ML25280A089
| ML25280A089 | |
| Person / Time | |
|---|---|
| Site: | Kemmerer File:TerraPower icon.png |
| Issue date: | 10/08/2025 |
| From: | Hart M, Hanh Phan, Radel T NRC/NRR/DANU/UTB2 |
| To: | |
| References | |
| Download: ML25280A089 (1) | |
Text
NRC Review of Kemmerer Unit 1 Construction Permit Application Safety Analysis Methods Hanh Phan, Michelle Hart, Tracy Radel NRR/DANU/UTB2 ACRS Subcommittee Meeting October 8-9, 2025
Probabilistic Risk Assessment Hanh Phan, Senior PRA Analyst, NRR/DANU/UTB2 2
PRA Topics
- 1. Regulations and Guidance
- 2. Kemmerer Unit 1 (KU1) PRA Overview
- 3. Staff Review Approach
- 4. PRA Review and Confirmation
- 6. Conclusion 3
- 1. Regulations and Guidance Applicable regulatory requirements:
- 10 CFR 50.34(a)(1) and (a)(4) - Contents of applications; technical information
- 10 CFR 50.35 - Issuance of construction permits Relevant guidance documents for PRA evaluation:
- RG 1.233 - Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology for Non-LWR Licensing Basis
- RG 1.253 - Guidance for Technology-Inclusive Content-of-Application Methodology for Non-LWR Licensing 4
- 2. KU1 PRA Overview Plant risks and evaluation criteria:
- Total frequency of events exceeding 100 mrem at site boundary 1/plant-year
- Average individual risk of early fatality within 1 mile of EAB 5x10/plant-year
- Average individual risk of latent cancer fatality within 10 miles of EAB 2x10/plant-year KU1 CP-stage PRA supports:
- Reliability and integrity management (RIM)
- Design reliability assurance program (D-RAP) 5
KU1 PRA Overview (Cont.)
- 1. Plant Operating State Analysis
- 2. Initiating Event Analysis
- 3. Event Sequence Analysis
- 4. Success Criteria Analysis
- 5. Systems Analysis
- 6. Human Reliability Analysis 6
KU1 CP-stage PRA addresses 12 ASME/ANS PRA standard technical elements:
7.
Data Analysis 8.
Hazard Screening 9.
Event Sequence Quantification
- 10. Mechanistic Source Term Analysis
- 11. Radiological Consequence Analysis
- 12. Risk Integration
KU1 PRA Overview (Radionuclide Source Terms)
Radionuclide sources included in the KU1 PRA:
Reactor enclosure system, including spent fuel in-vessel storage Primary sodium processing system and sodium cover gas systems Intermediate sodium processing system Gaseous radiological waste system Ex-vessel handling machine and storage tank Spent fuel processing systems and pool 7
- 3. Staff Review Approach Key areas evaluated in an integrated manner:
- 1. PRA Scope: internal events, radiological source terms, plant operating states, risk metrics
- 2. Level of Detail: modeling to capture plant behavior, system interactions, and interdependencies
- 4. Plant Representation & Configuration Control: accuracy of design representation and mechanisms in place to manage changes 8
Staff Review Approach (Cont.)
