ML25345A481
| ML25345A481 | |
| Person / Time | |
|---|---|
| Site: | 99902049 |
| Issue date: | 12/21/2025 |
| From: | NRC/NRR/DNRL/NLIB |
| To: | |
| References | |
| EPID L-2025-TOP-0019, HI-2241098 | |
| Download: ML25345A481 (0) | |
Text
SMR, LLC (A HOLTEC INTERNATIONAL COMPANY) - DRAFT SAFETY EVALUATION OF LICENSING TOPICAL REPORT HI-2241098, SMR-300 DESIGN BASIS ACCIDENT RADIOLOGICAL CONSEQUENCES ANALYSIS METHODOLOGY, REVISION 1 (EPID L-2025-TOP-0019)
SPONSOR AND SUBMITTAL INFORMATION Sponsor:
SMR, LLC, a Holtec International Company Sponsor Address:
Krishna P. Singh Technology Campus 1 Holtec Blvd Camden, NJ 08104 Docket /Project No.:
99902049 Submittal Date:
June 6, 2025 Submittal Agencywide Documents Access and Management System (ADAMS) Accession Nos.: ML25157A142 and ML25248A283 Brief Description of the Topical Report: By letter dated June 6, 2025, SMR, LLC (SMR),
a Holtec International Company (Holtec), submitted Licensing Topical Report (LTR)
HI-2241098, Revision 0, SMR-300 Design Basis Accident Radiological Consequences Analysis Methodology Licensing Topical Report (ML25157A142) to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. The NRC staff accepted the LTR for review on June 26, 2025 (ML25161A170). From August 15, 2025, through August 29, 2025, the NRC staff conducted an audit to gain a detailed understanding of the LTR methodology and identify any additional information that required docketing to support the NRC staffs safety evaluation (SE) for the LTR (ML25181A199 and ML25237A021). By letter dated September 5, 2025, SMR (Holtec) submitted Revision 1 of the LTR (ML25248A283) that addressed items discussed during the NRC staffs audit. The audit report summarizing the NRC staffs observations was issued on September 23, 2025 (ML25258A007).
The LTR describes SMR (Holtec)s proposed methodology for evaluating design basis accident (DBA) radiological consequences. The LTR is specific to the SMR-300 design and generally follows, with a few case-by-case provisions, the methods in Revision 1 to Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (ML23082A305).
REGULATORY EVALUATION The NRC staff considered the following regulations and guidance during its review of the LTR:
Title 10 of the Code of Federal Regulations (10 CFR) Section 50.34, "Contents of applications; technical information," requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.
10 CFR 50.67, "Accident source term," provides an optional provision for licensees to revise their design basis radiological analyses.
Appendix A to 10 CFR Part 50 - General Design Criteria for Nuclear Power Plants, Criterion 19 - Control Room, defines regulatory requirements for radiation exposure limits in the control room during design basis accidents.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms, " and Section 6.5.2, Containment Spray as a Fission Product Cleanup System. The Standard Review Plan provides guidance to the NRC staff in performing safety reviews. Section 15.0.1 provides guidance to the NRC staff in reviewing radiological consequence analysis that use alternative source terms (AST). Section 6.5.2 provides guidance to the NRC staff when reviewing containment spray as a fission product cleanup system. For this review, the staff used Section 6.5.2 to evaluate the proposed method for determining the elemental iodine removal coefficient for natural wall deposition inside containment.
RG 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, describes approaches that are acceptable to the NRC staff for complying with regulations for DBA dose consequence analysis using AST.
RG 1.23, Revision 1, Meteorological Monitoring Programs for Nuclear Power Plants, provides guidance on establishing meteorological monitoring programs at nuclear power plants to support safety analyses and emergency planning. It specifies requirements for accurate data collection of wind, temperature, and other parameters essential for atmospheric dispersion modeling and radiological impact assessments.
RG 1.249, Revision 0, Use of ARCON Methodology for Calculation of Accident-Related Offsite Atmospheric Dispersion Factors, describes an approach acceptable to the NRC staff for the use of the ARCON computer code to calculate offsite atmospheric dispersion values.
RG 1.194, Revision 0, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, outlines methods acceptable to the NRC staff for calculating accident-related onsite atmospheric dispersion values including the use of the ARCON computer code.
