ML24345A006

Safety Evaluation for SMR, LLC Licensing Topical Report HI-2230875, Revision 1, Holtec PSA Risk Significance Determination Methodology Licensing Topical Report

ML24345A006
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HI-2230875, Rev 1
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1 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to SMR, LLC Licensing Topical Report HI-2230875, Revision 1, Holtec PSA Risk Significance Determination Methodology Licensing Topical Report 1.0 Introduction By letter dated October 18, 2024 (Ref. 1), SMR, LLC (A Holtec International Company, hereafter referred to as Holtec) submitted licensing topical report (TR) HI-2230875, Revision 1, Holtec PSA Risk Significance Determination Methodology Licensing Topical Report, (Ref. 2), to the U.S. Nuclear Regulatory Commission (NRC) staff for review and approval. This TR describes Holtecs proposed methodology for identifying candidate risk-significant structures, systems, and components (SSCs) using the SMR-300 probabilistic risk assessment (PRA) and the basis for the risk significance criteria used. This methodology is specific to the SMR-300 design and uses alternative risk significance criteria that deviate from Regulatory Guide (RG) 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, (Ref. 3).

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, (SRP), Section 19.0, Revision 3, Probabilistic Risk Assessment and Severe Accident Evaluations for New Reactors, (Ref. 4), states that the term significant is intended to be consistent with the definition provided in RG 1.200, and any other definition shall be subject to additional NRC staff review and approval. This safety evaluation (SE) describes the NRC staffs review and approval of the TR and the limitations and conditions on its use.

2.0 Regulatory Criteria The NRC staff considered the following regulatory guidance during its review of the TR:

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, (Ref. 5),

describes an approach that is acceptable to the NRC staff for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights.

RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, describes an approach that is acceptable to the NRC staff for determining whether a base PRA, in total or in the portions that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, (Ref. 6),

endorses NEI 00-04, 10 CFR 50.69 Categorization Guideline, (Ref. 7), and describes an approach that is acceptable to the NRC staff for complying with the Commissions requirements in Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-informed categorization and treatment of structures, systems and components for

2 nuclear power reactors, with respect to the categorization of SSCs that are considered in risk-informing special treatment requirements.

SRP Section 17.4, Revision 1, Reliability Assurance Program, (Ref. 8), provides the NRC staff review guidance for the reliability assurance program description in design certification and combined license applications.

SRP Section 19.0, Revision 3, Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors, provides the NRC staff review guidance for the design-specific PRA for a design certification and the plant-specific PRA for a combined license application.

SRP Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, dated June 2007 (Ref. 9),

provides the NRC staff review guidance for risk-informed reviews of licensees proposals for changes to the licensing basis of nuclear power plants.

3.0 Summary of Technical Information Licensing TR HI-2230875 describes Holtecs proposed methodology for identifying candidate risk-significant SSCs for the SMR-300 design using the SMR-300 PRA. Holtec stated that this methodology is specific to the SMR-300 design, and it can only be used with a PRA and analysis of core damage frequency (CDF) and either large release frequency (LRF) or large early release frequency (LERF), as applicable, that is technically adequate.

In the TR, Holtec stated that RG 1.200 discusses the term significant as it relates to the relative risk criteria and defines basic events that have a Fussell-Vesely (FV) importance greater than 0.005 or a risk achievement worth (RAW) greater than 2 as significant. Holtec described that the relative risk criteria used in RG 1.200 for FV and RAW are based on the relative risk associated with the operating fleet of reactors and, therefore, do not account for the lower risk profile of the passive SMR-300 design. Holtec stated that applying the relative risk criteria in RG 1.200, which are determined as a ratio to the total CDF of LRF/LERF, to the SMR-300 design would artificially elevate the significance of SSCs that do not have commensurate contribution to risk and that this artificially inflated significance of SSCs would not be risk-informed because it would result in unnecessary resource allocation for both the licensee and regulatory staff.

Holtec proposed a methodology for identifying candidate risk-significant SSCs for the SMR-300 design that adjusts the thresholds for RAW and FV based on the baseline CDF and LRF to ensure that measurable contributors to risk are identified regardless of the risk profile. Holtec stated that this methodology is based on the approach in RG 1.174 for acceptable increases in risk based on the baseline risk. Holtecs proposed methodology is summarized in the following sections.

