ML19228A263

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August 15, 2019 Advanced Reactors Stakeholders Meeting Presentation Slides
ML19228A263
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Issue date: 08/15/2019
From: Vechioli-Feliciano L
Office of New Reactors
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Vechioli L, NRO/DAR/ARLB, 415-6035
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Public Meeting on Possible Regulatory Process Improvements for Advanced Reactor Designs August 15, 2019 Telephone Bridge: (800) 593-0695 Passcode: 3798624 1

Public Meeting

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

Outline 9:00-9:10 am Opening Remarks 9:10-9:40 am Status Update 9:40-10:25 am Technology Inclusive Content Applications Project (TICAP) 10:25-10:35 am Break 10:35-11:00 am Micro-Reactors Regulatory Topics 11:00-11:30 am Concrete and Elevated Temperature 11:30-12:15 pm Environmental Interim Staff Guidance and Generic EIS Developments 12:15-1:15 pm Lunch 1:15-1:45 pm 10 CFR Part 53: Risk-informed, Technology Inclusive Regulatory Framework for Advanced Reactors Rulemaking 1:45-2:15 pm Nuclear 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 pm Break 3:30-4:00 pm Protecting Sensitive Information 4:00-4:30 pm Open Discussion 3

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

- Amir Afzali, Southern Company Services 5

Break Meeting/Webinar will begin shortly Telephone Bridge 800-593-0695 Passcode: 3798624 6

  • 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

Lunch Meeting/Webinar will begin shortly Telephone Bridge 800-593-0695 Passcode: 3798624 10

  • 10 CFR Part 53: Risk-Informed, Technology Inclusive Regulatory Framework for Advanced Reactors Rulemaking

- William Reckley, NRC 11

- NRC Section 103 Activities

- Establishing Metrics and Milestones

  • 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, NRC 13

Break Meeting/Webinar will begin shortly Telephone Bridge 800-593-0695 Passcode: 3798624 14

  • Protecting Sensitive Information

- Stewart Magruder, NRC 15

2019 Tentative Schedule for Periodic Stakeholder Meetings October 10 December 12 16

Open Discussion and Closing 17

TICAP Technology Inclusive Content of Application Project Amir Afzali Southern Company NRC Advanced Reactor Public Meeting Briefing August 15, 2019

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 all potential 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 Commissions 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 NRCs interaction (e.g., WP for NRCs official review) not yet defined.
  • Necessary Event
  • Regulatory Guide endorsing NEI 18 Expected Date Dec. 2019 3

TICAP Success Criteria and Guiding Principles Success Criteria Provide 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.

Guiding Principles

  • Right-size content of application to be commensurate with the complexity of the applicants safety case.
  • Improve the overall efficiency of review and optimize generation of applications information in terms of scope and level of details.
  • Is limited to utilizing the outputs developed through application of the Licensing 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, 3rd quarter).
  • Should contribute beneficially to the development of the NRCs new regulatory framework for advanced reactors (10 CFR Part 53) 4

Siting near Functional densely populated Containment areas EP for SMRs (SECY-18-0096) and ONTs Licensing Modernization (SECY-18-0103)

Project Insurance and Liability Simplified Environmental Reviews representation role of Consequence LMP within regulatory Based Security (SECY-18-0076) framework 5

Fundamental Consideration Based on Use of LMP Process TICAP 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 Standard Advanced Reactor Stakeholder Meeting August 15, 2019

Outline of Plan

  • Objectives and Scope
  • Tasks Task 1: Supporting development of the standard Task 2: Preparation for review of the standard Task 3: Reviewing the standard Task 4: Maintaining PRA standard Task 5: Development of schedule Task 6: Identification of resources Task 7: Development of communication plan August 15, 2019 Page 2

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 applications Define 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 issues Review each technical element associated with each risk level and hazard, for each application type, and identify possible technical or policy issues Describe the significance of the issues Identify 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 review Criteria for determining acceptance (no objection)

August 15, 2019 Page 3

Status of NRC RG, ASME/ANS Standard, and NEI Guidance Under development

  • Initiated staff position NRC Regulatory Position (RP ) ( RG x.xxx )

PRA Acceptability Consensus Standard Peer Review to demonstrate to demonstrate conformance with RP conformance with Standard Under development Under development ?????

