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{{#Wiki_filter: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
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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
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 Slide 6    Reactor Initiatives, 02/17/2022
 
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ACRS meeting on Integration of Source Term Activities in Support of Advanced 11 Slide 15    Reactor Initiatives, 02/17/2022
 
NRC Analytical Tools and Past Studies-Severe Accident Progression and Source Term Hossein Esmaili, RES/DSA Jason Schaperow, NRR/DANU Slide 16
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 2 Slide 17    Reactor Initiatives, 02/17/2022
 
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ACRS meeting on Integration of Source Term Activities in Support of Advanced 29 Slide 44    Reactor Initiatives, 02/17/2022
 
BREAK ACRS meeting on Integration of Source Term Activities in Support of Advanced Slide 45    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 30 Slide 46    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 31 Slide 47    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 32 Slide 48    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 33 Slide 49    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 34 Slide 50    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 35 Slide 51    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 36 Slide 52    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 37 Slide 53    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 38 Slide 54    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 39 Slide 55    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 40 Slide 56    Reactor Initiatives, 02/17/2022
 
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ACRS meeting on Integration of Source Term Activities in Support of Advanced 42 Slide 58    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 43 Slide 59    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 44 Slide 60    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 45 Slide 61    Reactor Initiatives, 02/17/2022
 
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
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 47 Slide 63    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 48 Slide 64    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 49 Slide 65    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 50 Slide 66    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 51 Slide 67    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 52 Slide 68    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 53 Slide 69    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 54 Slide 70    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 55 Slide 71    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 56 Slide 72    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 57 Slide 73    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 58 Slide 74    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 59 Slide 75    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 60 Slide 76    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 61 Slide 77    Reactor Initiatives, 02/17/2022
 
ACRS meeting on Integration of Source Term Activities in Support of Advanced 62 Slide 78    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 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
* 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 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)
* 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 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
* 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 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
  - 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 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
            - 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 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 &sect; 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
* 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 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
* 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)
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}}

Revision as of 18:03, 18 November 2024

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
Person / Time
Issue date: 02/17/2022
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NRC/NRR/DANU
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Costa A
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