gained, and gaps and uncertainties are quantified
  • Plan will be updated as more experience is gained and as new information regarding specific reactor design becomes available. Current focus on:

-How data transfer will work between SCALE and MELCOR/MACCS

-Moving fuel and power history presents challenges

  • Demonstrate the sufficiency of bounding analysis for licensing use

-Correct level of chemistry modelling between SCALE and MELCOR 7

8MACCS Overview

  • MACCS is the only code used in U.S. for probabilistic offsite consequence analysis
  • Highly flexible code that 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 9MACCS Code Development Areas for Non

-LWRs*Code development plans for site

-related issues

-Near-field atmospheric transport

  • Code development plans for design

-specific issues

-Radionuclide screening

-Radionuclide particle size distribution

-Radionuclide chemical form

-Radionuclide particle shape factor

-TritiumLloyd 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 Protecting Sensitive InformationAdvanced Reactors Stakeholder Meeting

-August 15, 2019Stu MagruderSenior Project ManagerAdvanced Reactor Licensing BranchDivision of Advanced Reactors Protecting Information

  • NRC must protect classified and sensitive information

-Classified information

-Safeguards Information

-Sensitive unclassified non

-safeguards information SUNSI (e.g., proprietary, security

-related, export controlled information (ECI))Note: The NRC does not designate ECI. ECI designation should be coordinated with appropriate federal agency (e.g., Department of Energy, Department of Commerce).

2 SUNSI Categories* Allegation information* Investigation information* Critical Electric Infrastructure Information (CEII)* Security

-related information* Proprietary information* Privacy Act information* Federal-, State-, foreign government

-, and international agency

-controlled information (ECI)* Sensitive internal information 3

Requests for Withholding

  • Per 10 CFR 2.390, prospective applicants may request that proprietary information be withheld from public disclosure
  • Requests for withholding must be accompanied by an affidavit

-Affidavit should be either notarized or signed under oath or affirmation

-Identify what information is considered proprietary

-Explain why the release of information would cause harm

  • Sensitive information, including proprietary information and ECI should include appropriate portion and page markings
  • Non-proprietary (public) versions of documents should be provided with proprietary submittals
  • NRC staff will evaluate requests and determine whether information should be withheld from public disclosure 4

Marking GuidanceMARKINGWhat documents should be marked?

  • Mark all documents containing Trade Secrets or Confidential Commercial or Financial Information.
  • Do not mark documents from INPO designated INPO Private.Who may authorize document marking?NRC recipient or originator (or supervisor) pursuant to 10 CFR 2.390.How should a document be marked?NRC Generated Documents
  • The top and bottom of each page should be marked

-"Official Use Only

-Proprietary Information."Incoming Documents

  • Marking requirements are defined in 10 CFR 2.390(b) and require marking only at the top of page, and each successive page containing proprietary Information, and adjacent to the specific proprietary information.When is portion or page marking required?
  • Required for all documents.
  • If the entire page is not affected, indicate the basis (i.e., trade secret, etc.) for the designation adjacent to the protected information. See 10 CFR 2.390 (b)(1)(i)(B).5 Staff Guidance
  • NRC Office Instruction LIC

-204, "Handling Requests to Withhold Proprietary Information from Public Disclosure." (ADAMS Accession No. ML093240489)

-Provides specific information on reviewing and dispositioning requests to withhold proprietary information

-Publicly available

-recommend reviewing before submitting documents 6

Additional Thoughts

  • Take care with redacting process
  • Expectations for level of detail in withholding may change with maturity of application
  • What about requests for withholding information during public meetings?
  • "No comment" policy for staff re SUNSI 7

Why is this Important?

  • Final NRC records and documents are generally made public per 10 CFR 2.390(a)

-Balance interests of industry and public

-Documentation for withholding is important 8

Questions?

9This Photoby Unknown Author is licensed under CC BY-SA

Retrieved from "https://kanterella.com/w/index.php?title=ML19228A263&oldid=1882920"