  • Trial Use issued
  • Final, publication in 2020/2021 August 15, 2019 Page 4

Micro-Reactors Regulatory Topics Marc Nichol, Director New Reactor Deployment August 15, 2019

©2019 Nuclear Energy Institute

Micro-Reactor Regulatory Issues Priority Issues Addressed in Broader Non-Urgent Efforts

1. Review Scope, Duration,
  • Siting
  • Transportation Level of Effort
  • Environmental Reviews
  • Annual Licensee Fees
2. Aircraft Impact
  • Fuel
3. Operations (auto/remote)
  • Generic License
4. Resident Inspector
5. Physical Security
6. Emergency Preparedness No issues identified to-date
  • Liability Insurance
  • Decommissioning Funding

©2019 Nuclear Energy Institute 2

Unique Micro-Reactor Considerations*

Typically 1 MWe to 10 Mwe Very small size

  • Site <0.1 acres, building ~size of a house, reactor fits in shipping container Very small potential consequences
  • Source terms as low as 1% of todays reactors
  • Fail-safe: shuts itself off, cannot meltdown
  • Proliferation resistant fuel Operational simplicity
  • Few to zero moving parts
  • Automatic operations
  • Minimal maintenance
  • General description, all features may not be applicable to all designs ©2019 Nuclear Energy Institute 3

Environmental Interim Staff Guidance and Generic EIS Developments Donald Palmrose, PhD Mallecia Sutton Sr. Reactor Engineer Sr. Project Manager Office of New Reactors U.S. Nuclear Regulatory Commission Advanced Reactor Stakeholders Meeting August 15, 2019 International Conference on Radioecology and Environmental Radioactivity, Bergen, 1 Norway

Agenda

  • Staff seeking input on:

- Draft ISG

- GEIS

- Possible EA for some Advanced Reactors

  • Open discussion 2

Advanced Reactor Stakeholders Meeting on August 15, 2019

Interim Staff Guidance Environmental 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 3

Advanced Reactor Stakeholders Meeting on August 15, 2019

Interim Staff Guidance (cont.)

  • 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 NRCs NEPA obligations
  • Issuance of a draft ISG for comment and use by December 2019 4

Advanced Reactor Stakeholders Meeting on August 15, 2019

Considerations for Advanced Reactor Reviews

- 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 5

Advanced Reactor Stakeholders Meeting on August 15, 2019

Considerations for Advanced Reactor Reviews

  • Can an EA address some advanced reactor reviews

- 10 CFR 51.21

- Rulemaking needed?

- Considering staff guidance 6

Advanced Reactor Stakeholders Meeting on August 15, 2019

Open Discussion 7

Advanced Reactor Stakeholders Meeting on August 15, 2019

Concrete and Elevated Temperature Madhumita Sircar*, Jose Pires*, Ata Istar**

U.S. Nuclear Regulatory Commission

  • Office of Nuclear Regulatory Research
    • Office of New Reactors Advanced Reactors Stakeholders Meeting Nuclear Energy Institute Washington, DC.

August 15, 2019

Scope of Presentation:

  • Effects of High Temperature on Concrete Current Code Requirements by ACI 349:
  • Provision E.4, Concrete Temperature, 150oF for Normal Operations 350oF 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 Temperature Temperature, °C Temperature, °C Ultimate Compressive Strength, psi Ultimate Compressive Strength, MPa Modulus of Elasticity, 106 psi Modulus of Elasticity, MPa Temperature, °F Temperature, °F Ultimate compressive strength and modulus of elasticity of Type I Portland cement paste (w/c:0.33) 6

[Source: Harmathy et 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]

Concrete Under High Temperature, cont.

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.

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.

[Source: Fig. 2.6 as referenced in NUREG/CR-7031]

11

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, Technology Inclusive Regulatory Framework for Advanced Reactors Rulemaking Advanced Reactor Stakeholder Meeting August 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
  • Todays meeting marks first staff outreach to external stakeholders
  • A focused public workshop is being planned for October 2019

Rule Applicability

  • NEIMAs 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?

Developing Functional Requirements Next Generation Nuclear Plant (NGNP) Concepts

Licensing Modernization Project Figure 4.2. Definition of Risk-Significant and Safety-Significant SSCs 8

Integrated Approach Siting near densely populated Functional areas EP for SMRs Containment and ONTs Licensing Modernization Project Insurance and Liability Environmental Reviews Consequence Based Security 9

NRC Path Forward

  • Draft a rulemaking plan, taking into consideration todays 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.gov website by searching for the NRC Docket ID NRC-2019-0062

Nuclear Energy Innovation and Modernization Act:

Establishing Metrics and Milestones Steven Lynch Acting Chief, Advanced Reactor Licensing Branch Division of Advanced Reactors 1

