ML20274A057

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October 1, 2020 Advanced Reactor Stakeholder Meeting Presentation Slides
ML20274A057
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Issue date: 10/01/2020
From: Jordan Hoellman
NRC/NRR/DANU/UARP
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Hoellman J, NRR/DANU/UARP, 301-415-5481
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Advanced Reactor Stakeholder Public Meeting October 1, 2020 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 894 701 300#

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Time Agenda Speaker 10:00 - 10:20 am Opening Remarks NRC 10:20 - 11:30 Update on Regulatory Analysis Review of Applicable B. Travis, NRC am Regulations for Non-Light-Water Reactors 11:30 am - Status Update on the Advanced Reactor Generic M. Sutton and J.

12:00 pm Environmental Impact Statement (GEIS) Cushing, NRC 12:00 - 1:00 pm BREAK All Discussion of Advanced Reactor Fuel Qualification T. Drzewiecki, 1:00 - 2:00 pm White Paper NRC Assessment of the MACCS Code Applicability for 2:00 - 2:30 pm D. Clayton, SNL Nearfield Consequence Analysis Update on NRC Endorsement of ORNL Report on 2:30 - 2:45 pm Preparing and Reviewing a Molten Salt Non-Power W. Kennedy, NRC Reactor Application 2:45 - 3:00 pm Concluding Remarks and Future Meeting Planning NRC/All 2 of 82

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced.html 3 of 82

NRC Staff Draft White Paper Analysis of Applicability of NRC Regulations for Non-Light Water Reactors October 2020 4 of 82

Purpose

  • Discuss NRC Staff Draft White Paper - Analysis of Applicability of NRC Regulations for Non-Light Water Reactors
  • Provide staff position on the presumed applicability of various regulations to non-LWR applicants under either Part 50 or Part 52
  • Discuss expected exemptions from regulations identified by NRC staff in specific areas applicable to non-LWRs 2

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Background

  • NRC published Draft Non-Light Water Review Strategy Staff White Paper in September 2019
  • Intended to facilitate discussion on possible approaches for NRC staff review of the licensing basis information of non-LWR applications independent of the specific design or methodology; no plans to finalize.
  • Attachment 1 of that review strategy provided preliminary NRC position on regulatory applicability for discussion 3

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Analysis

  • Providing an expanded standalone version of regulatory applicability was perceived as a priority, so NRC staff reviewed existing regulations with an eye towards non-LWR reactors
  • White paper presents the NRC staffs generic analysis of regulations; does not constitute a new interpretation.
  • Regulations were evaluated generically. If it was not possible to preclude all non-LWR designs from the regulation, it was denoted as applicable.
  • In a few cases, the NRC conclusion regarding applicability differs from the Appendix of the draft review strategy as a result of a more rigorous review 4

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Analysis

  • As part of any future application review, NRC staff will continue to evaluate current NRC regulations that do not apply to non-LWR designs to ensure that any particular non-LWR design satisfactorily meets adequate protection of public health and safety or common defense and security
  • NRC staff acknowledges that some of the regulations identified as applicable in the tables may not be required to meet the underlying purpose of the rule when applied for certain non-LWR designs.

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Exemptions

  • Applicants may request exemptions from the NRCs regulations on a case-by-case basis
  • In reviewing an exemption request, the NRC must determine that the proposed exemption

- is authorized by law,

- will not present an undue risk to the public health and safety,

- is consistent with the common defense and security.

- meets at least one special circumstance identified in 10 CFR 50.12(a)(2).

  • The NRC staff anticipates that non-LWRs applicants will request exemptions from some of the NRCs regulations. This should not be perceived as a negative.

