ML22046A312
| ML22046A312 | |
| Person / Time | |
|---|---|
| Issue date: | 02/17/2022 |
| From: | NRC/NRR/DANU |
| To: | |
| Costa A | |
| References | |
| Download: ML22046A312 (140) | |
Text
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
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 ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 2 Reactor Initiatives, 02/17/2022
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 2 Slide 4 Reactor Initiatives, 02/17/2022
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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BREAK ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 45 Reactor Initiatives, 02/17/2022
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https://www.nrc.gov/reactors/new-reactors/advanced/details.html#non-lwr-ana-code-dev ACRS meeting on Integration of Source Term Activities in Support of Advanced 46 Slide 62 Reactor Initiatives, 02/17/2022
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NuScale EPZ Sizing Methodology Topical Report, Rev. 2 Light Water SMR Design Certification Source Term Approach Source Term Approach for Early non-LWR Movers ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 79 Reactor Initiatives, 02/17/2022
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 2 Slide 81 Reactor Initiatives, 02/17/2022
SMR and Non-LWR Accident Source Terms Recent Experience
- 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 3 Slide 82 Reactor Initiatives, 02/17/2022
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 4 Slide 83 Reactor Initiatives, 02/17/2022
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 5 Slide 84 Reactor Initiatives, 02/17/2022
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)
- 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 6 Slide 85 Reactor Initiatives, 02/17/2022
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 7 Slide 86 Reactor Initiatives, 02/17/2022
NuScale EPZ Sizing Methodology Topical Report
- Source term refers to fission product release to the environment as a function of time
- 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 8 Slide 87 Reactor Initiatives, 02/17/2022
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 9 Slide 88 Reactor Initiatives, 02/17/2022
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 10 Slide 89 Reactor Initiatives, 02/17/2022
NuScale TR: Core Damage Event
- Derive source term from range of accident scenarios that result in significant damage to the core
- 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 11 Slide 90 Reactor Initiatives, 02/17/2022
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 12 Slide 91 Reactor Initiatives, 02/17/2022
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 13 Slide 92 Reactor Initiatives, 02/17/2022
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
- Hermes CP application (under review)
- Evaluates MHA, deterministic
- Refers to MST TR ACRS meeting on Integration of Source Term Activities in Support of Advanced 14 Slide 93 Reactor Initiatives, 02/17/2022
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 15 Slide 94 Reactor Initiatives, 02/17/2022
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 16 Slide 95 Reactor Initiatives, 02/17/2022
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 17 Slide 96 Reactor Initiatives, 02/17/2022
LUNCH ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 97 Reactor Initiatives, 02/17/2022
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 19 Slide 99 Reactor Initiatives, 02/17/2022
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 20 Slide 100 Reactor Initiatives, 02/17/2022
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 21 Slide 101 Reactor Initiatives, 02/17/2022
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 22 Slide 102 Reactor Initiatives, 02/17/2022
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 23 Slide 103 Reactor Initiatives, 02/17/2022
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 24 Slide 104 Reactor Initiatives, 02/17/2022
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 26 Slide 106 Reactor Initiatives, 02/17/2022
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 27 Slide 107 Reactor Initiatives, 02/17/2022
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 28 Slide 108 Reactor Initiatives, 02/17/2022
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
- 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 29 Slide 109 Reactor Initiatives, 02/17/2022
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 30 Slide 110 Reactor Initiatives, 02/17/2022
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 31 Slide 111 Reactor Initiatives, 02/17/2022
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 32 Slide 112 Reactor Initiatives, 02/17/2022
