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Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, February 17, 2022
	ML22046A312
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Issue date:	02/17/2022
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Advisory Committee on Reactor Safeguards Future Plant Designs Subcommittee Meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives February 17, 2022 Slide 1

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 AGENDA Opening Remarks Staff Introduction History and Evolution of LWR Source Term NRC analytical tools and past studies SCALE/MELCOR non-LWR reference plant analysis Break Agenda Item IV Continued NuScale EPZ Sizing Methodology Topical Report, Rev. 2 Light water SMR design certification source term approach Source term approach for early non-LWR movers Lunch Accident-consequence-related regulation activities Break Guidance and information for developing advanced reactor source term Guidance for developing advanced reactor source term (long-term)

Opportunity for Public Comment Member Discussion Adjourn Slide 2

Integration of Source Term Activities in Support of Advance Reactor Initiatives John Segala NRR/DANU February 17, 2022 Slide 3

Staff Introduction

Determining source terms is a critical component in the NRCs licensing process

NRC team presenting today:

- Mark Blumberg - NRR/DRA

- Michelle Hart - NRR/DANU

- Jason Schaperow - NRR/DANU

- Bill Reckley - NRR/DANU

- Tim Drzewiecki - NRR/DANU

- Hossein Esmaili - RES/DSA ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 2

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Integration of Source Term Activities in Support of Advance Reactor Initiatives John Segala NRR/DANU February 17, 2022 Slide 5

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NRC Analytical Tools and Past Studies-Severe Accident Progression and Source Term Hossein Esmaili, RES/DSA Jason Schaperow, NRR/DANU Slide 16

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ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 BREAK Slide 45

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ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 46 https://www.nrc.gov/reactors/new-reactors/advanced/details.html#non-lwr-ana-code-dev Slide 62

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ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 NuScale EPZ Sizing Methodology Topical Report, Rev. 2 Light Water SMR Design Certification Source Term Approach Source Term Approach for Early non-LWR Movers Slide 79

Accident Source Term in Recent and Near-term Applications Michelle Hart NRR/DANU/UTB2 Slide 80

Outline

SMR and non-LWR accident source terms recent experience

Emergency planning zone size justification consequence analyses

Example: SMR design certification source term approach

Source term approaches for non-LWR early movers ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 2

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SMR and Non-LWR Accident Source Terms Recent Experience

SMR topical report reviews and SMR DC application review

Advanced reactor pre-application interactions, topical report reviews, and license applications

Source term development contractor reports ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 3

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Emergency Planning Zone Size Justification Consequence Analyses

Concept based on NUREG-0396

- Technical basis for plume exposure and ingestion pathway EPZ radius of ~10 and ~50 miles, respectively

- Identification of area within which prompt protective actions may be necessary to provide dose savings in the event of a radiological release

Calculate dose at distance for a spectrum of accidents

- Analysis includes design basis accidents and severe accidents ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 4

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Emergency Planning Zone Size Justification Consequence Analyses

No separate/unique source terms developed especially for EPZ size analysis

- Re-use source terms and accident release information developed for safety analysis report and PRA ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 5

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Emergency Planning Zone Size Justification Consequence Analyses

Methodology to support exemptions to 10-mile requirement

- Clinch River ESP EPZ size methodology described in SSAR

Methodology to support plume exposure pathway EPZ size determination on case-by-case basis for reactors <250 MWt

- NuScale EPZ sizing methodology topical report (under review)

EPZ size determination required in EP for SMRs and ONTs alternative framework, once issued

- SECY-22-0001 issued for Commission review and approval

- Guidance on analysis in appendices to RG 1.242 ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 6

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NuScale EPZ Sizing Methodology Topical Report

TR-0915-17772, Revision 2, submitted in 2020, currently under review

- Not part of DC review

- Applicable to light-water SMRs such as NuScale, although not limited to the NuScale designs

- Rev. 3 under development

Analysis methodology to determine plume exposure pathway EPZ size ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 7

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NuScale EPZ Sizing Methodology Topical Report

Source term refers to fission product release to the environment as a function of time

Uses source terms from DBAs (DC FSAR Ch. 15) and PRA severe accident scenarios scoped into analysis

- No separate/unique source terms developed especially for EPZ size analysis

- Uses CDF from PRA to categorize severe accidents and select accident sequences to evaluate against relevant dose criteria ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 8

