ML25280A085

2 - Lmp Overview

ML25280A085
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	10/08/2025
From:	Reed Anzalone, Hart M, Hanh Phan, Radel T NRC/NRR/DANU/UTB2
To:
References
Download: ML25280A085 (1)

Text

NRC Review of Kemmerer Unit 1 Construction Permit Application LMP Overview Tracy Radel, Reed Anzalone, Hanh Phan, Michelle Hart NRR/DANU/UTB2 ACRS Subcommittee Meeting October 8-9, 2025

Topics

• Probabilistic risk assessment (PRA) development

• Supporting analyses

• Licensing basis event (LBE) selection and categorization

• Identification of safety-significant PRA safety functions (PSFs)

• Safety classification of structures, systems, and components (SSCs)

• Defense-in-depth (DID) adequacy evaluation

• Identification of design requirements and selection of special treatments 2

From PRA to Design of SSCs Under the Licensing Modernization Project (LMP)

• PRA used in a multi-step process to

• Identify and categorize LBEs

• Identify required safety functions (RSF) and select associated safety-related (SR) SSCs

• Identify risk-significant non-safety-related with special treatment (NSRST) functions and associated SSCs

• Identify safety-significant, NSRST functions and associated SSCs required to meet DID adequacy PRA and Supporting Analyses LBE Selection and Categorization Safety-Significant PSFs SSC Safety Classification Design Requirements and Special Treatments 3

It is all connected Design Probabilistic Risk Assessment (PRA)

Licensing Basis Events (LBEs)

Classification and Special Treatments and all iterative 4

PRA Development

• RG 1.253, Appendix A provides guidance on the acceptability of a PRA that supports an advanced reactor construction permit application

• Thorough and systematic approach with strong documentation

• Comprehensive identification of initiating events, should use appropriate methods to evaluate the unique plant design

• The design is not complete at the construction permit stage, therefore the PRA will not be complete

• Unavailability of design information generally handled in two ways: design scope excluded from the PRA or placeholders added

• Tracking of assumption and open items is critical 5

Supporting Analyses 6

Consequence Results Neutronics/

Reactivity Thermal Hydraulics Fuel Performance Mechanistic Source Term Radionuclide Transport Criticality Safety

• Detailed analyses used to support the determination of consequence values used on the Frequency-Consequence (F-C) chart

• Uncertainties in these analyses should be appropriately propagated to the final consequence results, contributing to the uncertainty bands

• Alternatively, conservative values may be used

Identification and Categorization of LBEs LBE Type Minimum Frequency [per plant year]

Maximum Frequency [per plant year]

Anticipated operational occurrence (AOO) Mean frequency > 1.0E-02 N/A Design basis event (DBE)

Mean frequency > 1.0E-04 Mean frequency <1.0E-02 Beyond design basis event (BDBE) 95th percentile frequency > 5.0E-07 Mean frequency < 1.0E-04

• Risk-significant LBEs are LBEs with a consequence exceeding 2.5 mrem that are within 1% of the F-C Target. This evaluation should be done using the 95th percentile frequency and consequence values.

• Design basis accidented (DBAs) are generated for DBEs and AOOs and BDBEs with uncertainty bands that extend into the DBE region. The DBA is evaluated in a bounding way, with only safety-related SR SSCs available to perform their functions. If a SR SSCs has failed within the event being mapped into the DBA, it must also be evaluated as failed in the DBA.

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Frequency-Consequence Chart

• AOOs, DBEs, BDBEs, and DBAs are collectively known as LBEs

• DBAs are evaluated deterministically and not plotted on the F-C chart

• Distribution on frequency comes from failure probabilities in PRA

• Distribution on consequence comes from uncertainty in plant states, fractions that escape confinement,

/Q dispersion factors, etc.