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 33 Slide 113 Reactor Initiatives, 02/17/2022
MST Formulation
=
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)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 34 Slide 114 Reactor Initiatives, 02/17/2022
Technology-Inclusive Source Term Methodology Determination ACRS meeting on Integration of Source Term Activities in Support of Advanced 35 Slide 115 Reactor Initiatives, 02/17/2022
INL Report Methodology Steps 1: Identify Regulatory 8. Establish Adequacy of MST Requirements Simulation Tools 2: Identify Reference Facility 9. Develop and Update PRA Design Model 3: Define Initial Radionuclide 10. Identify or Revise the List of Inventories LBEs
- 4. Perform Bounding Calculations 11. Select LBEs to Include Design
- 5. Conduct SHA and Perform Basis External Hazard Level for Simplified Calculations Source Term Analysis
- 6. Consider Risk-informed System 12. Perform Source Term Design Changes Modeling and Simulation for LBEs
- 7. Select Initial List of LBEs and 13. Review LBEs List for Adequacy Conduct PIRT of Regulatory Acceptance
- 14. Document Completion of Source Term Development ACRS meeting on Integration of Source Term Activities in Support of Advanced 36 Slide 116 Reactor Initiatives, 02/17/2022
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 37 Slide 117 Reactor Initiatives, 02/17/2022
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 38 Slide 118 Reactor Initiatives, 02/17/2022
Licensing Modernization
- Risk-informed approach to selection and analysis of licensing basis events
- Combined with assessment of cumulative risks
- Key roles for PRA and MST ACRS meeting on Integration of Source Term Activities in Support of Advanced 39 Slide 119 Reactor Initiatives, 02/17/2022
Licensing Modernization 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 ACRS meeting on Integration of Source Term Activities in Support of Advanced 40 Slide 120 Reactor Initiatives, 02/17/2022
Licensing Modernization
- Flexibility provided on how to develop safety case
- NRC Advanced Reactor Policy Statement encourages use of passive and inherent features ACRS meeting on Integration of Source Term Activities in Support of Advanced 41 Slide 121 Reactor Initiatives, 02/17/2022
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 42 Slide 122 Reactor Initiatives, 02/17/2022
FQ Assessment Framework Goal: Fuel is qualified for use A fuel manufacturing specification controls the key fabrication parameters that significantly affect fuel Safety criteria can be satisfied [G2]
performance [G1]
ACRS meeting on Integration of Source Term Activities in Support of Advanced 43 Slide 123 Reactor Initiatives, 02/17/2022
G2: Safety Criteria Safety criteria can be satisfied [G2]
Margin to design limits can be Margin to radionuclide demonstrated under conditions Ability to achieve and release limits under accident of normal operation, including maintain safe shutdown can conditions can be the effects of anticipated be assured [G2.3]
demonstrated [G2.2]
operational occurrences [G2.1]
10 CFR 50.34(a)(1)(ii)(D) GDC/ARDC 2 GDC/ARDC 10 10 CFR 52.47(a)(2)(iv) GDC 27/ARDC 26 10 CFR 52.79(a)(1)(vi) GDC/ARDC 35 ACRS meeting on Integration of Source Term Activities in Support of Advanced 44 Slide 124 Reactor Initiatives, 02/17/2022
G2.2: Radionuclide Release Limits Margin to radionuclide release limits under accident conditions can be demonstrated [G2.2]
Radionuclide retention The fuel performance requirements of the fuel Radionuclide retention envelope is defined under accident and release behavior of Criteria for barrier degradation the fuel matrix under
[G2.1.1] conditions is and failure under accident specified [G2.2.1] accident conditions is conditions are suitably modeled conservatively conservative [G2.2.2] [G2.2.3]
ACRS meeting on Integration of Source Term Activities in Support of Advanced 45 Slide 125 Reactor Initiatives, 02/17/2022
G2.2.2 Criteria for Barrier Degradation Criteria for barrier degradation and failure under accident conditions are suitably conservative [G2.2.2]
Criteria are shown to provide conservative Experimental data is prediction of barrier appropriate degradation and failure [G2.2.2(b)]
[G2.2.2(a)]
Note: Testing at environmental conditions consistent with accident conditions is expected (e.g., elevated temperatures)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 46 Slide 126 Reactor Initiatives, 02/17/2022