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Example: SMR Design Certification Source Term Approach

SECY-19-0079, August 16, 2019

-Describes staff review approach to evaluate accident source terms for both the TR and the NuScale SMR DC application

-Provides basis for using source term without core damage for environmental qualification ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 9

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Example: SMR Design Certification Source Term Approach - NuScale TR

NuScale TR-0915-17565, Accident Source Term Methodology, Revision 4, February 2020

- Methods to develop accident source terms are consistent with RG 1.183 guidance for PWRs except for:

Core damage source term for Core Damage Event

Iodine spike design basis source term (no fuel damage)

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 10 Slide 89

NuScale TR: Core Damage Event

Derive source term from range of accident scenarios that result in significant damage to the core

- Informed by NuScale SMR PRA

NuScale-design-specific analyses using MELCOR to be performed by applicant referencing the TR

Radionuclide transport phenomena

- Iodine retention in containment based on pH

- Aerosol natural deposition in containment ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 11 Slide 90

NuScale SMR DC Application: Core Damage Event

Implemented the NuScale TR methodology to determine the core damage source term

Core inventory calculated using SCALE code

Scenario selection

- Based on NuScale SMR PRA, internal events

- 5 surrogate scenarios

Various failures of ECCS, with decay heat removal system available

Intact containment ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 12 Slide 91

NuScale SMR DC Application: Core Damage Event

MELCOR used to estimate release timing and magnitude for each scenario

- Release onset and duration from scenario with minimum time to core damage

- Core release fractions taken as median of scenarios

Time-dependent aerosol removal rates calculated using STARNAUA code

- Design-specific input thermal hydraulic conditions calculated by MELCOR for surrogate scenario with minimum time to core damage ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 13 Slide 92

Source Term Approaches for Non-LWR Early Movers

Kairos Power

- MST methodology TR (under review)

Methodology for applicants to develop event-specific radiological source terms

- DBAs for siting and safety analysis

- AOOs and DBEs for LMP

- Hermes CP application (under review)

Evaluates MHA, deterministic

Refers to MST TR ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 14 Slide 93

Source Term Approaches for Non-LWR Early Movers

X-energy

- Proposed to use developer-made source term code (XSTERM) which includes modeling of radionuclides from generation to release (and dose)

- TR was submitted, but withdrawn to clarify and resubmit in future (not currently under review)

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 15 Slide 94

Source Term Approaches for Non-LWR Early Movers

Oklo Aurora COL application (review ended)

- Proposed maximum credible accident without release

TerraPower

- Development of source term methodology described in 1/13/2022 public meeting (ML22011A072)

- Topical report planned for April 2023

Terrestrial, Westinghouse, Others

- Source terms to be determined

- Public website information on non-LWR pre-application activities ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 16 Slide 95

Acronyms AOO anticipated operational occurrence CDF core damage frequency COL combined license CP construction permit DBA design basis accident DBE design basis event DC design certification ECCS emergency core cooling system EP emergency preparedness EPZ emergency planning zone ESP early site permit FSAR final safety analysis report LMP Licensing Modernization Project MHA maximum hypothetical accident MST mechanistic source term MWt megawatts thermal Non-LWR non-light water reactor ONTs other new technologies PRA probabilistic risk assessment PWR pressurized water reactor RG regulatory guide SMR small modular reactor SSAR site safety analysis report TR topical report ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 17 Slide 96

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 LUNCH Slide 97

Accident Consequence-Related Regulation Activities Michelle Hart NRR/DANU/UTB2 Slide 98

Petition for Rulemaking

PRM-50-121, Voluntary Adoption of Revised Design Basis Accident Dose Criteria

- Received 11/23/2019, docketed 2/19/2020 (85 FR 31709)

- Under evaluation - no disposition yet

Requests voluntary rule to allow power reactor licensees to adopt alternative to the accident dose criteria specified in § 50.67, Accident source term.