Background

  • 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

  • Generic milestone schedules for requested activities are provided on the NRC public webpage:

https://www.nrc.gov/about-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

- 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 Activities John Segala Chief, Advanced Reactor Policy Branch Division of Advanced Reactors August 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-93-092, NRC contract with INL to develop guidance)
  • Other Policy Issues (Siting as it relates to population, Physical Security, and Micro Reactors)
  • Coordination and stakeholder input

- Public Stakeholder meeting on March 28th

  • 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

RES Implementation Action Plan for Advanced Non-LWR ; Codes and Tools Non-LW R Vision and Strategy Com puter Code R eports M odeling and Sim ulation of non-LW Rs 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 Stakeholders Meeting August 15, 2019

2 NRCs Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Strategy 4 Knowledge, Skills, Industry Codes and and Capacity Standards Strategy 2 Strategy 5 Technology Inclusive Computer Codes Issues Strategy 3 Strategy 6 Flexible Review Communication Process Slide 2

NRCs Implementation Action Plan, Strategy 2 - Computer Codes ML19093B266 ML19093B322 ML19093B404 Introduction Volume 1 Volume 3 Volume 2 = Fuel Performance Volume 4 = Radiation Protection Under Development 4

Event Selection

  • 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 4 4

Some Recent & Upcoming Events . . .

  • 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 5 5

Volume 1 Design Basis Event Analysis Slide 6 6

Introduction / Outline

  • Volume 1 Design Basis Event Analysis :

- Phenomena Identification and Ranking Tables (PIRTs)

- Event Scenarios

- New Physical Phenomena for non-LWRs

- Gaps

- Tasks Slide 7 7

Slide 8 Characterization of Design Types Plant Description Fuel Type No.

1 HTGR; prismatic core, thermal spectrum TRISO (rods or plates) 2 PBMR; pebble bed core, thermal spectrum TRISO (pebbles) 3 GCFR; prismatic core, fast spectrum SIC clad UC (plates) 4 SFR; sodium cooled, fast spectrum Metallic (U-10Zr) 5 LMR; lead cooled, fast spectrum Not available.

(Possibly nitride fuel.)

6 HPR; heat pipe cooled, fast spectrum Metallic (U-10Zr) 7 MSR; prismatic core, thermal spectrum TRISO (plates) 8 MSPR; pebble bed, thermal spectrum TRISO (pebbles) 9 MFSR; fluoride fuel salt, thermal/epithermal spectrum Fuel salt 10 MCSR; chloride fuel salt, fast spectrum Fuel salt Slide 9

Modeling Gaps Identified by PIRTs

  • 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 10 10

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 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 11 11

Comprehensive Reactor Analysis Bundle (CRAB)

SCALE SERPENT MAMMOTH Cross-sections Cross-sections Neutronics FLUENT CFD PRONGHORN Core T/H PARCS Nek5000 CFD Neutronics TRACE MOOSE System T/H BISON SAM System and Core T/H MELCOR Fuel Performance Containment / FP FAST MAMMOTH Fuel Performance Neutronics SERPENT NRC Code Intl Code Commercial Cross-sections DOE Code Slide 12

Approach to Validation (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 nodalization scheme and modeling options.

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

Slide 13 13

Summary & 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 product than can be attained with further development of only the NRCs conventional LWR codes.

Slide 14

Extra Stuff Slide 15

Codes for Design Basis Event Analysis

  • 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 16 16

Unique Capabilities Available in CRAB

  • Examples

- Multiphysics Coupling

- Geometric Fidelity

- Advanced Equivalence Methods

- Multi-Scheme Capability

- 3D Reduced Order Flow Model Goal: Enable analysis of advanced designs without over-simplifying assumptions to provide intermediate fidelity model for modest computational resources.

Slide 17 17

Multiphysics Coupling SAM: System Level Thermo-Fluids MAMMOTH: Rx Kinetics Temperatures & Densities Power Temperatures Displacements Tensor Mechanics Module Slide 18 18

Geometric Fidelity PRONGHORN Mesh HTR-10 D Channels for upcomer & CRs Bottom Reflector & Conus D Porous body for conus Slide 19 19

Advanced Equivalence Methods HTR-PM HTTR

  • Pebble bed HTR (420k pebbles)
  • Prismatic HTR
  • Diffusion problem size - 54,656 cells
  • Diffusion problem size - 15,552 cells HTR-PM keff pcm RMS % Err Max %Err HTTR keff pcm RMS % Err Max %Err (Tfuel = 1100 K) f f ( Tfuel = 1300 K) Power Power Serpent 1.01159 +/-1.8 - - Serpent 1.00259 +/-2.7 - -