PSAR Figure 3.5-1. F-C Chart for LBEs with Uncertainty Bands 8

Initiating Event Families

• Grouping licensing basis events (LBE) into initiating event (IE) families can provide insights into:

• Relationships between LBEs

• What LBEs may not be included in the application

• Impacts of individual PRA safety functions (PSFs)

• Preventive, mitigative, or both 9

INL/EXT-20-60396 Figure 3-1. Capability and Reliability of an SSC to Mitigate and Prevent LBEs INL/EXT-20-60396 Table 3-2. Performance Attributes for SSC Prevention and Mitigation Functions

Fundamental Safety Functions (FSFs)

• PSFs contribute to meeting one of three fundamental safety functions:

• Controlling heat generation (RC)

• Controlling heat removal (HR)

• Retaining radionuclides (RR) 10

Identification of Safety-Significant PSFs

• Evaluations that assume failure of PSFs and assess the impacts on individual LBEs

• PSFs necessary to maintain LBEs below the F-C target are risk-significant

• Can be SR or NSRST depending on exact evaluation

• Additional risk-significant NSRST functions identified through integrated risk evaluation

• SR but not risk-significant PSFs are identified through DBA analysis

• Additional NSRST PSFs identified through DID evaluation 11 NEI 18-04 Figure 4-2. Definition of Risk-Significant and Safety-Significant SSCs

Safety Classification of SSCs

• SSCs are classified based on the functions they perform

• Some SSCs may perform more than one function

• Safety classification based on the most limiting function (SR > NSRST > NST)

• If the SSC does not perform any safety-significant functions, it is non-safety-related with no special treatment (NST)

• All safety-significant functions must be accounted for in the design requirements and selection of special treatments 12 Designers have the flexibility to select the SSCs that will be credited to perform the RSFs (SR functions) to optimize the set of SR SSCs INL/EXT-20-60396 Table A-2. Evaluation of MHTGR SSCs for Core Heat Removal Safety Function

DID Adequacy Evaluation

• Evaluation of uncertainties and margins, compensates for residual unknowns

• Determines whether any single feature is excessively relied upon to achieve public safety objectives

• Confirms that a balance between event prevention and mitigation is reflected in the layers of defense for risk-significant LBEs

• Results reviewed by integrated decision-making process panel (IDPP) 13 NEI 18-04 Table 5-2. Guidelines for Establishing the Adequacy of Overall Plant Capability Defense-in-Depth

Design Requirements and Special Treatments

• SR design criteria (SRDC) established for each SR function (i.e., RSF)

• Reliability and capability targets developed for SR and NSRST SSCs from performance relied on in the PRA, iteration between PRA and design

• Codes and standards, additional design requirements, and special treatments are selected to ensure reliability and capability targets are met

• Special treatments go beyond the application of normal industrial codes and standards and may include:

• Programmatic special treatments

• Adding requirements beyond what is required by the commercial code (e.g.,

additional inspections and testing)

• Selection of a nuclear codes and standard 14

Summary

• LMP provides a systematic safety analysis process that is technology independent

• The PRA forms the foundation for the process

• The process is iterative and should be exercised continuously as the design develops to achieve the most efficient safe design

• Review by the IDPP is an important step to ensure completeness and accuracy, especially prior to final design when there may be missing scope or large uncertainties due to design information being unavailable 15

Acronyms ACRS - Advisory Committee on Reactor Safeguards AOO - Anticipated Operational Occurrence ARCAP - advanced reactor content of application BDBE - Beyond Design Basis Event CP - Construction Permit DBA - Design Basis Accident DBE - Design Basis Event DID - Defense In Depth F-C - Frequency-Consequence IDPP - Integrated Decision-making Process Panel IE - Initiating Event LBE - Licensing Basis Event LMP - Licensing Modernization Project NRC - Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation NSRST - Non-safety-related with Special Treatment PRA - Probabilistic Risk Assessment PSF - PRA Safety Functions PSAR - Preliminary Safety Analysis Report QHO - Quantitative Health Objective RG - Regulatory Guide RSF - Required Safety Functions SE - Safety Evaluation SR - Safety Related SRDC - Safety Related Design Criteria SSC - Structures, Systems, and Components TICAP - technology inclusive content of application 16