TECHNICAL EVALUATION In the LTR, SMR (Holtec) provides a site-independent methodology for determining the accident consequences from DBAs that are required to be evaluated in accordance with 10 CFR 50.34, Contents of applications; technical information, and for demonstrating compliance with control room acceptance criteria reflected in Appendix A to 10 CFR Part 50, Criterion 19. This methodology adheres to NRC regulatory requirements and guidance, and in those instances where the methodology applies case-by-case approaches, SMR (Holtec) provides adequate technical justifications. The methodology is used to determine the total effective dose equivalent (TEDE) at the exclusion area boundary (EAB); the low population zone (LPZ); the main control room (MCR); and the technical support center (TSC) resulting from postulated design basis accidents. During the application stage for a new facility, these values would be compared to the acceptance criteria provided in Appendix A to 10 CFR Part 50, Criterion 19 and RG 1.183, Revision 1, as applicable.
SMR (Holtecs) methodology is summarized in Section 4.0, Methodology, of the LTR and the use of computer codes is described in Section 3.0, Code Descriptions and Analyses Flowchart. SMR (Holtec) proposes to follow a traditional approach to consequence analysis in that a radionuclide inventory is developed using a depletion software code from which a containment source term is developed using appropriate depletion calculations and release fractions. The radionuclide transport is then evaluated using a software code that incorporates applicable source term reductions, for example, from natural deposition and filtering from components such as control room ventilation systems. Finally, the radiological dose consequences are determined at the EAB, LPZ, and in the MCR and TSC through application of dose conversion factors.
During the NRC staffs review of Revision 0 of the LTR, the NRC staff found that a regulatory audit would be beneficial to obtain clarification on SMR (Holtecs) use of the pH in the containment flood-up pool for determining if iodine re-evolution from the containment flood-up pool should be included in the containment source term. Following completion of the audit, SMR (Holtec) submitted Revision 1 of the LTR to correct and clarify Section 4.3.1.2, Iodine Chemical Form, and the NRC staff determined that it is not necessary for the LTR to evaluate iodine re-evolution from the containment flood-up pool. This SE applies to Revision 1 of the LTR.
The following sections of this SE describe the NRC staffs technical evaluation of the methodology proposed in the LTR. The section numbers below correspond to the section numbering scheme in Section 4.0 of the LTR.
4.1 Accident Release Durations and Dose Acceptance Criteria The LTR proposes to use analysis release durations and dose acceptance criteria that are consistent with Table 7 of RG 1.183, Revision 1. This assumption is acceptable because it is consistent with the NRC-approved methodology.
4.2 Radionuclide Inventory In the SMR-300 design the spent fuel pool (SFP) is located inside the containment structure.
However, for DBA consequence analysis, the radionuclide inventory of the SFP (i.e., spent fuel and water in SFP) is not included in the development of the source term used to evaluate the radiological consequences of design basis accidents. This assumption is acceptable because the radionuclide inventory of the SFP is not liberated by design basis accidents analyzed in the LTR. In the case of the fuel handling accident (FHA) the source term from the damaged fuel assembly is first determined and the release from the SFP is calculated based on the transport of the source term through the SFP column of water. For the FHA, specific assumptions account for radionuclides released to the SFP and to the environment as described in Section 4.3.5 of this SE.
4.2.1 Core Radionuclide Inventory The LTR proposes to develop the radionuclide inventory consistent with the guidance in Regulatory Position 3 of RG 1.183, Revision 1. Specifically, the core radionuclide inventory is calculated using the SCALE depletion software code and the approach for distinguishing between the inventory that applies to the maximum hypothetical accident (MHA) and that which applies to DBAs that do not involve the entire core is consistent with Regulatory Position 3.1 of RG 1.183, Revision 1. Therefore, this approach is acceptable.
The radial peaking factor is a correction factor that accounts for differences in power level across the radial axis of the core. It is used to modify consequence analyses for DBAs that only impact parts of the core to ensure results are conservative. The LTR proposes three at-instant, core inventories that apply radial peaking factors in a situationally-dependent manner. First, the radial peaking factor is applied to analyses that do not involve the entire core. Second, the radial peaking factor is not applied to analyses that assume full-core damage. The third at-instant, core inventory applies to the FHA and it includes a radial peaking factor and a conservative cooling time that is specific to the FHA. This approach is acceptable because it is conservative and bounding of specified accidents.