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Exemptions

  • An exemption request may not always be required - in many cases, NRC staff expects that non-LWR designs may meet a rule through design- and application-specific implementations.
  • In other cases, such as in applying definitions or listing codes and standards, the regulations may be applicable but not impose a requirement.
  • For straightforward exemptions, NRC staff believes it is possible to simplify the applicants exemption process (example preceding Table 2 in the white paper)

- Involves NRC staff applying 50.12 or 52.7, as appropriate 7

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Tables Table Title Purpose - Intended Use For all non-LWR applicants; Provides 1 additional context regarding probable Areas with anticipated exemptions exemption areas For Part 52 applicants - application of Part 2

Part 52 Regulations Referencing Part 50 52 requirement differs from Part 50 for Regulations Limited to LWRs non-LWRs 3 10 CFR Part 50 Requirements, as applicable to applications under Part 50 for non-LWRs For Part 50 non-LWR applicants Provides a list of TMI-items including entry 4

Applicability of 10 CFR 50.34(f) TMI conditions for technical relevancy listed Requirements to non-LWRs under Part 52 for some items 5

Selected 10 CFR Part 52 Requirements, as applicable to non-LWR DCs, COLs, and SDAs For Part 52 non-LWR applicants Other regulations (excluding 10 CFR Parts 50 and 6

52) that may apply to non-LWRs at some stage in licensing For all non-LWR applicants 8

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Table 1: Fission Product Release

  • This language is LWR-centric and the prescriptive nature of the implementation of the regulation is not consistent with Commission policy regarding non-LWRs; underlying purpose of the rule remains applicable
  • One focus of the regulation is on demonstrating mitigation of consequences of a radiological release (vs. prevention); Not intended to be a non-representative fission product release derived from LWR operation, however
  • Possible approaches could include:

- Using LMP and the associated comprehensive evaluation of potential events to select sequences for consequence evaluation

- Demonstrably limiting event sequence(s) (not limited to design basis) with corresponding mechanistic evaluation of consequences 9

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Table 1: Criticality Requirements

  • 10 CFR 50.68 provides two options for monitoring to detect criticality for stored nuclear material.

- 10 CFR 70.24

- 10 CFR 50.68(b), which contains LWR-specific considerations for fuel storage

  • Non-LWR fuel differs significantly in form from traditional fuel types used in LWRs and in many cases have higher enrichment.
  • Staff expects that applicants may elect to provide an alternative to 10 CFR 70.24 similar to 50.68(b) for their specific fuel type.

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Table 1: Reactor Coolant Pressure Boundary

  • The reactor coolant pressure boundary for an LWR provides a fission product retention barrier for the release of radionuclides.
  • However, in some non-LWRs, the reactor coolant boundary may not serve this function - may instead use a functional containment.
  • For these designs, references to the integrity of the reactor coolant pressure boundary may not be applicable and an exemption is anticipated.
  • How these exemptions are addressed will depend on the importance of a coolant boundary to demonstrating the specific reactors safety.

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Table 2 - Part 52 References to Part 50

  • Table 2 provides a list of the regulations in 10 CFR Part 52 that apply to all power reactors, but reference a 10 CFR Part 50 regulation that refers specifically to LWRs.
  • To address the regulations as written, an exemption would be required.
  • Justification for these exemptions should be straightforward as the referenced 10 CFR Part 50 regulations do not apply to non-LWRs.
  • Without rulemaking, this issue cannot be resolved generically, so in the white paper NRC staff has proposed a streamlined exemption path 12 15 of 82

Table 2 - Part 52 References to Part 50

  • In order to process these exemptions expeditiously, staff requests applicants provide the following:

- a statement that the design need not comply with the requirements of a specific subsection or subsections below; and

- a statement or reference to associated docketed application material explaining why the design need not comply with the regulation (e.g., a design overview that makes it clear the reactor is not an LWR and the technology type employed need not include the safety function required by the regulation or accomplishes a required safety function through a means other than that required by the regulation).

  • In performing the exemption review, NRC staff will provide the information necessary to comply with the applicable exemption regulations in 10 CFR 52.7 and 50.12 13 16 of 82

Tables 3, 5 and 6 - Regulatory Applicability

  • Table 3 provides a list of regulations with presumed applicability for non-LWR designers applying under 10 CFR Part 50.
  • Table 5 provides a list of regulations to be considered by non-LWR designers applying under 10 CFR Part 52.
  • Table 6 provides a list of regulations outside of 10 CFR Part 50 and Part 52 that may apply to non-LWRs. This list does not assess specific regulations within each part.
  • Although the lists are intended to be comprehensive, lack of inclusion of any regulation should not be interpreted as a non-applicability.