Proposes a uniform value of 100 milli-Sieverts (10 rem) for offsite locations and for the control room ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 19 Slide 99

Emergency Preparedness for SMRs and Other New Technologies Rulemaking

Final rule in development

- New section 10 CFR 50.160, and related/conforming changes

- ACRS meetings in September and November 2021

RG 1.242 (to be issued with final rule)

- Appendices

Generalized analysis methodology

Information on source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 20 Slide 100

Emergency Preparedness for SMRs and Other New Technologies Rulemaking

Appendix A, General Methodology for Establishing Plume Exposure Pathway Emergency Planning Zone Size

- Provides general guidance on the consequence analysis to support plume exposure pathway EPZ size determination

- Discusses event selection and consideration of accident likelihood ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 21 Slide 101

Emergency Preparedness for SMRs and Other New Technologies Rulemaking

Appendix B, Development of Information on Source Terms

- Provides guidance to develop source terms for plume exposure pathway EPZ size evaluations ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 22 Slide 102

Alternative Physical Security for Advanced Reactors Rulemaking

Draft rule and guidance in development

Voluntary alternative physical security requirements commensurate with potential safety and security consequences

Analyses (guidance under development)

- Develop relevant scenarios

- Site-specific potential offsite radiological consequences ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 23 Slide 103

Acronyms CFR Code of Federal Regulations EPZ emergency planning zone FR Federal Register PRM petition for rulemaking RG Regulatory Guide SMR small modular reactor ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 24 Slide 104

Guidance and Information for Developing Source Terms for Non-LWRs Michelle Hart, NRR/DANU/UTB2 Bill Reckley, NRR/DANU/UARP Tim Drzewiecki, NRR/DANU/UTB1 Slide 105

Outline

Accident consequence analysis for advanced reactors

Mechanistic source term

Recent reports on Non-LWR source term development

Non-LWR PRA standard and source term

Licensing Modernization Project and source term

Overview of method in NUREG-2246, Fuel Qualification for Advanced Reactors

Non-LWR accident source term information website ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 26 Slide 106

Accident Consequence Analysis for Advanced Reactors

Regulatory nexus

- Siting and safety analysis regulatory requirement

- Newer uses for advanced reactors

LMP

Plume exposure pathway EPZ size determination

Alternative security requirements - ongoing rulemaking

Part 53 - ongoing rulemaking ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 27 Slide 107

Accident Consequence Analysis for Advanced Reactors

Accident source term development considerations

- Event selection, scenarios

- Balance of prevention vs. mitigation

- Relationship to functional containment

A barrier, or set of barriers taken together, that effectively limit the physical transport of radioactive material to the environment (SECY-18-0096)

- Relationship to PRA

- Uncertainty ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 28 Slide 108

Accident Consequence Analysis for Advanced Reactors

Mechanistic or deterministic evaluation

- LMP assumes MST and use of PRA

- Some non-LWRs may choose to provide a postulated MHA, similar to non-power reactor licensees

No current specific RG on MST or non-LWR source terms, however

- RG 1.183, regulatory position C.2, Attributes of an Acceptable AST, may be useful

- SECY-93-092 included staff recommendations on non-LWR source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 29 Slide 109

Mechanistic Source Term

SECY-93-092 definition of MST A mechanistic source term is the result of an analysis of fission product release based on the amount of cladding damage, fuel damage, and core damage resulting from the specific accident sequences being evaluated. It is developed using best-estimate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs.

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 30 Slide 110

SECY-93-092: Provisions for Staff Assurance The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis.

Sufficient data should exist on the reactor and fuel performance through the research, development, and testing programs to provide adequate confidence in the mechanistic approach.

The transport of fission products can be adequately modeled for all barriers and pathways to the environs, including specific consideration of containment design. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured.

The events considered in the analyses to develop the set of source terms for each design are selected to bound severe accidents and design-dependent uncertainties ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 31 Slide 111

National Lab Non-LWR Source Term Reports

Technology inclusive, what to do to develop accident source terms, not specific on how to do it

No specific methods or phenomenological models

Do not provide technology-related source terms or releases ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 32 Slide 112

Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities INL/EXT-20-58717, Revision 0, June 2020, ML20192A250

Summarizes a risk-informed, performance-based, and technology-inclusive approach to determine source terms

Graded process

- Conservative non-mechanistic approach

- MST calculation methods

Design-specific scenarios for a range of licensing basis events

Best-estimate models with uncertainty quantification ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 33 Slide 113

MST Formulation ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 34

=

Figure 1-2 INL/EXT-20-58717, Revision 1. From Illustration of radionuclides retention and removal process for one non-LWR concept (reproduced from SAND2020-0402)

Slide 114

Technology-Inclusive Source Term Methodology Determination ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 35 Slide 115