4.2.2 Coolant Radionuclide Inventory In Sections 4.2.2.1 and 4.2.2.2, the LTR describes considerations for determining the radionuclide inventory in the primary and secondary coolant, respectively. To the extent that the coolant radionuclide inventory serves as an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, the assumed value requires a plant-specific technical specification under Criterion 2 of 10 CFR 50.36(c)(2(ii). Detailed evaluation of this aspect of the LTR is beyond the scope of this SE. However, the considerations described in Sections 4.2.2.1 and 4.2.2.2 are reasonable approaches to determining the coolant radionuclide inventory because they include bounding factors to account for the presence of radionuclides in the primary and secondary coolant such as fuel defects and leakage from steam generator tubes.
4.3 Source Terms 4.3.1 MHA LOCA [Loss of Coolant Accident] Source Terms In Section 4.3.1.1, the LTR proposes to use release fractions from Table 2 of RG 1.183, Revision 1 and the release timings from Table 5 of RG 1.183, Revision 1.
In Section 4.3.1.2, the LTR proposes to use the chemical form that applies to the iodine that is assumed to be released to the containment atmosphere from Regulatory Position 3.5 of RG 1.183, Revision 1. These fractions are 95% cesium iodide, 4.85% elemental iodine, and 0.15%
organic iodide.
Regulatory Position A-1.1 in RG 1.183 states, If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of radioiodine released to the containment should be assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, Section 3.5, Chemical Form, provides a basis for the pH value assumed in RG 1.183, Revision 0. This basis and the pH value was unchanged in RG 1.183, Revision 1.
As stated in NUREG/CR-5732, Section 3.3.2, which provides detailed background information for the staff position on pH found in NUREG-1465, a pool pH of greater than 7 would result in an elemental iodine re-evolution of less than 1%, a negligible amount. Using mathematical models from NUREG-5950, which are substantially similar to the approaches described in NUREG/CR-5732, SMR (Holtec) demonstrated that with conservative assumptions for pool concentrations, temperature, and a pH of 6, that the percentage of re-evolved elemental iodine remains below 1%. The NRC staff confirmed this through independent calculations.
Since the elemental iodine that would be expected to re-evolve in the SMR-300 containment flood-up pool is bounded by the existing assumption of 1%, use of the assumed elemental iodine value of 4.85% from Regulatory Position A-1.1 in RG 1.183, Revision 1, as described in Section 4.3.1.2 of the LTR, is acceptable. Therefore, the assumptions used to determine the MHA LOCA source term for the SMR-300 are acceptable because they are consistent with the NRC-approved methodology.
4.3.2 REA [Rod Ejection Accident] Source Terms for Containment Release Path In Sections 4.3.2.1 and 4.3.2.2, the LTR proposes to evaluate the REA (containment release path) consistent with guidance in Appendix H of RG 1.183, Revision 1. Specifically, release fractions are consistent with Regulatory Position H-1, and the assumed instantaneous timing of the release is consistent with Regulatory Position H-3. Additionally, the chemical form of iodine released to the containment environment is consistent with Regulatory Position H-4.
The application of the iodine chemical form fractions from Regulatory Position H-4 is acceptable on the same basis as described in Section 4.3.1 of this SE. These assumptions are acceptable because they are consistent with the NRC-approved methodology.
4.3.3 REA Source Terms for Secondary Plant Release Path In Sections 4.3.3.1 and 4.3.3.2, the LTR proposes to evaluate the REA (secondary release path) consistent with guidance in Appendix H of RG 1.183, Revision 1. Specifically, release fractions are consistent with Regulatory Position H-1, and the assumed instantaneous timing of the release is consistent with Regulatory Position H-3. The chemical form of the iodine releases from the steam generator is consistent with Regulatory Position H-5 of RG 1.183, Revision 1.
These assumptions are acceptable because they are consistent with the NRC-approved methodology.