3.1 Risk Achievement Worth Criteria The proposed methodology applies the following RAW criteria at a single-unit level for CDF and LRF. The RAW criteria are applicable to all initiating events collectively and aggregated across all hazards and operating modes (i.e., internal events, low-power and shutdown conditions, internal flooding, internal fires, and external hazards).

3 The proposed methodology considers the RAW for basic events representing equipment unavailability and human failure. The proposed methodology does not consider the RAW for internal initiator basic events or external initiator basic events.

3.1.1 Core Damage Frequency Holtec proposed the following RAW criteria for identifying component-level basic events as candidate risk significant using CDF:

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a component-level basic event with a RAW greater than 5 is identified as candidate risk significant.

When the baseline CDF is less than 1x10-7 per year, a component-level basic event with a RAW greater than 30 is identified as candidate risk significant.

The proposed methodology implements a tailored approach that increases the component-level RAW criteria from the current criterion in RG 1.200 and is consistent with the criterion approved for the NuScale design (Ref. 10) based on the baseline CDF. Holtec stated that the proposed component-level RAW criteria meet the intent of RG 1.200 and RG 1.174.

For system-level basic events (i.e., basic events that represent a common-cause failure (CCF) of the system), Holtec stated that it considered the criterion that a CCF event is risk significant if it has a RAW greater than 20, which is contained in NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline, and endorsed by RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance. Holtec stated that most systems expected to provide or assist safety missions in the SMR-300 design typically include some inter-and intra-system redundancy, which was the rationale for using an order of magnitude increase (i.e., a multiplication factor of 10) for the system-level criterion compared to the component-level criterion. To ensure conservatism, Holtec reduced the multiplication factor as the baseline CDF decreases:

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a multiplication factor of 7 is used, so a system-level basic event with a RAW greater than 35 is identified as candidate risk significant.

When the baseline CDF is less than 1x10-7 per year, a multiplication factor of 2 is used, so a system-level basic event with a RAW greater than 60 is identified as candidate risk significant.

Holtec stated that the proposed system-level RAW criteria are similar to the criteria approved for the economic simplified boiling water reactor (ESBWR) design (Ref. 11) and the criteria approved for the NuScale design (Ref. 10). In the approved ESBWR methodology, a system-level basic event with a RAW greater than 50 is identified as risk significant. In the approved NuScale methodology, a system-level basic event with a CDF conditional on the failure of the basic event greater than 1x10-5 per year is identified as candidate risk significant, and the conditions and limitations require the core damage frequency to be very low (i.e., approximately 1x10-7 per year or less). When the baseline CDF is 1x10-7 per year, a conditional CDF of 1x10-5 per year is equivalent to a RAW greater than 100. Holtec stated that the proposed system-level RAW criteria meet the intent of RG 1.200 and RG 1.174.

4 3.1.2 Large Release Frequency The proposed methodology uses the same approach for LRF as it does for CDF, except the baseline LRF values are reduced by an order of magnitude from the baseline CDF values. This is consistent with the Commissions conditional containment failure probability (CCFP) goal of less than 0.1 for new reactors (Ref. 12) and the approach in RG 1.174.

Holtec proposed the following RAW criteria for identifying component-level basic events as candidate risk significant using LRF:

When the baseline LRF is greater than or equal to 1x10-8 and less than 1x10-7 per year, a component-level basic event with a RAW greater than 5 is identified as candidate risk significant.

When the baseline LRF is less than 1x10-8 per year, a component-level basic event with a RAW greater than 30 is identified as candidate risk significant.

Holtec proposed the following RAW criteria for identifying system-level basic events as candidate risk significant using LRF:

When the baseline LRF is greater than or equal to 1x10-8 and less than 1x10-7 per year, a system-level basic event with a RAW greater than 35 is identified as candidate risk significant.

When the baseline LRF is less than 1x10-8 per year, a system-level basic event with a RAW greater than 60 is identified as candidate risk significant.