Diffusion 1.03653 2435 6.0 40.6 Diffusion 1.01978 1715 3.12 6.20 SPH-Diffusion 1.01159 0 1.55E-03 2.94E-03 SPH-Diffusion 1.00259 0 7.0E-02 2.0E-01 Transport level accuracy for the price of a diffusion calculation Slide 20 20

Multi-Scheme Capability

  • MAMMOTH

- Allows usage of transport where more detail is needed with efficiency of diffusion for remainder of domain Slide 21 21

SAM: 3-D Flow Model

  • Validation Examples Natural Convection Ra = 105 Lid-Driven Cavity Flow Slide 22 22

Verification & Validation Gaps Slide 23 23

Code Assessment Issues

  • Code Assessment = Verification & Validation represents 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 24 24

Code Validation Matrix

  • 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 25 25

Validation Status GCRs: HTR-10, PBMR-268,-400, SANA, HTTU, AVR, . . .

SFRs: EBR-II, FFTF, CEFR, ZPPR, Monju, . . .

LMRs: Helios HPRs: KRUSTY, Godiva Completed MSRs: MSRE, UCB-Ciet, UW-Loop, . . . In-progress RCCS: NSTF, UW-Loop, . . . Planned

  • 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.

Slide 26 26

Validation / Experimental Gaps GCRs: Prismatic gas-cooled IET (i.e. HTTR, OSU-HTTF)

SFRs: Pool type IET data, International data LMRs: Additional IET data, SET data for T/H, fuel, kinetics HPRs: Monolith conduction and heat release SET data MSRs: Pool type IET data, natural circulation loop data Scaling of IETs and Range of Conditions of existing data to full-scale prototypes remains to be established.

Slide 27 27

Molten Salt Reactor (Inventory Control Gap)

Cover Gas Gaseous Fission Products System Filtering Corrosion Product, Particulate Removal Fission Product Generation Chemical Reaction Fission Product Filtering Core Primary Flow Fuel Cycle Facility Fissile Material Depletion Fission Product Decay Fissile Material Addition Solid Material Plateout, Sediment Slide 28

NRC non-LWR COMPUTER CODE DEVELOPMENT PLANS FOR SEVERE ACCIDENT PROGRESSION, SOURCE TERM, AND CONSEQUENCE ANALYSIS https://www.nrc.gov/docs/ML1914/ ML19093B404 ML19143A120.pdf Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission August 15, 2019

Design Basis Source Term Development Process (example: MOX & High Burnup Fuel)

Experimental Basis PIRT process Oxidation/Gas Generation Melt Progression Fission Product Release Fission Product Transport Accident Analysis Design Synthesize MELCOR Scenario # 1 Scenario # 2 timings and Basis

. . release Source fractions Term Scenario # n-1 Scenario # n D. Powers, et al. Accident Source Terms for Light Water Nuclear Power Plants Using High-Burnup or MOX Fuel, SAND2011-0128 January 2011 Cs Diffusivity

  • Differences not from change of fuel but from code advances 2

Phenomenology & Release Paths (common processes)

Condensation / Condensation / Resuspension /

Evaporation / Deposition Evaporation Containment Agglomeration Leak/Failure LWR HTGR SFR FHR MSR Primary System Primary System Primary System Primary System Primary System Deposition / Deposition / Deposition /

Condensation / Deposition / Deposition / Condensation Condensation Chemisorption Condensation Condensation FP Release Resuspension/ Resuspension/

Revaporization Revaporization Resuspension/

Resuspension / Revaporization Revaporization Resuspension/

Bubble FP Release Bubble Revaporization Transport & Transport &

Bubble Entrainment / Entrainment /

RN Vaporization RN Vaporization FP Release Transport FP Release Condensation & Condensation &

Dissolution of Vapors Dissolution of Vapors Entrainment & Dissolution Entrainment & Dissolution of Aerosols of Aerosols Vessel Leak Vessel Leak Vessel Leak Vessel Leak Deposition / Vaporization Deposition / Vaporization Release of RNs Sodium Fire & Condensation Condensation Bubble Transport Aerosol Generation and Aerosols Pool Scrubbing Vessel Leak Air/Moisture Ingress Molten Core Sodium Concrete Concrete Interaction Interaction 3

Reactor Physics Kinetics Evaluated Nuclear Data File Parameters (ENDF-B/VII+)

Power Distributions Development of Cross-Section Library Generation Decay Heat (AMPX)