4.3.4 Locked Rotor Accident Source Terms In Sections 4.3.4.1 and 4.3.4.2, the LTR proposes to evaluate the reactor coolant pump locked rotor accident consistent with guidance in Appendix G of RG 1.183, Revision 1. Fuel melt resulting from a locked rotor accident is outside of the design basis of the SMR-300; therefore, the LTR only evaluates gap release resulting from cladding failure that may result from a locked rotor event. Consequently, pressurized water reactor steady-state fission product release fractions residing in the fuel rod plenum and gap from Table 4 of RG 1.183, Revision 1 are assumed to be released. Consistent with Regulatory Position G-3, the release is assumed to occur instantaneously. The chemical form of the iodine releases from the steam generator is consistent with Regulatory Position G-4 of RG 1.183, Revision 1. These assumptions are acceptable because they are consistent with the NRC-approved methodology.
4.3.5 FHA Source Terms Consistent with Regulatory Position B-1.1 of RG 1.183, Revision 1, SMR (Holtec) performed a conservative analysis to determine the number of fuel rods that are assumed damaged during the accident in the most limiting case. The LTR states that the SMR-300 fuel handling crane is single failure-proof and designed such that a fuel assembly cannot be inadvertently released by the crane. This ensures that an assembly cannot be dropped onto another assembly.
Additionally, the SMR-300 design includes measures to ensure that fuel assemblies do not damage adjacent assemblies when being lifted. A detailed evaluation of the overhead heavy load handling systems is not within the scope of this SE. However, for the purposes of the FHA analysis, the description in the LTR is sufficient to support the assumption that only a single fuel assembly would be damaged during the FHA.
In Sections 4.3.5.1 and 4.3.5.2, the LTR proposes to evaluate the FHA consistent with guidance in Appendix B of RG 1.183, Revision 1. Specifically, release fractions and the assumed instantaneous timing of the release are consistent with Regulatory Position B-1.2. The chemical form of the iodine releases from the steam generator are consistent with Regulatory Position B-1.3 of RG 1.183, Revision 1. These assumptions are acceptable because they are consistent with the NRC-approved methodology.
4.3.6 Primary Coolant Source Terms for DBAs with No Fuel Breach For accidents that do not predict fuel failure, the source term for follow-on radiological consequence analysis is determined by the maximum allowable reactor coolant system activity per licensee technical specifications. For the steam generator tube rupture and main steam line break accidents, Regulatory Positions E-2 and F-2 of RG 1.183, Revision 1 apply, respectively.
Additionally, in accordance with Regulatory Positions E-2 and F-2 of RG 1.183, Revision 1, the LTR evaluates two cases of iodine spiking: the pre-accident spike and the concurrent spike.
These assumptions are acceptable because they are consistent with the NRC-approved methodology.
In Section 4.3.6.1, the LTR proposes to evaluate the pre-accident spike source term for the steam generator tube rupture and main steam line break accidents consistent with Regulatory Positions E-2.1 and F-2.1 of RG 1.183, Revision 1. Additionally, the LTR proposes to use the dose conversion factors for converting between isotopic concentrations and Dose Equivalent I-131 using Table 2.1 of Federal Guidance Report (FGR) 11. FGR-11 is one of the approved sources for dose conversion factors that apply to the definition of Dose Equivalent I-131 per the Standard Technical Specifications. These assumptions are acceptable because they are consistent with NRC-approved methodologies.
To determine the source term that applies during the concurrent iodine spike for the steam generator tube rupture and main steam line break, Section 4.3.6.2 of the LTR first calculates an iodine appearance rate. The iodine appearance rate is equivalent to the iodine concentration at equilibrium, which, per Criterion 2 of 10 CFR 50.36(c)(2)(ii), would be provided in technical specifications. The LTR proposes to use Equation 4-1, which yields an equilibrium iodine concentration by balancing iodine production and removal rates as described in the NRC memorandum titled Results of Initial Screening of Generic Issue 197, Iodine Spiking Phenomena, dated May 8, 2006 (ML061100331). The equilibrium iodine concentration is then multiplied by a spiking factor and spike duration consistent with Regulatory Positions E-2.2 and F-2.2 of RG 1.183, Revision 1, for the steam generator tube rupture and main steam line break accidents, respectively. While Regulatory Position E-2.2 prescribes a spiking factor of 335 for the steam generator tube rupture, the LTR conservatively applies a spiking factor of 500 for both the steam generator tube rupture and main steam line break accidents. These assumptions are acceptable because they are consistent with or more conservative relative to the NRC-approved methodologies.