Holtec stated that the proposed methodology is based on LRF since LRF and CCFP are used during modern application reviews. Since the Commission approved the NRC staffs recommendation to transition from LRF to LERF at or before initial fuel load and discontinue regulatory use of LRF and CCFP thereafter in SRM-SECY-12-0081, Staff Requirements -

SECY-12-0081 - Risk-Informed Regulatory Framework for New Reactors, (Ref. 13), the proposed methodology uses the same criteria for LERF as it does for LRF. Holtec stated that this is conservative based on the LRF goal of less than 10-6 per year being more restrictive than the LERF goal of 10-5 per year.

3.2 Fussell-Vesely Criteria To supplement the RAW criteria, the proposed methodology uses the FV importance to identify those SSCs that have the largest fractional contribution to risk. Holtec stated that the focus of these criteria is on identifying SSCs for which reliability and availability have the greatest influence on the risk profile.

The proposed methodology applies the following FV criteria at a single-unit level for CDF and LRF. The FV criteria are applied individually to each hazard group and mode of plant operation.

The proposed methodology considers the FV for basic events representing equipment unavailability and human failure and for internal initiator basic events because they represent failures of plant components. The proposed methodology does not consider the FV for external initiator basic events because they do not represent failures of plant components.

5 The proposed methodology sums the FV for each basic event (failure mode) of an SSC (contributor) to calculate the total FV for the SSC. An SSC is identified as candidate risk significant if the total FV for the SSC exceeds the FV criteria.

3.2.1 Core Damage Frequency Holtec proposed the following FV criteria for identifying SSCs as candidate risk significant using CDF:

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a basic event with a FV greater than 0.02 is identified as candidate risk significant.

When the baseline CDF is less than 1x10-7 per year, a basic event with a FV greater than 0.2 is identified as candidate risk significant.

The proposed methodology implements a tailored approach that increases the FV criteria from the criterion in RG 1.200 to a criterion consistent with that approved for the NuScale design based on the baseline CDF. Holtec stated that the proposed component-level RAW criteria meet the intent of RG 1.200 and RG 1.174.

3.2.2 Large Release Frequency The proposed methodology uses the same approach for LRF as it does for CDF, except the baseline LRF values are reduced by an order of magnitude from the baseline CDF values. This is consistent with the Commissions CCFP goal of less than 0.1 for new reactors and the approach in RG 1.174.

Holtec proposed the following FV criteria for identifying SSCs as candidate risk significant using LRF:

When the baseline LRF is greater than or equal to 1x10-8 and less than 1x10-7 per year, a basic event with a FV greater than 0.02 is identified as candidate risk significant.

When the baseline LRF is less than 1x10-8 per year, a basic event with a FV greater than 0.2 is identified as candidate risk significant.

3.3 Summary of Risk Significance Criteria The proposed methodology for identifying candidate risk-significant SSCs is summarized in Table 1 and Table 2.

6 Table 1. Risk significance criteria using CDF RAW Baseline CDF (per year)

Component Level System Level FV 1x10-7 CDF < 1x10-6 5

35 0.02 CDF < 1x10-7 30 60 0.2 Table 2. Risk significance criteria using LRF RAW Baseline LRF (per year)

Component Level System Level FV 1x10-8 LRF < 1x10-7 5

35 0.02 LRF < 1x10-8 30 60 0.2 4.0 Technical Evaluation In the absence of specific review procedures for evaluating methods for assessing risk significance, the NRC staff identified the following three key areas of review for this TR:

selection of the risk metrics for assessing risk significance selection of the risk metric thresholds application of the risk metrics 4.1 Selection of the Risk Metrics for Assessing Risk Significance The proposed methodology for identifying candidate risk-significant SSCs for the SMR-300 design uses the relative risk metrics RAW and FV. The NRC staff finds the use of RAW and FV acceptable because it is consistent with the guidance in RG 1.200.

4.2 Selection of the Risk Achievement Worth Thresholds The proposed methodology for identifying candidate risk-significant SSCs for the SMR-300 design adjusts the thresholds for RAW based on the baseline CDF and LRF to ensure that measurable contributors to risk are identified regardless of the risk profile. Holtec stated that this methodology is based on the approach in RG 1.174 for acceptable increases in risk based on the baseline risk. The basis for the proposed thresholds for RAW is summarized in the following sections.