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Table 4 - TMI Requirements

  • TMI-related requirements written considering LWRs (operating experience and lessons learned):

- Some requirements are clearly only technically relevant for specific reactor types.

These were omitted from the table;

- Some requirements may be narrowly applicable to some technology types but not to others - staff provided entry conditions for some of these, and

- Some requirements are general high-level programmatic requirements that are technology-neutral.

  • Technically relevant as it applies in this case allows for a greater degree of flexibility. If the requirement in question can be justified as not technically relevant to a design under review, the requirement is satisfied without a need for an exemption.
  • In some cases, may also be met through application of a different regulation (e.g. Appendix E requirements)
  • An exemption may nonetheless be required in some cases 15 18 of 82

Questions/Discussion 16 19 of 82

Status Update on the Advanced Reactor Generic Environmental Impact Statement Jack Cushing Senior Environmental Project Manager Environmental Center of Expertise 20 of 82

Status of ANR GEIS

  • Staff Requirement Memorandum 20-0020 issued September 21, 2020 directing the staff to codify the results of the ANR GEIS (ADAMS Accession No. ML202065A112)
  • Scoping Summary Report issued on September 25, 2020 (ADAMS Accession No. ML20260H180)
  • Staff is drafting sections of the GEIS 21 of 82 2

Scoping Summary Report

  • GEIS will use a technology neutral, performance-based plant parameter envelope (PPE) and site parameter envelope (SPE) approach that is inclusive of as many advanced reactor technologies as possible.
  • Power level will not be used in most resource areas.

Reactor of any size can use the GEIS provide that it is bounded by the performance measures and assumptions.

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Scoping Summary Report cont.

  • Reactor applications can reference individual resources that it meets and evaluate the ones it does not meet in its application.
  • Goal is to develop an effective GEIS to disposition generically as many issues as practicable.

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Advanced Reactor GEIS Status

  • The staff is evaluating the schedule impacts of rulemaking on issuance of the GEIS.

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Questions 25 of 82 6

Break Meeting/Webinar will resume shortly Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 894 701 300#

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Fuel Qualification (FQ) for Advanced Reactors Advanced Reactor Stakeholder Meeting October 1, 2020 27 of 82 1

Outline Background/Review of past engagement Activity affecting FQ guidance FQ report overview Incorporation of comments from May 7th stakeholder meeting Next steps and discussion 28 of 82 2

NEIMA SEC. 103. ADVANCED NUCLEAR REACTOR PROGRAM (c) REPORT TO INCREASE THE USE OF RISK-INFORMED AND PERFORMANCE-BASED EVALUATION TECHNIQUES AND REGULATORY GUIDACNE 29 of 82 3

Regulatory Aspects of Nuclear Fuel Qualification Regulations No requirements specific to nuclear fuel qualification Requirements on fuel qualification are provided by top level requirements attributed to the facility 10 CFR 50.43(e) - Demonstrate safety features (e.g., data)

GDC/ARDC 2, Design bases for protection against natural phenomena GDC/ARDC 10, Reactor design Coolable Geometry/Dose:

GDC 27/ARDC 26 - Reactivity control systems GDC/ARDC 35 - Emergency core cooling system 10 CFR 50.34(a)(1)(ii)(D), 10 CFR 52.47(a)(2)(iv), and 10 CFR 52.79(a)(1)(vi) 30 of 82 4

Regulatory Aspects of Nuclear Fuel Qualification

  • Guidance

- NUREG-0800, Standard Review Plan

  • Section 4.2, Fuel System Design

- Identifies acceptance criteria derived from know fuel failure/degradation mechanisms for light water reactor fuel

- ATF-ISG-2020-01

  • Significant changes to fuel design must be assessed for potentially new failure/degradation mechanisms

- Reg Guide 1.233, Licensing Modernization

  • Emphasis on risk - requires understanding of accident sequence consequences (i.e., source term) 31 of 82 5

May 7, 2020 Advanced Reactor Stakeholder Meeting

  • NRC presented framework:

- Top-down approach with ~58 terminal goals identified

- Supporting/clarifying language was still being developed

- Standards for evidence with clarifying examples were still being developed

  • Input from stakeholders has been incorporated into the draft report:

- Union of Concerned Scientists (UCS)

- United States Nuclear Industry Council (USNIC)

- Southern Nuclear Company (SNC)

- Kairos Power

- Public stakeholder 32 of 82 6

Outline Background/Review of past engagement Activity affecting FQ guidance FQ report overview Incorporation of comments from May 7th stakeholder meeting Next steps and discussion 33 of 82 7

FQ Activity NRC reviewed and approved:

EPRI-AR-1, "Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance," May 2019 ANL/NE-16/17, Rev. 1, "Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification," May 2019 NRC supported activity:

MSR fuel qualification:

ORNL/LTR-2018/1045, "Molten Salt Reactor Fuel Qualification Considerations and Challenges," 2018 (ML18347A303)

ORNL/TM-2020/1576, "MSR Fuel Salt Qualification Methodology," 2020 (ML20197A257)

Source term:

SAND2020-0402, "Simplified Approach for Scoping Assessment of Non-LWR Source Terms," 2020 (ML20052D133)

INL/EXT-20-58717, "Technology-inclusive determination of mechanistic source terms for offsite dose-related assessments for advanced nuclear reactor facilities," 2020 (ML20192A250) 34 of 82 8

FQ Activity White paper assessment:

TerraPower, Advanced Fuel Qualification Methodology (ML20209A155)

White paper development:

General Atomics - Accelerated Fuel Qualification (AFQ)

NEA - Working Group on the Safety of Advanced Reactors (WGSAR)

Fuel Qualification Report (Draft) 35 of 82 9

Outline Background/Review of past engagement Activity affecting FQ guidance FQ report overview Incorporation of comments from May 7th stakeholder meeting Next steps and discussion 36 of 82 10

FQ Framework - Literature JNM 2007 Paper JNM 2020 Paper 37 of 82 11

FQ Framework - Scope Broad interpretation of fuel qualification (many aspects of nuclear safety are impacted by the fuel)

Neutronic performance Thermal-fluid performance (e.g., margin to critical heat flux limits)

Seismic behavior Fuel transportation and storage Need to restrict the scope of the report The scope of this report focuses on the identification and understanding of fuel life limiting and degradation mechanisms that occur as a result of irradiation during reactor operation.

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FQ Framework - Other Considerations Definition of fuel qualification (from JNM 2007)

The objective of nuclear fuel qualification is the demonstration that a fuel product fabricated in accordance with a specification behaves as assumed or described in the applicable licensing safety case, and with the reliability necessary for economic operation of the reactor plant Clarify safety case The role of nuclear fuel in the safety case can vary significantly between different reactor designs (e.g. TRISO fuel contains fission product barriers within the fuel itself) 39 of 82 13

FQ Framework Development of a generic assessment framework for fuel qualification:

Top-down approach used to decompose the top level goal of fuel is qualified into lower level supporting goals Lower level supporting goals are further decomposed until clear objective goals are identified that can be satisfied with direct evidence NRC has used assessment framework approach to evaluation thermal-margin models (Evaluations similar to NUREG/KM-0013)

Significant reduction in review time Comprehensive/transparent review 40 of 82 14

FQ Assessment Framework: Goal Goal: Fuel is qualified for use

= High confidence exists that the fuel fabricated in accordance its specification will perform as described in the applicable licensing safety case Goal: Fuel is qualified for use A fuel manufacturing specification controls the key Safety criteria can be fabrication parameters that satisfied with high significantly affect fuel confidence [G2]

performance [G1]

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G1: Manufacturing Specification A fuel manufacturing specification controls the key fabrication parameters that significantly affect fuel performance [G1]

Microstructure attributes Key dimensions and Key constituents are for materials within the tolerances of fuel specified with fuel component are components are allowance for specified or otherwise specified [G1.1] impurities [G1.2]

justified [G1.3]

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G2: Safety Criteria Safety criteria can be satisfied with high confidence [G2]

Margin to design limits can be Margin to radionuclide demonstrated under conditions release limits under accident Ability to achieve and of normal operation, including conditions can be maintain safe shutdown can the effects of anticipated demonstrated with high be assured [G2.3]

operational occurrences with confidence [G2.2]

high confidence [G2.1]