In Sections 4.3.6.3 and 4.3.6.4, the LTR proposes to evaluate the steam generator tube rupture and main steam line break source terms consistent with guidance in Appendices E and F of RG 1.183, Revision 1. Specifically, release fractions and the assumed instantaneous timing of the release are consistent with Regulatory Positions E-3 and F-3 of RG 1.183, Revision 1, for the steam generator tube rupture and main steam line break accidents, respectively.
The chemical form of the iodine releases from the steam generator is consistent with Regulatory Positions E-5 and F-5 of RG 1.183, Revision 1, for the steam generator tube rupture and main steam line break accidents, respectively. These assumptions are acceptable because they are consistent with the NRC-approved methodology.
4.3.7 Secondary Coolant Source Terms In Sections 4.3.7.1 and 4.3.7.2, the LTR describes how radiological source terms in the secondary water are determined. The secondary coolant source term is determined as described in Section 4.2.2 of this SE and is conservatively assumed to be released to the environment via steaming without any mitigation. The chemical form of the iodine releases from the steam generator are consistent with Regulatory Positions E-5 and F-5 of RG 1.183, Revision 1, for the steam generator tube rupture and main steam line break accidents, respectively. These assumptions are acceptable because they are conservative and consistent with the NRC-approved methodologies.
4.4 Radionuclide Transport The following sections describe aspects of radionuclide transport that are specific to individual accident analyses.
4.4.1 Spent Fuel Pool Decontamination For the FHA analysis, the LTR evaluates the decontamination factor for radionuclides released into the spent fuel pool that originate from the damaged fuel rods. In this case, the column of water above the damaged fuel serves as a filter medium depending on the chemical form of the radionuclides. Consistent with Regulatory Position B-4 of RG 1.183, Revision 1, noble gases are assumed to escape the pool fully and radionuclides in particulate form are assumed to be fully retained in the spent fuel water. Isotopes in elemental form are assumed to be released instantaneously, and their decontamination factor is determined consistent with Regulatory Position B-2 of RG 1.183, Revision 1. Equation 4-5 of the LTR converts the decontamination factor to a filter efficiency using a generally accepted equation. These assumptions are acceptable because they are consistent with the NRC-approved methodologies.
4.4.2 Natural Aerosol Deposition In Section 4.4.2, the LTR proposes to credit natural aerosol deposition consistent with Regulatory Position A-2.2 of RG 1.183, Revision 1. The continued applicability of the Powers Deposition Model was evaluated by the NRC staff and was the subject of a public meeting on September 24, 2025 (public meeting notice at ML25262A164; meeting presentation materials at ML25261A154). This assumption is acceptable because it is consistent with the NRC-approved methodologies.
4.4.3 Elemental Iodine Removal In Section 4.4.3, the LTR proposes to assume iodine removal via iodine gas interaction with free surfaces inside containment in radiological analyses for all accidents with containment path releases. This approach is consistent with Regulatory Position A-2.2 of RG 1.183, Revision 1.
The elemental iodine removal by wall deposition is determined consistent with the method described in Subsection III.4.C of Section 6.5.2 of NUREG-0800, and the LTR proposes to use the value for mass-transfer which ensures that all experimental data used to develop Equation 4-6 of the LTR are bounded. This approach is acceptable because it is consistent with the NRC-approved methodologies.
4.4.4 Containment Leakage In Section 4.4.4, the LTR proposes assumptions relevant to containment leakage that are consistent with Regulatory Position A-2.7 of RG 1.183, Revision 1. Specifically, radionuclides available for release are assumed to leak to the atmosphere for all accidents with containment path releases at the peak pressure technical specification (TS) leakage rate. The leakage is assumed to reduce to 50% of the TS leakage rate after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For the FHA, the containment air lock is assumed to be open. Therefore, radioisotopes released to the containment are assumed to be released to the environment instantaneously.