Complete FQ Assessment Framework GOAL Fuel is qualified for use GOAL Evaluation model is acceptable for use G1 Fuel is manufactured in accordance with a specification EM G1 Evaluation model contains the appropriate modeling capabilities G1.1 Key dimensions and tolerances of fuel components are specified EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system G1.2 Key constituents are specified with allowance for impurities EM G1.2 Evaluation model is capable of modeling the material properties of the fuel G1.3 End state attributes for materials within fuel components are specified or system otherwise justified EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel G2 Margin to safety limits can be demonstrated performance G2.1 Margin to design limits can be demonstrated under conditions of normal EM G2 Evaluation model has been adequately assessed against experimental data operation and AOOs EM G2.1 Data used for assessment are appropriate (see ED Assessment G2.1.1 Fuel performance envelope is defined Framework)
G2.1.2 Evaluation model is available (see EM Assessment Framework) EM G2.2 Evaluation model is demonstrably able to predict fuel failure and G2.2 Margin to radionuclide release limits under accident conditions can be degradation mechanisms over the test envelope demonstrated EM G2.2.1 Evaluation model error is quantified through assessment G2.1.1 Fuel performance envelope is defined against experimental data G2.2.1 Radionuclide retention requirements are specified EM G2.2.2 Evaluation model error is determined throughout the fuel G2.2.2 Criteria for barrier degradation and failure are suitably conservative performance envelope EM G2.2.3 Sparse data regions are justified (a) Criteria are conservative EM G2.2.4 Evaluation model is restricted to use within its test envelope (b) Experimental data are appropriate (see ED Assessment Framework)
G2.2.3 Radionuclide retention and release from fuel matrix are modeled GOAL Experimental data used for assessment are appropriate conservatively ED G1 Assessment data are independent of data used to develop/train the evaluation model (a) Model is conservative ED G2 Data has been collected over a test envelope that covers the fuel performance (b) Experimental data are appropriate (see ED Assessment envelope Framework) ED G3 Experimental data have been accurately measured G2.3 Ability to achieve and maintain safe shutdown is assured ED G3.1 The test facility has an appropriate quality assurance program G2.3.1 Coolable geometry is ensured ED G3.2 Experimental data are collected using established measurement techniques (a) Criteria to ensure coolable geometry are specified ED G3.3 Experimental data account for sources of experimental uncertainty (b) Evaluation models are available (see EM Assessment ED G4 Test specimens are representative of the fuel design Framework) ED G4.1 Test specimens are fabricated consistent with the fuel manufacturing G2.3.2 Negative reactivity insertion can be demonstrated specification (a) Criteria are provided to ensure that negative reactivity ED G4.2 Distortions are justified and accounted for in the experimental data insertion path is not obstructed (b) Evaluation model is available (see EM Assessment Framework)
- For illustrative purposes only. Please see Appendix A to NUREG-2246 for a legible list.
ACRS meeting on Integration of Source Term Activities in Support of Advanced 47 Slide 127 Reactor Initiatives, 02/17/2022
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 48 Slide 128 Reactor Initiatives, 02/17/2022
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 49 Slide 129 Reactor Initiatives, 02/17/2022
Guidance for developing advanced reactor source term (long-term)
Bill Reckley Michelle Hart John Segala NRR/DANU Slide 130
General Approach
- 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)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 Slide 131 Reactor Initiatives, 02/17/2022
DOE/National Laboratories ACRS meeting on Integration of Source Term Activities in Support of Advanced 3 Slide 132 Reactor Initiatives, 02/17/2022
NRC Activities ACRS meeting on Integration of Source Term Activities in Support of Advanced 4 Slide 133 Reactor Initiatives, 02/17/2022
Next Generation Nuclear Plant (NGNP)
ACRS meeting on Integration of Source Term Activities in Support of Advanced 5 Slide 134 Reactor Initiatives, 02/17/2022
Model Development ACRS meeting on Integration of Source Term Activities in Support of Advanced 6 Slide 135 Reactor Initiatives, 02/17/2022
Applications & Pre-App Interactions ACRS meeting on Integration of Source Term Activities in Support of Advanced 7 Slide 136 Reactor Initiatives, 02/17/2022
Moving Forward
- 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 ACRS meeting on Integration of Source Term Activities in Support of Advanced 8 Slide 137 Reactor Initiatives, 02/17/2022
Opportunity for Public Comment ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 138 Reactor Initiatives, 02/17/2022
Member Discussion ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 139 Reactor Initiatives, 02/17/2022
Adjourn ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 140 Reactor Initiatives, 02/17/2022