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G2.1: Design Limits for Normal and Anticipated Operational Occurrences Margin to design limits can be demonstrated under conditions of normal operation, including the effects of anticipated operational occurrences with high confidence [G2.1]

An evaluation model is available to assess fuel performance The The fuel fuel performance performance envelope envelope against design limits to protect isis defined defined [G2.1.1]

[G2.1.1] against fuel failure and degradation (i.e., life-limiting) mechanisms [G2.1.2]

Note: The fuel performance envelope specifies the environmental conditions and radiation exposure that the fuel is expected to encounter. The envelope is typically specified by fuel designers and provides constraints on the design of the reactor and associated systems.

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Evaluation Model (EM)

Assessment Framework Goal: The evaluation model is acceptable The evaluation model contains The evaluation model has the appropriate modelling been adequately assessed capabilities against experimental data

[EM G1] [EM G2]

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Experimental Data (ED)

Assessment Framework Goal: Experimental data used for assessment is appropriate Assessment data is Data has been collected Experimental data Test specimens are independent of data over a test envelope have been representative of used to develop/train that covers the fuel accurately measured prototypical fuel the evaluation model performance envelope

[ED G3] [ED G4]

[ED G1] [ED G2]

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Summary of FQ Assessment Framework Supported by two additional assessment frameworks Evaluation Models Experimental Data A total of 60 terminal goals 11 in the main FQ Assessment Framework 2 x (14 in the Evaluation Model Assessment Framework) 3 x (7 in the Experimental Data Assessment Framework) 47 of 82 21

Outline Background/Review of past engagement Activity affecting FQ guidance FQ report overview Incorporation of comments from May 7th stakeholder meeting Next steps and discussion 48 of 82 22

Incorporation of Stakeholder Feedback

  • Clarify how to make judgements against the criteria, Ed Lyman (UCS)

- Added language to support criteria and to provide clarifying language/examples to demonstrate how criteria can be satisfied

  • Need adequate data to account for uncertainties, Cyril Draffin (USNIC)

- Added language in support of the evaluation model goal EM G2, The evaluation model has been adequately assessed against experimental data, which includes considerations of error quantification (i.e., uncertainty)

  • Align safety criteria with fundamental safety functions, Clint Medlock (SNC)

- Added Section 2.2.3 to clarify interfaces between fuel qualification and RG 1.233 49 of 82 23

Incorporation of Stakeholder Feedback

  • Address lead test specimens, Darrell Gardner (Kairos)

- Added Section 2.4, Lead Test Specimens, and supporting language in Section 3.4.2, ED G2 Test Envelope, to address the use of lead test specimens

  • Clarify the use of the term prototype as to not confuse with the prototype provisions of 10 CFR 50.43(e), Darrell Gardner (Kairos)

- Updated FQ framework and report to replace prototypical fuel with proposed fuel design as needed

  • The assessment framework is similar to the objective hierarchy concept established in NUREG/BR-0303

- Added reference to the objective hierarchy in Section 2.5, Assessment Frameworks 50 of 82 24

Outline Background/Review of past engagement Activity affecting FQ guidance FQ report overview Incorporation of comments from May 7th stakeholder meeting Next steps and discussion 51 of 82 25

Next Steps

  • Legal reviews

- Congressional Review Act (CRA)

  • Convert report into a regulatory document (e.g. NUREG) 52 of 82 26

Contact Information

  • Tim Drzewiecki Timothy.Drzewiecki@nrc.gov 301-415-5184 53 of 82 27

Assessment of the MACCS Code Applicability for Nearfield Consequence Analysis PRESENTED BY Dan Clayton, Nate Bixler Sandia National Laboratories Presented at the Advanced Reactor Stakeholder Meeting , October 1, 2020 Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell 54 of 82 International Inc., for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525.