Staffs review confirms:
- PRA consistent with preliminary design and site
- Modeling assumptions reasonable and documented
- Methods consistent with industry practices
- Risk profile reasonable, including risk-significant contributors
- Supports LMP methodology, licensing basis development, and CPA
- Adequate configuration control ensures PRA maintenance throughout detailed design and construction, in preparation for the OLA 9
Staff Review Approach (Documentation &
Audit)
KU1 PSAR PRA content limited; the staff supplemented its review using additional documentation made available in the regulatory audit Virtual and three-day in-person audit conducted Over 40 PRA-related documents posted in eRR PRA model developed using CAFTA software Sensitivity analyses performed to understand uncertainties Self-assessment against Non-LWR PRA Standard ASME/ANS RA-S-1.4-2021 and RG 1.247 10
- 4. PRA Review and Confirmation: POS &
Initiating Events PRA Element 1 - Plant Operating State Analysis (POS)
- POSs represent distinct, stable plant conditions and were reasonably defined
- LPSD evolutions segmented based on plant response
- POSs characterized using design information
- Sources of uncertainty identified
- IEs logically grouped by mitigation similarity
- Most IE frequencies reasonably estimated
- Uncertainties identified
- Documentation provides traceable methodology and results 11
PRA Review and Confirmation: Event Sequence & Success Criteria PRA Element 3 - Event Sequence Analysis (ES)
Barriers and safety functions reasonably defined Functional dependencies reflected Success criteria and mission times considered Relevant modeling uncertainties identified Documentation ensures traceability PRA Element 4 - Success Criteria Analysis (SC)
- Detail level reasonable for CP-stage
- FSFs, supporting SSCs, and operator actions addressed
- Uncertainties identified
- Documentation ensures technical traceability 12
PRA Review and Confirmation: Systems Analysis & HRA PRA Element 5 - Systems Analysis (SY)
Models reflect as-designed, as-to-be-built, and as-to-be-operated plant Support systems and dependencies included Active/passive components, failure modes, common cause failures, and human errors incorporated Mission times and component actuations considered Uncertainties identified and documented Documentation ensures traceability PRA Element 6 - Human Reliability Analysis (HR)
- HRA analyzed on POS-by-POS basis
- Pre/post-initiator HFEs affecting mitigation considered
- No credit for recovery actions
- Human error probabilities estimated with dependencies
- Uncertainties identified and evaluated
- Documentation ensures traceability 13
PRA Review and Confirmation: Data Analysis
& Hazard Screening PRA Element 7 - Data Analysis (DA)
- Parameter estimation incorporated generic and design-specific data
- Parameters defined relative to PRA logic and probability framework
- Component boundaries established
- Uncertainties identified
- Documentation sufficient for CP-stage; OL-stage traceability to be added PRA Element 8 - Hazard Screening (HS)
- Hazards systematically identified, including site-and design-specific
- Qualitative and quantitative screening applied
- Decisions supported by data
- Uncertainties identified
- Documentation ensures traceability 14
PRA Review and Confirmation: ES Quantification & MS PRA Element 9 - Event Sequence Quantification (ESQ)
- Integrated modeling elements
- Event sequences quantified with low truncation limits
- Risk-significant contributors identified
- Uncertainties addressed
- Documentation ensures traceability PRA Element 10 - Mechanistic Source Term (MS)
- Releases reasonably grouped into categories
- Timing, location, magnitude, barriers, and transport assessed
- Computational tools applied to the extent practicable
- Uncertainties identified
- Documentation ensures traceability 15
PRA Review and Confirmation: Radiological Consequence & Risk Integration PRA Element 11 - Radiological Consequence Analysis (RC)
- Analysis conducted using justified methods and tools
- TEDE and associated uncertainties quantified
- Documentation ensures traceability PRA Element 12 - Risk Integration (RI)
- Assumptions identified
- Impacts on sequence frequencies and consequences evaluated
- Sensitivity analyses performed
- Uncertainties characterized
- Documentation ensures traceability 16
KU1 Spent Fuel PRA
- IE identification, event trees, fault trees, quantification, and uncertainty evaluation included
- Release frequencies estimated from single fuel pin damage to large-scale SFP overheating
- PRA will be refined as design/operational details become available
PRA Configuration Control Program Generally, CCP addressed key elements consistent with the non-LWR PRA standard, including:
Monitors plant design, operations, PRA methods, and industry experience Collects and updates data Maintains/updates PRA for as-designed, as-built, and as-operated plant conditions Evaluates cumulative impact of pending changes Controls configuration of computer codes and files Documentation ensures traceability Overall, USOs PRA CCP reasonably developed. Establishes effective processes and controls to maintain and enhance the PRA as the plant design evolves.