INL Report Methodology Steps 1: Identify Regulatory Requirements 2: Identify Reference Facility Design 3: Define Initial Radionuclide Inventories

4. Perform Bounding Calculations

5. Conduct SHA and Perform Simplified Calculations

6. Consider Risk-informed System Design Changes

7. Select Initial List of LBEs and Conduct PIRT

8. Establish Adequacy of MST Simulation Tools

9. Develop and Update PRA Model

10. Identify or Revise the List of LBEs

11. Select LBEs to Include Design Basis External Hazard Level for Source Term Analysis

12. Perform Source Term Modeling and Simulation for LBEs

13. Review LBEs List for Adequacy of Regulatory Acceptance

14. Document Completion of Source Term Development ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 36 Slide 116

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

Primarily qualitative means to identify the dominant considerations that affect a release mitigation strategy

Classifies release mitigation strategies based on a range of barriers, physical attenuation processes, and system performance under sample accident scenarios

Did NOT develop quantitative estimates of radiological release magnitudes and compositions to the environment

Looked at high temperature gas reactors, sodium fast reactors, and liquid fueled molten salt reactors ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 37 Slide 117

Non-LWR PRA Standard ASME/ANS RA-S-1.4-2021

Full scope PRA (includes consequence analysis)

Mechanistic Source Term Analysis (MS) element provides useful information on what to do to develop mechanistic source terms ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 38 Slide 118

Licensing Modernization ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 39

Risk-informed approach to selection and analysis of licensing basis events

Combined with assessment of cumulative risks

Key roles for PRA and MST Slide 119

Licensing Modernization ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 40 See: SECY-18-0096, Functional Containment Performance Criteria for Non-Light-Water-Reactors, and INL/EXT-20-58717, Technology-Inclusive Determination of Mechanistic Source Terms for Offsite Dose-Related Assessments for Advanced Nuclear Reactor Facilities Slide 120

Licensing Modernization ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 41

Flexibility provided on how to develop safety case

NRC Advanced Reactor Policy Statement encourages use of passive and inherent features Slide 121

Assessment Frameworks Fuel Qualification (FQ)

Top-down approach to identify criteria (goals) to support a finding that fuel is qualified ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 42 Slide 122

FQ Assessment Framework ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 43 Goal: Fuel is qualified for use Safety criteria can be satisfied [G2]

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

Slide 123

G2: Safety Criteria ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 44 Safety criteria can be satisfied [G2]

Margin to design limits can be demonstrated under conditions of normal operation, including the effects of anticipated operational occurrences [G2.1]

Margin to radionuclide release limits under accident conditions can be demonstrated [G2.2]

Ability to achieve and maintain safe shutdown can be assured [G2.3]

GDC/ARDC 10 10 CFR 50.34(a)(1)(ii)(D) 10 CFR 52.47(a)(2)(iv) 10 CFR 52.79(a)(1)(vi)

GDC/ARDC 2 GDC 27/ARDC 26 GDC/ARDC 35 Slide 124

G2.2: Radionuclide Release Limits ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 45 Margin to radionuclide release limits under accident conditions can be demonstrated [G2.2]

Radionuclide retention and release behavior of the fuel matrix under accident conditions is modeled conservatively

[G2.2.3]

Criteria for barrier degradation and failure under accident conditions are suitably conservative [G2.2.2]

Radionuclide retention requirements of the fuel under accident conditions is specified [G2.2.1]

The fuel performance envelope is defined

[G2.1.1]

Slide 125

G2.2.2 Criteria for Barrier Degradation ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 46 Note: Testing at environmental conditions consistent with accident conditions is expected (e.g., elevated temperatures)

Criteria for barrier degradation and failure under accident conditions are suitably conservative [G2.2.2]

Experimental data is appropriate

[G2.2.2(b)]

Criteria are shown to provide conservative prediction of barrier degradation and failure

[G2.2.2(a)]

Slide 126

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 47 Complete FQ Assessment Framework GOAL Fuel is qualified for use G1 Fuel is manufactured in accordance with a specification G1.1 Key dimensions and tolerances of fuel components are specified G1.2 Key constituents are specified with allowance for impurities G1.3 End state attributes for materials within fuel components are specified or otherwise justified G2 Margin to safety limits can be demonstrated G2.1 Margin to design limits can be demonstrated under conditions of normal operation and AOOs G2.1.1 Fuel performance envelope is defined G2.1.2 Evaluation model is available (see EM Assessment Framework)