These assumptions are acceptable because they are consistent with the NRC-approved methodologies.
4.4.5 Secondary Plant Release Path Accident Sources In Section 4.4.5, the LTR provides considerations for transport analyses involving secondary plant release paths. These analyses are consistent with the source term considerations in Sections 4.3.6 and 4.3.7 of this SE. Additionally, consistent with Regulatory Positions E-3 and E-6, the LTR conservatively assumes that for the steam generator tube rupture accident, primary-to-secondary normal leakage and break flow due to the rupture in the SMR-300 once-through steam generator occur in the steam space and completely flash vapor. These assumptions are acceptable because they are consistent with the NRC-approved methodologies.
4.4.6 Atmospheric Dispersion The NRC staff reviewed SMR (Holtecs) description of the atmospheric dispersion analysis.
SMR (Holtec) states in Section 4.4.6 of the LTR that in the event of an accident, radionuclides may be released to the environment from containment leakage or secondary plant release paths to form a plume that could travel to receptor sites at the EAB, at the LPZ outer boundary, in the MCR, and in the TSC. To quantify this dispersion and assess the potential exposure for a receptor at various locations, SMR (Holtec) will conduct a site-specific atmospheric dispersion analysis. SMR (Holtec) cites Regulatory Position 5.3 of RG 1.183 which lists the application of RG 1.194 and RG 1.249 as guidance for calculating atmospheric dispersion factors (/Q).
Guidance in RG 1.194 is typically followed for MCR and TSC dispersion analysis. If source-to-receptor distance is less than 1200 meters (m), procedures consistent with RG 1.249 can be used to support application of the modeling for EAB and LPZ /Q calculations. SMR (Holtec) states that the NRC-sponsored computer code ARCON 2.0 is used for the analysis and is endorsed by RG 1.249. As discussed in RG 1.249, ARCON 2.0 uses the same underlying dispersion algorithms as ARCON96, which is endorsed by RG 1.194. Therefore, ARCON 2.0 and ARCON96 are equally applicable to the calculations. The NRC staff finds SMR (Holtecs) description of the atmospheric dispersion analysis consistent with NRC guidance and therefore acceptable.
4.4.6.1 Overview Figure 4.2 of the LTR provides a flowchart of the proposed atmospheric dispersion analysis.
The flowchart shows that SMR (Holtec) plans to use the meteorological data, source-receptor distances and directions, and other release characteristics to make ARCON 2.0 model runs producing the atmospheric dispersion factors. The NRC staff reviewed this section of the LTR, and since this methodology corresponds with NRC guidance, the NRC staff finds it acceptable.
4.4.6.2 Meteorological Data The NRC staff reviewed SMR (Holtecs) proposed plan for the meteorological data. SMR (Holtec) stated in Section 4.4.6.2 of the LTR that the methodology described in the LTR is site-independent and therefore, when applied to a particular site, hourly meteorological data must be collected using a meteorological monitoring program compliant with RG 1.23. SMR (Holtec) states that the data is then processed into the required input format for the atmospheric dispersion analysis code, ARCON 2.0. Section 4.4.6.2 of the LTR lists the inputs required by ARCON 2.0 as discussed in Section 4.4.2 of NUREG/CR-6331, Atmospheric Relative Concentrations in Building Wakes. SMR (Holtec) also states that the wind stability is determined based on temperature differences captured by instruments at the lower and higher levels and provides Table 4.4 to show the classification of atmospheric stability that will be used in the analysis. The NRC staff reviewed SMR (Holtecs) proposed plan for the meteorological data and found it acceptable since it will be compliant with the referenced NRC guidance documents.