18
Assessment vs RG 1.253 RG 1.253 provides guidance for CP-stage non-LWR PRA for a CPA implementing LMP methodology in NEI 18-04 Guidance on the interrelated areas of PRA acceptability and documentation, such as:
scope, level of detail, elements, plant representation, configuration control, and documentation Staff confirms KU1 PRA meets regulatory positions in Appendix A of RG 1.253:
A.2 - General A.3 - PRA Scope A.4 - PRA Elements A.5 - PRA Level of Detail A.6 - Plant Representation and PRA Configuration Control A.7 - PRA Documentation A.8 - Demonstrating PRA Acceptability 19
- Completing internal/external hazards (internal fire, internal floods, seismic, tornadoes, volcanoes, wind)
- Reassessing and resolving assumptions
- Updating PRA to address self-assessment findings and staff observations and include other elements from ASME/ANS PRA standard
- Integrating models and requantifying risks
- Conducting peer review and addressing findings and observations
- NRC staff engagement continues to:
- Track resolution of staff, self-assessment, and peer review findings and observations
- Ensure configuration control and final plant representation
- Confirm OL-stage PRA scope, level of detail, and methodology adequacy 20
- 6. Conclusion
- KU1 CP-stage PRA consistent with guidance and standards, demonstrating low risk and providing important insights
- Living document; assumptions, hazards, and detail updated as design matures
- NRC staff intends to ensure at OL:
- Updated scope, level of detail, and assumptions
- Peer review and resolution of findings
- Configuration control implementation 21
Functional Containment, Mechanistic Source Term, and Consequence Analysis Michelle Hart, Senior Reactor Engineer, NRR/DANU/UTB2 Tracy Radel, Senior Nuclear Engineer, NRR/DANU/UTB2 22
Topics
- Functional containment
- Guidance and tie of LMP to functional containment
- KU1 functional containment strategy
- Primary/enveloping barriers
- Assumed barrier performance in analyses
- Mechanistic source term (MST)
- Consequence analysis 23
Functional Containment Retaining Radionuclides Fundamental Safety Function Accomplished in KU1 by a functional containment strategy that employs diverse passive barriers SECY-18-0096: Functional containment is a barrier, or set of barriers taken together, that effectively limits the physical transport of radioactive material to the environment Functional containment performance criteria are developed to address each barriers role in mitigating releases to meet plant-level performance criteria (regulatory dose requirements)
The NEI 18-04 LMP methodology is consistent with the approach described in the enclosure to SECY-18-0096 and provides the necessary information to develop the functional containment performance criteria MST modeling of the radionuclide barriers and release rates for the LBEs informs the functional containment performance criteria by providing details on the barrier performance assumed in the radiological consequence analyses 24
KU1 Functional Containment Strategy The diverse passive barriers that are used in the functional containment strategy begin at a radionuclide source and include all SSCs between that source and the environment 25 Fuel Cladding (SR)
Reactor Enclosure System (SR)
Head Access Area (NSRST)
Primary Enveloping At-power In-vessel Events Fuel Cladding (SR)
Reactor Enclosure System + Fuel Handling Equipment (SR)
In-vessel Fuel Handling Events System Barrier (NSRST)
Enveloping Cell (NSRST or NST)
Ex-vessel Events
Primary and Enveloping Functional Containment Barriers Primary Barrier
- An SSC, or portion of an SSC, which is required to perform a radionuclide retention function to keep offsite doses of DBA scenarios within 10 CFR 50.34(a)(1) dose criteria and/or keep the consequence of the associated DBE from violating the F-C target curve
- SSCs serving as a Primary Barrier are generally safety related Enveloping Barrier
- An SSC, or portions of an SSC, which provides a backup radionuclide retention function to Primary Barriers or the Primary Functional Containment Boundary, in the event of leakage or failure of the Primary Barriers it envelopes
- Enveloping Barriers working alone or in tandem with other barriers limit the radiological release in AOOs, DBEs, and BDBEs to below established dose criteria for the event type
- Typically, Enveloping Barriers are NSRST or NST 26
Functional Containment Barrier Release Rates
- Event-specific mechanistic source terms model release across barriers
- Releases from the fuel are consistent with ANL reports
- Consistent with approved source term methodology
- Leakage rates from other barriers are assumed values for the PSAR
- Barrier release rate assumptions will be updated as necessary based on the final design for the FSAR