SAND2020-8592 C

2 Outline Introduction/Setup Code Trends Code Comparisons Wrap up 55 of 82

3 Introduction (1/2)

1. The adequacy of the MELCOR Accident Consequence Code System (MACCS) in the nearfield is discussed in a non-Light Water Reactor (LWR) vision and strategy report that discusses computer code readiness for non-LWR applications developed by the Nuclear Regulatory Commission (NRC)
2. MACCS currently includes a simple model for building wake effects. The MACCS2 Users Guide suggests that this simple building wake model should not be used at distances closer than 500 m. This statement raises the first question of whether MACCS can reliably be used to assess nearfield doses, i.e., at distances less than 500 m 56 of 82

4 Introduction (2/2)

3. MACCS is a highly flexible Gaussian model and the user can choose whether to model a variety of physical phenomena, including such things as building wake effects, plume buoyancy, and plume meander. Furthermore, the user has flexibility in choosing how to model the Gaussian dispersion parameters
4. So, a second question goes beyond the first question of whether MACCS can be used in the nearfield to the related question of how can MACCS be used to generate results that are bounding of other codes intended for nearfield analysis 57 of 82

5 General Arrangement of Flow Zones Near a Sharp-edged Building Meteorology and Atomic Energy, 1968 58 of 82

6 Objective An evaluation of modeling approaches (methods) to estimate nearfield air concentrations and depositions was performed where several candidate codes were ranked for comparison and potential incorporation into the MACCS code In this report, it is assumed that the results from the selected codes are all adequate in the nearfield, which is reasonable because these codes are specifically intended to be used in the nearfield Hence, by comparing the results of these codes to the results from MACCS, the adequacy of MACCS for assessing exposures in the nearfield can be evaluated, along with determining how MACCS can be used to generating bounding results 59 of 82

7 Nearfield Code List Four candidate codes were selected from the three main methods of atmospheric transport and dispersion (ATD) in the nearfield and evaluated

  • CFD models - OpenFOAM
  • Simplified wind-field models - QUIC
  • Modified Gaussian models - AERMOD and ARCON96 Based on these rankings, QUIC, AERMOD, and ARCON96 and were selected for comparison with MACCS 60 of 82

8 Test Cases Two weather conditions

  • 4 m/s, neutrally-stable (D stability class) - typical condition
  • 2 m/s, stable (F stability class) - reduced dispersion condition Three building configurations (HxWxL)
  • 20m x 100m x 20m (5:1 W:H) - extreme width to height ratio
  • 20m x 40m x 20m (2:1 W:H) - typical building size
  • No building (point source) - evaluate differences for elevated releases with no building Two power levels (heat content)
  • 0 MW - without buoyancy
  • 5 MW - with buoyancy 61 of 82

Code Trends 62 of 82

10 MACCS Results Building and elevation effects greatly diminished at 800 m downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions Buoyant plumes that escape building wake produce significantly lower dilution values due to fast plume rise compared with dispersion 63 of 82

11 ARCON96 Results Minimal change due to inclusion of building or elevated release within 1 km Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions No plume rise model implemented; buoyant cases were not modeled 64 of 82

12 AERMOD Results Building and elevation effects greatly diminished at 500 m downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions Minor differences due to buoyancy 65 of 82

13 QUIC Results (1/2)

Building and elevation effects greatly diminished at 1 km downwind Building significantly increases dispersion at short distances Dilution for stable conditions generally higher than the corresponding dilution for neutrally-stable conditions No straightforward way to implement buoyancy; buoyant cases were not modeled 66 of 82

14 QUIC Results (2/2)

Horizontal and vertical slices for a 4 m/s, neutrally-stable weather condition with a non-buoyant, elevated release from a 20 m x 100 m x 20 m building (Case 01) 67 of 82

Code Comparisons 68 of 82

16 Comparison Results At 50 m, order from highest to lowest dilution is ARCON96, AERMOD, QUIC, MACCS Order changes with distance

  • ARCON96 shifts from highest to lowest
  • AERMOD shifts from 2nd highest to 2nd lowest
  • Relative order between QUIC and MACCS is consistent 69 of 82

17 Potential Modifications to MACCS Input

1. Specify a ground-level release, instead of a release at the height of the building
  • ARCON96 model showed little dependence on elevation of release
  • Wake-induced building downwash observed in QUIC output
  • Regulatory Guide 1.145 discusses releases less than 2.5 times building height should be modeled as ground-level releases
2. Specify no buoyancy (plume trapped in building wake)
  • AERMOD model showed little dependence on buoyancy
3. If additional conservatism needed or desired, model as a point source
  • ARCON96 model showed little dependence on building size
  • DOE approach used for collocated workers
  • If point source too bounding, use an intermediate building wake size 70 of 82