4.2.1 Core Damage Frequency Holtec provided the following basis for the proposed RAW criteria for identifying component-level basic events as candidate risk significant using CDF:

7 When the baseline CDF is greater than or equal to 1x10-6 per year, a component-level basic event with a RAW greater than 2 is identified as candidate risk significant. The basis for this criterion is RG 1.200, which utilizes the same RAW criterion for CDF.

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a component-level basic event with a RAW greater than 5 is identified as candidate risk significant. The basis for this criterion is that the current RAW criterion for CDF in RG 1.200 corresponds to an increased risk of 2x10-5 per year when the basic event is assumed to fail, based on an operating reactor with a nominal CDF of 1x10-5 per year.

When the baseline CDF is 1x10-6 per year, using (1) an absolute risk criterion of an increased risk of 2x10-5 per year when the basic event is assumed to fail would result in very few basic events identified as candidate risk significant and (2) a relative risk criterion of RAW greater than 2 would not represent a significant loss in safety margin with respect to the 1x10-4 per year safety goal for CDF. To account for the lower baseline CDF of the SMR-300 design, and still identify basic events that drive the risk, Holtec adjusted the threshold for RAW and proposed that a component-level basic event with a RAW greater than 5 is identified as candidate risk significant.

When the baseline CDF is less than 1x10-7 per year, a component-level basic event with a RAW greater than 30 is identified as candidate risk significant. The basis for the selection of this criterion is that it is equivalent to the risk criterion approved for the NuScale design. In the approved NuScale methodology, a component-level basic event with a CDF conditional on the failure of the basic event greater than 3x10-6 per year is identified as candidate risk significant. When the baseline CDF is 1x10-7 per year, a conditional CDF of 3x10-6 per year is equivalent to a RAW greater than 30.

Holtec stated that the proposed component-level RAW criteria provide sufficient margin to the NRC safety goal for CDF to account for PRA uncertainties. In the approval for the NuScale methodology, the NRC staff stated that the ratio of the 95th percentile to mean value for CDF was less than 10 for the nuclear power plants documented in NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, (Ref. 14), and the PRA results for two nuclear power plant designs certified by the NRC that include passive safety systems.

Holtec stated that the SMR-300 design is also expected to have a ratio of the 95th percentile to mean value for CDF that is less than 10 for the following reasons:

The SMR-300 design uses proven light-water reactor technology similar to the nuclear power plants analyzed in NUREG-1150.

The SMR-300 design relies on automatic actuation of passive safety systems to accomplish its safety functions.

PRA modeling practices used to evaluate the SMR-300 design are consistent with industry standard practices described in RG 1.200.

Holtec combined the proposed component-level RAW criteria with the expected uncertainty ratio of 10 to demonstrate sufficient margin to the NRC safety goal for CDF as follows.

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a RAW of 5 combined with an uncertainty ratio of 10 provides a factor of 2 margin to the NRC safety goal for CDF. Holtec described that using a larger threshold for RAW would

8 reduce or eliminate the margin and using a smaller threshold for RAW would be inconsistent with the intent to reduce the number of SSCs identified as candidate risk significant compared to the number of SSCs that would be identified as candidate risk significant using the current RAW criterion for CDF in RG 1.200.

When the baseline CDF is less than 1x10-7 per year, a RAW of 30 combined with an uncertainty ratio of 10 provides a factor of 3.33 margin to the NRC safety goal for CDF, which is consistent with the approved NuScale methodology.

The NRC staff finds that, as the baseline CDF decreases, the proposed criteria result in a decreasing absolute risk threshold for an SSC to be identified as candidate risk significant. This decreasing absolute risk threshold accounts for uncertainty, which is expected to increase as the baseline risk decreases. The NRC staff finds that this basis adequately addresses uncertainty considerations associated with the proposed component-level RAW criteria.

Holtec provided the following basis for the proposed RAW criteria for identifying system-level basic events as candidate risk significant using CDF:

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a multiplication factor of 7 is used, so a system-level basic event with a RAW greater than 35 is identified as candidate risk significant. The basis for this criterion is that, due to the potentially significant impact of a loss of a system due to CCF, Holtec lowered the multiplication factor from 10 to 7 to reflect the lower CDF.