- FSAR MST analyses will have barrier release modeling based on final design information and will inform the functional containment performance criteria for the OL application 27
Functional Containment Conclusions
- Staff concluded that the overall functional containment approach is acceptable to apply to the KU1 design
- Staff did not come to a final determination on the adequacy and acceptability of functional containment performance due to the preliminary nature of the design and analysis
- Staff will confirm the acceptability of the functional containment performance criteria in the OL application review 28
Mechanistic Source Terms
- Specific evaluation of the justification of user inputs that were not provided in NAT-9392-A Staff reviewed the applicants implementation of the NAT-9392-A TR methodology to develop the LBE source terms
- Some LBE consequence analyses applied a representative source term based on another related scenario
- Some also with additional source term scaling factors to account for different LBE conditions Limited number of source terms for the PSAR
- Release to environment over time
- Lists isotopes which contribute to at least 95 percent of the total dose
- Cumulative activity release of each isotope in curies at time since initiation PSAR source term tables 29
Source Term Modeling
- Radionuclide inventories for the sources of radiological material at risk of release (MAR) in the fuel, primary sodium and cover gas are conservatively estimated using approved methods
- PSAR LBE MSTs modeled the release rates of MAR from the initial barrier (e.g., fuel pin) into subsequent spaces defined by physical barriers (e.g.,
reactor vessel system, head access area, reactor building) for the relevant event
- Modeling of radionuclide transport and removal phenomena within the barriers (e.g., sodium pool aerosol scrubbing) use approved methods
- Based on preliminary design information,
- Assumptions consistent with data on source term phenomena for SFRs, or
- Assumptions that would conservatively bias the transport and release 30
Source Term Uncertainty Treatment
- PSAR LBE scenarios include many conservatively biased assumptions in the progression including the amount and timing of fuel damage or other MAR release rates
- LBE MST analyses also include many conservatively biased assumptions and code input to model the radionuclide transport across the release barriers and radionuclide retention and removal phenomena within the sodium coolant and in the gas space
- Release rates across barriers were based on preliminary design information or with conservative assumptions
- Expectation that these barrier release rate assumptions will be updated in the FSAR analyses based on the final design, as necessary 31
Staff Analyses
- Staff performed calculations on the LBE MSTs to ascertain the effective radionuclide retention within the credited barriers for the event
- Included estimation of the effective removal rate of each of the radionuclide transport and removal phenomena within those barriers, such as bubble scrubbing in the sodium pool, SFP scrubbing, aerosol natural deposition, and filtration system
- Staffs analysis of the LBE MSTs and radionuclide retention aided in the staff's review of the applicant's use of the LMP process to classify SSCs and evaluate the relationship to the retaining radionuclides fundamental safety function and the Natrium reactor functional containment strategy 32
Source Term Conclusions
- Staff determined that USO addressed the NAT-9392-A TR limitations and conditions sufficiently for the CP stage
- PSAR analyses are acceptable for the CP stage and further considerations are reasonably left to the OL application review
- Scenario-specific LBE source terms based on final design information and approved analysis methods will be developed to support the OL application
- MST evaluation model treatment of sodium pool scrubbing and iodine releases will be further examined for the OL stage to confirm the implementation and supporting methodology is appropriate for the application 33
Consequence Analyses Staff reviewed the applicants implementation of the NAT-9391-A approved methodology to develop event-specific consequence analyses DBAs and Major Accident
- Enveloping 95th percentile short-term /Q values in lieu of site characteristic values for EAB, LPZ and main control room
- Developed based on the evaluation of atmospheric dispersion factors for a range of power reactor sites
- Shown to be bounding of the site characteristic values for each time period and result in bounding doses for KU1 Non-DBA LBEs (AOOs, DBEs, and BDBEs)
- Generic meteorological data based on Annex B of the EPRI URD representative of a reasonable number of sites in the contiguous U.S.