18 Updated Comparison Results MACCS input modified to reflect a ground-level (1), non-buoyant (2) release (grey) bounds AERMOD and QUIC up to 1 km and ARCON96 from 200 m up to 1 km MACCS input modified to reflect a ground-level (1), non-buoyant (2),

point-source (3) release (light blue) bounds all three up to 1 km 71 of 82

Wrap up 72 of 82

20 Summary (1/3)

ARCON96, AERMOD, and QUIC selected for comparison with MACCS based on initial evaluation Test cases developed to give a broad range of conditions, not to be exhaustive

  • Two weather conditions
  • Three building configurations
  • Two buoyancy variations 73 of 82

21 Summary (2/3)

MACCS calculations configured with point-source, ground-level, nonbuoyant plumes provide conservative nearfield results that bound the centerline, ground-level air concentrations from ARCON96, AERMOD, and QUIC .

MACCS calculations with ground-level, nonbuoyant plumes that include the effects of the building wake (area source) provide nearfield results that bound the results from AERMOD and QUIC and the results from ARCON96 at distances >200 m If using a point-source is too conservative and it is desired to bound the results from all three codes, another alternative is to use area source parameters in MACCS that are less than the standard values, i.e., an area source intermediate between the standard recommendation and a point source.

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22 Summary (3/3)

MACCS can be used at distances significantly shorter than 500 m downwind (50 - 200 m) from a containment or reactor building However, the MACCS user needs to select the MACCS input parameters appropriately to generate results that are adequately conservative for a specific application A conservative nearfield result may be obtained using the following MACCS parameter choices:

  • The parameterization of Eimutis and Konicek for the dispersion model.
  • The plume meander model based on Regulatory Guide 1.145. This model is selected by setting the value of the MACCS parameter MNDMOD to NEW.
  • The release modeled as a point-source, ground-level, nonbuoyant plume.

Additional information available from final technical report (Clayton D.J and N.E. Bixler, Assessment of the MACCS Code Applicability for Nearfield Consequence Analysis Sandia National 75 of 82 Laboratories, Albuquerque, NM, February 2020, ADAMS Accession Number ML20059M032)

Overview of the Oak Ridge National Laboratory Report on Preparing and Reviewing a Non-Power Liquid Fueled Molten Salt Reactor License Application William B. Kennedy Project Manager Non-Power Production and Utilization Facility Licensing Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities U.S. Nuclear Regulatory Commission 76 of 82

Background

  • Under contract with NRC, Oak Ridge National Laboratory developed a report titled, Proposed Guidance for Preparing and Reviewing a Molten Salt Non-Power Reactor Application 77 of 82

Overview of the Report

  • An information resource for stakeholders interested in licensing of non-power MSRs
  • Based on NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors
  • Focuses on the technical information needed to apply NUREG-1537 to a non-power MSR license application 78 of 82

Overview of the Report

  • Covers topics including:

- Siting

- Design of structures, systems, and components

- Reactor description

- Reactor cooling systems

- Engineered safety features

- Instrumentation and control systems

- Auxiliary systems

- Radiation protection and waste management

- Accident analysis

- Technical specifications 79 of 82

Future Plans

  • The NRC staff is considering endorsing the report for use by potential non-power MSR applicants by January 2021
  • Subsequently, the NRC staff will consider incorporating appropriate information from the report in an existing NRC guidance document, such as the next revision of NUREG-1537, a process that would include a formal public comment period
  • Any feedback is welcome 80 of 82

How to Get the Report

  • Available on the NRCs Agencywide Documents Access and Management System (ADAMS) at Accession No. ML20219A771
  • Contact me at william.kennedy@nrc.gov 81 of 82

Future Meeting Planning and Open Discussion 2020 Tentative Schedule for Periodic Stakeholder Meetings October 22 (TICAP, ARCAP)

November 5 82 of 82

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