When the baseline CDF is less than 1x10-7 per year, a multiplication factor of 2 is used, so a system-level basic event with a RAW greater than 60 is identified as candidate risk significant. The basis for this criterion is that, due to the potentially significant impact of a loss of a system due to CCF, Holtec lowered the multiplication factor from 7 to 2 to reflect the lower CDF.

The NRC staff finds that, as the baseline CDF decreases, the proposed criteria result in a decreasing absolute risk threshold for an SSC to be identified as candidate risk significant. This decreasing absolute risk threshold accounts for uncertainty, which is expected to increase as the baseline risk decreases. The NRC staff finds that this basis adequately addresses uncertainty considerations associated with the proposed system-level RAW criteria.

The NRC staff reviewed the guidance in RG 1.200, RG 1.201, RG 1.174, the criteria approved for the ESBWR design, and the criteria approved for the NuScale design. The NRC staff finds that the proposed component-level RAW criteria, based on order of magnitude differences in the baseline CDF, are consistent with the approach in RG 1.174 for acceptable increases in risk based on order of magnitude differences in the baseline risk, provide sufficient margin to the NRC safety goal for CDF to account for PRA uncertainties, and provide additional margin as the baseline risk decreases. The NRC staff finds that the proposed system-level RAW criteria are consistent with the guidance in RG 1.201 and provide additional margin as the baseline risk decreases. Based on its review, the NRC staff finds the proposed thresholds for RAW based on the baseline CDF acceptable for identifying candidate risk-significant SSCs for the SMR-300 design.

9 4.2.2 Large Release Frequency The proposed methodology uses the same approach for LRF as it does for CDF, except the baseline LRF values are reduced by an order of magnitude from the baseline CDF values. This is consistent with the Commissions CCFP goal of less than 0.1 for new reactors and the approach in RG 1.174.

Holtec did not explicitly address uncertainty considerations relative to LRF. However, the proposed methodology uses the same approach for LRF as it does for CDF, except the baseline LRF values are reduced by an order of magnitude from the baseline CDF values. Holtec stated that this is consistent with the Commissions CCFP goal of less than 0.1 for new reactors and the approach in RG 1.174. In addition, Holtec stated that the proposed component-level and system-level RAW criteria meet the intent of RG 1.200 and RG 1.174.

In SRM-SECY-90-016, the Commission approved the overall mean frequency of a large release of radioactive material to the environment from a reactor accident as less than 1x10-6 per year of reactor operation. In SRM-SECY-89-102, SECY-89-102 - Implementation of the Safety Goals (Ref. 15), the Commission described that this frequency is inherently more conservative, but within an order of magnitude of, the quantitative health objectives. As such:

When the baseline LRF is greater than or equal to 1x10-8 and less than 1x10-7 per year, a RAW of 5 provides a factor of 2 margin to the NRC safety goal for LRF.

When the baseline LRF is less than 1x10-8 per year, a RAW of 30 provides a factor of 3.33 margin to the NRC safety goal for LRF.

The NRC staff finds that, as the baseline LRF decreases, the proposed criteria result in a decreasing absolute risk threshold for an SSC to be identified as candidate risk significant. The NRC staff finds that this adequately addresses uncertainty considerations associated with the proposed component-level LRF RAW criteria.

Holtec provided a similar basis for the proposed RAW criteria for identifying system-level basic events as candidate risk significant using LRF as the proposed RAW criteria for identifying system-level basic events as candidate risk significant using CDF. Based on the above discussion, the NRC staff finds that this basis adequately addresses uncertainty considerations associated with the proposed system-level RAW criteria.

The NRC staff reviewed the guidance in RG 1.174 and the Commission direction in SRM-SECY-90-16 and SRM-SECY-12-0081. The NRC staff finds that the proposed LRF criteria are consistent with the Commissions CCFP goal of less than 0.1 for new reactors and the approach in RG 1.174 and provide additional margin as the baseline risk decreases. Based on its review, the NRC staff finds the proposed thresholds for RAW based on the baseline LRF or LERF, as applicable, acceptable for identifying candidate risk-significant SSCs for the SMR-300 design.