- Staffs judgment is that use of the generic meteorological data based on the EPRI URD in the non-DBA LBE consequence analyses will result in doses that are likely to be bounding for KU1 and sufficient to use in the NEI 18-04 risk-informed process
- At the OL stage the meteorological data used to evaluate the radiological consequences of AOOs, DBEs, and BDBEs must be shown to be conservatively representative of the site-specific meteorological data
- Generic meteorological data also used in PRA, including OQEs 34
Consequence Analysis for Integrated Risk Assessment
- PSAR analysis deviates from the NAT-9391-A methodology because the long-term phase was not evaluated
- The method for estimating the consequences of chronic exposure was added during the TR review, after the PSAR analyses were done
- Only the evaluation of the average individual risk of latent cancer fatality is affected by the omission of chronic exposure in the integrated risk assessment
- Increase in consequence is expected to be well within the PSAR margin to the LMP cumulative risk metric target
- Deviation from the consequence methodology with respect to chronic exposure is not likely to challenge the LMP individual risk of latent cancer cumulative risk target
- The long-term phase chronic exposure will be evaluated for the OL application 35
Consequence Uncertainty Treatment
- The LBE consequence analyses implemented the NAT-9391-A TR methodology, including treatment of uncertainty
- Atmospheric transport and dispersion modeling
- DBAs : Use conservative 95th percentile /Q values
- Non-DBA LBEs: Evaluation of weather uncertainty through weather trials based on 1 year of meteorological data using the consequence calculation code consistent with NAT-9391-A
- Other uncertain parameters in the consequence analyses were handled with bounding conservative values
- Source term uncertainty for input to the non-DBA LBE consequence analyses handled in most cases by using
- Nominal source terms to generate the mean and 5th percentile doses
- Limiting source terms from a series of source term uncertainty runs to generate 95th percentile doses 36
Consequence Analysis Conclusions
- Staff determined that USO addressed the NAT-9391-A TR limitations and conditions sufficiently for the CP stage
- PSAR analyses provides results that are representative of the potential radiological consequences of events at KU1 for use in the LMP and are acceptable for the CP stage
- Further considerations are reasonably left to the OL application review 37
Major Accident Michelle Hart, Senior Reactor Engineer, NRR/DANU/UTB2 Tracy Radel, Senior Nuclear Engineer, NRR/DANU/UTB2 38
Regulatory Requirement
- 10 CFR 50.34(a)(1)(ii)(D) requires, in part, that the PSAR analyses consider safety features engineered into the facility and barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur
- Evaluation and analysis of the postulated fission product release from the core into the containment
- Footnote 3: assumed fission product release should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events
- Precedents used a deterministically-evaluated fission product release with consequences bounding all DBAs 39
Regulatory Criteria
- The plant design features intended to mitigate the radiological consequences of accidents, the site atmospheric dispersion characteristics, and the siting boundary distances are acceptable if the total calculated radiological consequences for the postulated fission product release meet the reference values for public dose given in 10 CFR 50.34(a)(1)(ii)(D):
(1) An individual located at any point on the boundary of the exclusion area (EAB) for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE)
(2) An individual located at any point on the outer boundary of the low population zone (LPZ), who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE 40
RG 1.253 Guidance for LMP-based License Applications
- LMP analyses do not directly address the requirement
- RG 1.253, staff position C.3.b, described two potential approaches to establish the acceptability of the EAB and LPZ in the context of LMP 41
- DBA analysis under an LMP-based approach is a deterministic, conservative analysis that is analogous to the DBA analyses performed for operating reactors and new LWRs