4.4.6.3 Statistical Confidence Level The NRC staff reviewed SMR (Holtecs) proposed plan to ensure statistical confidence. SMR (Holtec) stated in Section 4.4.6.3 of the LTR that for evaluation of the EAB and LPZ receptor locations, the more conservative of the 95th percentile overall or 99.5th percentile sector-wise
/Q is used to ensure the assessment is bounding. This approach is consistent with guidance in RG 1.249. The LTR states that the 95th percentile overall /Q is calculated using the distance from the nuclear island building to the nearest point on the site boundary, and for a multi-unit site, the distances from any nuclear island building to the nearest point on the site boundary are considered and the minimum distance is used. SMR (Holtec) points to Subsection 4.4.6.4 of the LTR for a description of how the source-to-receptor distances for the 99.5th percentile are determined. For evaluation of the MCR and TSC receptor locations, SMR (Holtec) states that RG 1.194 is followed by selecting the 95th percentile /Q. SMR (Holtec) points to Subsection 4.4.6.5 of the LTR for a description of how the source-to-receptor distances and directions are determined. The NRC staff finds process of selecting the 95th percentile overall or 99.5th percentile sector-wise /Q acceptable as it is consistent with NRC guidance.
4.4.6.4 Source-to-Receptor Distances and Directions for EAB and LPZ The NRC staff reviewed SMR (Holtecs) source-to-receptor distance and direction calculation methodology for the EAB and LPZ. SMR (Holtec) notes in Section 4.4.6.4 of the LTR that ARCON 2.0 is applicable for atmospheric dispersion analysis when source-to-receptor distance falls between 10 m and 1200 m, as noted in RG 1.194 and RG 1.249, respectively. SMR (Holtec) states that the source-to-receptor distance and direction calculation methodology for the EAB and LPZ outer boundary follows the guidance in Regulatory Position 2.2.1 of RG 1.249.
SMR (Holtec) states that for the SMR-300 plant design, the EAB and LPZ outer boundary will coincide with the site boundary, and the distance calculation will follow one of the two approaches outlined in RG 1.249 and reflected by SMR (Holtec) in the LTR. SMR (Holtec) lists the details of the first approach where the limiting distance from the nuclear island building to the nearest point on the site boundary is used for atmospheric dispersion analysis. SMR (Holtec) illustrated the approach in LTR Figure 4-3. SMR (Holtec) lists the details of the second approach where the limiting distance between the nuclear island building and the nearest point on the site boundary is evaluated in each of the 16 directional sectors (22.5° each). SMR (Holtec) illustrated that approach in LTR Figure 4-4. The NRC staff finds SMR (Holtecs) source-to-receptor distance and direction calculation methodology for the EAB and LPZ outer boundary to be acceptable as it is consistent with NRC guidance.
4.4.6.5 Source-to-Receptor Distances and Directions for MCR and TSC The NRC staff reviewed SMR (Holtecs) source-to-receptor distance and direction calculation methodology for the MCR and TSC. SMR (Holtec) states in Section 4.4.6.5 of the LTR that the methodology follows the guidance in Regulatory Position 3.4 of RG 1.194, and that all release points and air intakes are identified and the receptor is considered to be the point of air intake. SMR (Holtec) notes that the source-to-receptor distance is the shortest horizontal distance between the release point and the intake, and the direction is the wind direction that would carry the plume from the release point to the intake. For each combination of release points and intakes, SMR (Holtec) states that the minimum source-to-receptor distance and the direction are determined. SMR (Holtec) states that the largest /Q for each averaging time period is then identified among all release points and intake combinations. The NRC staff finds SMR (Holtecs) source-to-receptor distance and direction calculation methodology for the MCR and TSC to be acceptable as it is consistent with NRC guidance.
4.4.6.6 Release Characteristics The NRC staff reviewed the release characteristics that SMR (Holtec) will use as inputs for ARCON 2.0 model runs for the EAB, LPZ, MCR, and TSC atmospheric dispersion analysis.
SMR (Holtec) lists the release characteristics in Subsection 4.4.6.6 of the LTR. The NRC staff finds the characteristics that SMR (Holtec) will use for the dispersion modeling to be acceptable as they are consistent with NRC guidance.
4.4.6.7 Ground Deposition During Atmospheric Transport Section 4.4.6.7 of the LTR states that consistent with Regulatory Position 4.1 of RG 1.183, depletion of radioactive materials in the plume due to ground deposition is not credited.
The NRC staff finds this to be acceptable because it is consistent with NRC guidance.
The NRC staff reviewed the atmospheric dispersion section of the LTR and found that the proposed methodology is based on information provided in the applicable NRC guidance referenced in the LTR and is consistent with the methodologies outlined in those documents.