4.3 Selection of the Fussell-Vesely Thresholds To supplement the RAW criteria, the proposed methodology uses the FV importance to identify those SSCs that have the largest fractional contribution to risk. Holtec stated that the focus of these criteria is on identifying SSCs for which reliability and availability have the greatest influence on the risk profile.

10 The FV importance enables SSCs to be ranked according to their contribution to risk for each hazard group and mode of plant operation, and it is used to identify SSCs that contribute a significant fraction of the risk from a hazard with very low risk. The basis for the proposed thresholds for FV is summarized in the following sections.

4.3.1 Core Damage Frequency Holtec provided the following basis for the FV criteria for identifying SSCs as candidate risk significant using CDF:

When the baseline CDF is greater than or equal to 1x10-6 per year, a basic event with a FV greater than 0.005 is identified as candidate risk significant. The basis for this criterion is that it is the same as the current FV criterion for CDF in RG 1.200.

When the baseline CDF is greater than or equal to 1x10-7 and less than 1x10-6 per year, a basic event with a FV greater than 0.02 is identified as candidate risk significant. The basis for this criterion is that it maintains the same risk contribution of 2x10-8 per year that is used when the baseline CDF is 1x10-7 per year.

When the baseline CDF is less than 1x10-7 per year, a basic event with a FV greater than 0.2 is identified as candidate risk significant. The basis for this criterion is that the current FV criterion for CDF in RG 1.200 corresponds to a risk contribution of 5x10-8 per year, based on an operating reactor with a nominal CDF of 1x10-5 per year. When the baseline CDF is 1x10-7 per year, (1) a risk contribution of 5x10-8 per year corresponds to a FV of 0.5 and (2) using a relative risk criterion of FV greater than 0.5 does not reflect the intent to use FV for identifying SSCs that contribute a significant portion of the risk because some important contributors could be screened out. Holtec adjusted the relative risk criterion and proposed that a basic event with a FV greater than 0.2 is identified as candidate risk significant.

The NRC staff reviewed the guidance in RG 1.200, the criteria approved for the ESBWR design, and the criteria approved for the NuScale design. The NRC staff finds that the proposed FV thresholds are consistent with the guidance in RG 1.200 and provide additional margin as the baseline risk decreases. The NRC staff finds that uncertainty is addressed by maintaining the same risk contribution of 2x10-8 per year as the baseline risk decreases, which is more conservative than the current FV criterion for CDF in RG 1.200. Based on its review, the NRC staff finds the proposed thresholds for FV based on the baseline CDF acceptable for identifying candidate risk-significant SSCs for the SMR-300 design.

4.3.2 Large Release Frequency The proposed methodology uses the same approach for LRF as it does for CDF, except the baseline LRF values are reduced by an order of magnitude from the baseline CDF values. This is consistent with the Commissions CCFP goal of less than 0.1 for new reactors and the approach in RG 1.174.

The NRC staff reviewed the guidance in RG 1.200 and the Commission direction in SRM-SECY-90-16 and SRM-SECY-12-0081. The NRC staff finds that the proposed FV thresholds are consistent with the guidance in RG 1.200 and provide additional margin as the baseline risk decreases. The NRC staff finds that uncertainty is addressed by maintaining the same absolute

11 risk contribution of 2x10-9 per year as the baseline risk decreases, which is more conservative than the current FV criterion for LERF in RG 1.200. Based on its review, the NRC staff finds the proposed thresholds for FV based on the baseline LRF or LERF, as applicable, acceptable for identifying candidate risk-significant SSCs for the SMR-300 design.

4.4 Application of the Risk Metrics In the TR, Holtec proposed a methodology for identifying candidate risk-significant SSCs using the SMR-300 PRA. The NRC staff notes that important implementation details were not addressed in the TR. For example, the TR did not address (1) the way in which a RAW or FV is assigned to an SSC based on the RAW and FV computed for basic events associated with the failure of the SSC and (2) the specific techniques for assessing risk significance of SSC failures caused by specific hazards such as fires and floods. The NRC staff normally considers such issues in its review of a specific application that involves assessment of risk significance, such as the identification of SSCs to be included in the design reliability assurance program or categorization of SSCs for treatment under the requirements in 10 CFR 50.69. Such applications are submitted after the PRA has been completed and is available for audit by the NRC staff. For this reason, use of the TR in specific risk-informed applications will be reviewed on a case-by-case basis by the NRC staff when those risk-informed applications are submitted for review.