- Depending on the nature of the DBA, the application may need to include an exemption from the regulations in 10 CFR 50.34 or 10 CFR 52.79 that require an assumed major accident Option 1: Use the DBA dose consequence results from an LMP-based approach
- Precludes the need for an exemption from these requirements, as long as the bounding BDBE involves or bounds an event sequence meeting the description of a major accident Option 2: Use the greater of the dose consequence results from the bounding DBA and from a bounding BDBE, as identified in the LMP-based approach
Basis for KU1 Major Accident 42
- PSAR states that USO is using Option 1 as a basis for determining the Major Accident
- USO narrowed the consideration to in-vessel DBAs to be consistent with the 10 CFR 50.34(a)(1) description of the accident analysis as evaluating the postulated fission product release from the core into the containment
- Selected the core blockage and local faults DBA as the major accident
- At-power in-vessel scenario with releases from core damage
- Maximum core damage and associated in-core release of the at-power in-vessel DBAs
- Addresses the description of the major accident postulated fission product release in the footnote to 10 CFR 50.34
Staff Evaluation of Major Accident
- Offsite radiological consequences of the selected major accident are bounding for at-power in-vessel DBAs
- However, the major accident consequences are not bounding of all DBAs
- Fuel handling DBAs have higher offsite dose results
- Additional justification needed to address this departure from precedent
- LMP provides a more detailed and comprehensive look at the plant design safety and siting than potentially given by a bounding major accident deterministic evaluation
- All DBA analyses meet the 10 CFR 50.34 offsite dose criteria
- Staffs judgment is that the PSAR DBAs, as supported by evaluation of the LBEs, comprehensively address the safety and siting analyses consequence analysis requirement 43
Conclusions on Major Accident
- Staff determined that the PSAR described an evaluation of the major accident for the Kemmerer U1 facility that addresses the safety analysis regulatory requirements related to the evaluation of the radiological consequences at the EAB and LPZ
- Staff determined that the PSAR discussion of the major accident is consistent with the guidance for LMP-based safety analysis report contents given in RG 1.253 44
Acronyms 45 ACRS - Advisory Committee on Reactor Safeguards ANL - Argonne National Lab ANS - American Nuclear Society AOO - Anticipated Operational Occurrence ASME - American Society of Mechanical Engineers ARCAP - advanced reactor content of application BDBE - Beyond Design Basis Event CAFTA - Computer Aided Fault Tree Analysis System CCP - Configuration Control Program CFR - Code of Federal Regulations CP - Construction Permit CPA - Construction Permit Application DBA - Design Basis Accident DBE - Design Basis Event DID - Defense In Depth D-RAP - Design Reliability Assurance Program EAB - Exclusion Area Boundary EPRI - Electric Power Research Institute ERR - Electronic Reading Room ES - Event Sequence F-C - Frequency-Consequence POS - Plant Operating State PRA - Probabilistic Risk Assessment PSAR - Preliminary Safety Analysis Report PSID - Preliminary Safety Information Document PSF - PRA Safety Functions QHO - Quantitative Health Objective RG - Regulatory Guide RIM - Reliability and Integrity Management RSF - Required Safety Functions SC - Success Criteria SE - Safety Evaluation SFR - Sodium Fast Reactor SFP - Spent Fuel Pool SR - Safety Related SRDC - Safety Related Design Criteria SSC - Structures, Systems, and Components TEDE - Total Effective Dose Equivalent TICAP - technology inclusive content of application TR - Topical Report URD - Utility Requirements Document USO - US SFR Owner FSAR - Final Safety Analysis Report F+O - Findings and Observations HRA - Human Reliability Analysis IDP - Integrated Decision-making Process IDPP - Integrated Decision-making Process Panel IE - Initiating Event KU1 - Kemmerer Unit 1 LBE - Licensing Basis Event LMP - Licensing Modernization Project LPZ - Low Population Zone LWR - Light Water Reactor MAR - Material at Risk MST - Mechanistic Source Term NEI - Nuclear Energy Institute NRC - Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety-related with Special Treatment NST - No Special Treatment OL - Operating License OLA - Operating License Application OQE - Other Quantified Events