G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated G2.1.1 Fuel performance envelope is defined G2.2.1 Radionuclide retention requirements are specified G2.2.2 Criteria for barrier degradation and failure are suitably conservative (a)

Criteria are conservative (b)

Experimental data are appropriate (see ED Assessment Framework)

G2.2.3 Radionuclide retention and release from fuel matrix are modeled conservatively (a)

Model is conservative (b)

Experimental data are appropriate (see ED Assessment Framework)

G2.3 Ability to achieve and maintain safe shutdown is assured G2.3.1 Coolable geometry is ensured (a)

Criteria to ensure coolable geometry are specified (b)

Evaluation models are available (see EM Assessment Framework)

G2.3.2 Negative reactivity insertion can be demonstrated (a)

Criteria are provided to ensure that negative reactivity insertion path is not obstructed (b)

Evaluation model is available (see EM Assessment Framework)

GOAL Evaluation model is acceptable for use EM G1 Evaluation model contains the appropriate modeling capabilities EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system EM G1.2 Evaluation model is capable of modeling the material properties of the fuel system EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel performance EM G2 Evaluation model has been adequately assessed against experimental data EM G2.1 Data used for assessment are appropriate (see ED Assessment Framework)

EM G2.2 Evaluation model is demonstrably able to predict fuel failure and degradation mechanisms over the test envelope EM G2.2.1 Evaluation model error is quantified through assessment against experimental data EM G2.2.2 Evaluation model error is determined throughout the fuel performance envelope EM G2.2.3 Sparse data regions are justified EM G2.2.4 Evaluation model is restricted to use within its test envelope GOAL Experimental data used for assessment are appropriate ED G1 Assessment data are independent of data used to develop/train the evaluation model ED G2 Data has been collected over a test envelope that covers the fuel performance envelope ED G3 Experimental data have been accurately measured ED G3.1 The test facility has an appropriate quality assurance program ED G3.2 Experimental data are collected using established measurement techniques ED G3.3 Experimental data account for sources of experimental uncertainty ED G4 Test specimens are representative of the fuel design ED G4.1 Test specimens are fabricated consistent with the fuel manufacturing specification ED G4.2 Distortions are justified and accounted for in the experimental data

For illustrative purposes only. Please see Appendix A to NUREG-2246 for a legible list.

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Non-LWR Accident Source Term Webpage Information https://www.nrc.gov/reactors/new-reactors/advanced/related-documents/nuclear-power-reactor-source-term.html

One-stop shop for existing information, on public website under advanced reactors

- Discussion of accident source terms

- Linked list of documents relevant to development of non-LWR accident source terms for licensing

Staff will keep up to date ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 48 Slide 128

Acronyms AST alternative source term EPZ emergency planning zone INL Idaho National Laboratory LBE licensing basis event LMP Licensing Modernization Project LWR light water reactor MHA maximum hypothetical accident MST mechanistic source term Non-LWR non-light water reactor PIRT phenomena identification and ranking table PRA probabilistic risk assessment RG regulatory guide SHA system hazard analysis ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 49 Slide 129

Guidance for developing advanced reactor source term (long-term)

Bill Reckley Michelle Hart John Segala NRR/DANU Slide 130

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 2

Maintain traditional LWR approach (RG 1.183) as an acceptable option

Technology-inclusive methodology available as an option

Actual implementation is technology/design specific

NRC not planning to provide analytical inputs to applicants (beyond making available NRC developed models)

General Approach Slide 131

DOE/National Laboratories ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 3

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NRC Activities ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 4

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Next Generation Nuclear Plant (NGNP)

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 5

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Model Development ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 6

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Applications & Pre-App Interactions ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 7

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Moving Forward ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 8

Following the scientific work being done by national laboratories and developers

Engaging with developers

Continuing to develop NRC models and identify related uncertainties

Consider additional guidance based on experience with ongoing interactions

Consider feedback on the new webpage Slide 137

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 Opportunity for Public Comment Slide 138

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 Member Discussion Slide 139

ACRS meeting on Integration of Source Term Activities in Support of Advanced Reactor Initiatives, 02/17/2022 Adjourn Slide 140

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