5.0 Staff Conclusions The NRC staff reviewed the proposed methodology and risk significance criteria described in the TR and finds them acceptable for identifying candidate risk-significant SSCs for the SMR-300 design using the SMR-300 PRA. The NRC staffs conclusions for specific technical topics are found within the respective technical evaluation sections of this report. The NRC staff approves the use of the TR, subject to the conditions and limitations in section 6.0, by Holtec in support of licensing applications.

6.0 Conditions and Limitations

1. The NRC staffs approval of this TR is specific to the Holtec SMR-300 design. Any use in whole or in part for other designs would require an additional applicability review by the NRC staff. Use of the TR in specific risk-informed applications will be reviewed on a case-by-case basis by the NRC when those risk-informed applications are submitted for review.

2. The methodology in the TR can only be used in concert with a PRA and analysis of CDF and either LRF or LERF, as applicable, that the NRC staff has determined to be technically acceptable and addresses internal and external hazards and all operating modes, including low-power and shutdown, as required for specific licensing submittals.

The SMR-300 CDF must be less than 1x10-6 per year and the LRF must be less than 1x10-7 per year.

3. The methodology in the TR identifies candidate risk-significant SSCs from the SMR-300 PRA, but it is not the sole determinant of risk significance. To ensure that a holistic risk-informed approach is taken, additional consideration of uncertainties, sensitivities, traditional engineering evaluations and regulations, and maintaining sufficient defense in

12 depth and safety margin will be used to determine a complete list of risk-significant SSCs and will be identified in a future application that references this TR.

7.0 References

1. Brenner, Andrew, SMR, LLC, letter to U.S. Nuclear Regulatory Commission, SMR, LLC Submittal of Holtec PSA Risk Significance Determination Methodology Licensing Topical Report, Revision 1 (Project No. 99902049), October 18, 2024 (Agencywide Documents Access and Management System Accession No. ML24292A046).

2. SMR, LLC, HI-2230875, Revision 1, Holtec PSA Risk Significance Determination Methodology Licensing Topical Report, October 18, 2024 (ML24292A047).

3. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020 (ML20238B871).

4. U.S. Nuclear Regulatory Commission, NUREG-0800, Section 19.0, Revision 3, Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors, December 2015 (ML15089A068).

5. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256).

6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, May 2006 (ML061090627).

7. Nuclear Energy Institute, NEI 00-04, Revision 0, 10 CFR 50.69 Categorization Guideline, July 2005 (ML052910035).

8. U.S. Nuclear Regulatory Commission, NUREG-0800, Section 17.4, Revision 1, Reliability Assurance Program, May 2014 (ML13296A435).

9. U.S. Nuclear Regulatory Commission, NUREG-0800, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis:

General Guidance, June 2007 (ML071700658).

10. NuScale Power, LLC, TR-0515-13952-NP-A, Revision 0, Risk Significance Determination, October 10, 2016 (ML16284A016).

11. U.S. Nuclear Regulatory Commission, NUREG-1966, Volume 4, Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design, April 2014 (ML14100A187).

12. U.S. Nuclear Regulatory Commission, SRM-SECY-90-16, SECY-90 Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationships to Current Regulatory Requirements, June 26, 1990 (ML003707885).

13

13. U.S. Nuclear Regulatory Commission, SRM-SECY-12-0081, Staff Requirements -

SECY-12-0081 - Risk-Informed Regulatory Framework for New Reactors, October 22, 2012 (ML12296A158).

14. U.S. Nuclear Regulatory Commission, NUREG-1150, Volume 1, Severe Accident Risks:

An Assessment for Five U.S. Nuclear Power Plants, December 1990 (ML120960691).

15. U.S. Nuclear Regulatory Commission, SRM-SECY-89-102, SECY-89-102 -

Implementation of the Safety Goals, June 15, 1990 (ML003707881).