Library Evaluation Model Isotopic FP Reactor Physics Simulation Inventory (SCALE)

FHR/MSR only Tritium production Reactor- and State-specific Libraries and sequestration for Rapid Analysis LBE Transient Analysis System Accident Analysis (MELCOR)

Accident Progression

& Source Term Consequence Analysis (MACCS)

Dose, Health Effects, Economic/Societal Consequences HTGR/FHR LWR/SFR/HTGR/FHR LWR/SFR LWR/SFR/HTGR/FHR/MSR MSR 4

MELCOR Input & Data Requirements Input Data HTGR SFR MSR FHR FP Inventory SCALE SCALE SCALE SCALE FP diffusion coefficients Experiments (e.g., AGR) Experiments Experiments (e.g., AGR)

(D) and release and analysis (e.g., DOE and analysis (e.g., DOE tools) tools)

Core power shape Radial/Axial profiles Radial/Axial profiles Radial/Axial profiles Radial/Axial profiles (e.g., SCALE) (e.g., SCALE) (e.g., SCALE) (e.g., SCALE)

Fuel failure Experiments/other codes Experiments/other codes Experiments/other codes (e.g., DOE tools) (e.g., DOE tools) (e.g., DOE tools)

Dust generation & FP Experiments, historical transport data and other code (e.g., DOE tools)

FP release under Experiments air/water ingress &

interaction w/ graphite Kinetics parameters and Experiments/other codes Experiments/other codes Experiments/other codes Experiments/other codes reactivity feedback (e.g., SCALE) (e.g., SCALE) (e.g., SCALE) (e.g., SCALE) coefficients Equilibrium constants for Experiments/other codes Experiments/other codes Experiments/other codes release from pool and (e.g., DOE tools) (e.g., DOE tools) (e.g., DOE tools) vapor pressure data SCALE Distribution of Cs-137 in D S different layers as a function of time Experiments/Analysis 5

Spent fuel calorimeter SCALE (ORNL) NMSS/SFST Technical Basis for measurements 10CFR72 Decay Heat R.G. 3.54 CLAB NMSS/SFST Technical Basis for PWR (Sweden) 10CFR71 BUC Interim Staff guidance 10CFR72 ISG-8 & BWR BUC Actinides reactivity worth in spent fuel NRR HTC Catawba MOX Lead Test CRITICALS Assembly (France) Fission products NRR/NRO reactivity worth in Technical Basis for BUC spent fuel 10CFR50.68 for Spent Fuel Pool FP CRITICALS Criticality Safety (France)

NRR Radiochemical isotopic assays of HI BU SNF Licensing Amendments MALIBU/REBUS (i.e., MELLLA+)

ARIANE CSN TAKAHAMA NMSS/FCSS Domestic Fuel Cycle Analysis Programs MOX cycle exposure data NMSS/SFST HBU Fuel for Storage St. Laurent and Transportation MALIBU/REBUS Casks Criticality and PARCS (Univ. of Michigan)

HTGR criticality -

  • Reactor Kinetics Decay Heat neutronics operating
  • Core Design/Follow and safety parameters NRO
  • Flow Distribution IRPhEB/OECD: Non-LWR reactors
  • Peaking Factors HTTR (Japan)
  • Reactivity Coefficients NRO HTR10 (China)

PROTEUS Power Distributions/ New Reactor Design (Switzerland) Isotopic Composition/Decay Heat Limiting Operating Conditions (ESBWR, AP1000, PBMR400 APR1400, NuScale, )

MELCOR TRACE FAST 6

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

MACCS 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, uncertainty MACCS Gaussian plume segment ATD model animation for a single weather trial 8

MACCS Code Development Areas for Non-LWRs

  • Code development plans for site-related issues

- Near-field atmospheric transport 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

  • Code development plans for design-specific issues

- Radionuclide screening

- Radionuclide particle size distribution

- Radionuclide chemical form

- Radionuclide particle shape factor

- Tritium 9

Protecting Sensitive Information Advanced Reactors Stakeholder Meeting - August 15, 2019 Stu Magruder Senior Project Manager Advanced Reactor Licensing Branch Division 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 Guidance MARKING What documents should *Mark all documents containing Trade Secrets or Confidential Commercial or Financial Information.

be marked? *Do not mark documents from INPO designated INPO Private.

Who may authorize NRC recipient or originator (or supervisor) pursuant to 10 CFR 2.390.

document marking?

How should a document NRC Generated Documents:

be marked? *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 *Required for all documents.

marking required? *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?

This Photo by Unknown Author is licensed under CC BY-SA 9

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