Therefore, the NRC staff determined that the proposed atmospheric dispersion analysis methodology provides acceptable methods for collecting and processing meteorological data and calculating the atmospheric dispersion factors at the MCR, TSC, EAB, and LPZ outer boundary which will be used in the radiation consequences analysis.
4.4.7 MCR Habitability Section 4.4.8 of the LTR states that for the MHA LOCA analysis, radionuclide transport through the systems that support MCR habitability is credited. As stated in Section 4.5.2 of the LTR, these systems are not credited for any of the other DBAs. These systems consist of a non-safety-related non-radiologically controlled heating, ventilation, and air conditioning (NRV) system and a non-safety-related breathing air and pressurization (BAP) system. Detailed evaluation of these systems, including actuation time, redundancy, and seismic qualification, if applicable, is outside of the scope of the LTR and this SE. The NRV isolates normal air intake and routes air through a filtration unit when radiation levels in the outside air intake exceed a predefined threshold, as detected by a radiation monitor. The BAP provides an emergency source of pressurized air and means for recirculation through a filter unit to maintain a habitable environment for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the NRV is unavailable. After the 72-hour emergency operation period, the LTR assumes that the NRV can be reactivated to maintain habitable control room conditions. To the extent that the NRV and the BAP systems function as part of the primary success path and function or actuate to mitigate the MHA LOCA, their actuation and function require plant-specific technical specifications under Criterion 3 of 10 CFR 50.36(c)(2(ii).
Therefore, the evaluation of the MCR habitability analysis is plant specific.
4.4.8 TSC Habitability Section 4.4.8 of the LTR states that for the MHA LOCA analysis, radionuclide transport through the systems that support TSC habitability is credited. As stated in Section 4.5.2 of the LTR, these systems are not credited for any of the other DBAs. The non-safety-related normal TSC ventilation system automatically isolates normal air intake, routes air through a filtration unit and activates recirculation when radiation levels in the outside air intake exceed a predefined threshold, as detected by a radiation monitor. Detailed evaluation of these systems, including actuation time, redundancy, and seismic qualification if applicable, is outside of the scope of the LTR. Regulatory Position 4.4 of RG 1.183, Revision 1, describes the acceptance criteria for design basis accidents. For new reactor applicants, the TSC habitability acceptance criterion is based on the requirements to provide an onsite TSC from which effective direction can be given and effective control can be exercised during an emergency, as discussed in paragraph IV.E.8 of Appendix E to 10 CFR Part 50. The radiation protection design of the TSC is acceptable if the total calculated radiological consequences for the postulated fission product release fall within the exposure acceptance criteria specified for the control room. Therefore, the evaluation of the TSC habitability analysis is plant specific.
4.5 Dose Consequences Sections 4.5.1 and 4.5.2 of the LTR provide descriptions of the calculation methodologies used for determining dose consequences at the EAB, LPZ, MCR, and TSC.
4.5.1 Dose Consequences for EAB and LPZ Section 4.5.1 of the LTR proposes to evaluate the dose consequences at the EAB and LPZ consistent with Regulatory Positions 4.1.e and 4.1.f of RG 1.183, Revision 1, respectively.
In Section 4.5.1.1, the LTR proposes breathing rates that are consistent with Regulatory Position 4.1 of RG 1.183, Revision 1. In Section 4.5.1.2, the LTR describes the calculation that occurs in RADTRAD, which is an approved software code for dose consequence analysis.
Dispersion factors are described in Section 4.4.6 of this SE. These assumptions are acceptable because they are consistent with the NRC-approved methodologies.
4.5.2 Dose Consequences for MCR and TSC Section 4.5.2 of the LTR proposes to evaluate the dose consequences at the MCR and the TSC consistent with Regulatory Position 4.2.1 of RG 1.183, Revision 1. Specifically, the TEDE analysis should consider all sources of radiation that will cause exposure of control room personnel. The list of sources the LTR proposes to evaluate is consistent with Paragraphs a through e of Regulatory Positions 4.2.1 of RG 1.183, Revision 1.
Sections 4.5.2.1.1, 4.5.2.1.2, and 4.5.2.1.3 of the LTR describe assumptions relevant to the dose receptor that are consistent with Regulatory Position 4.2.6 of RG 1.183, Revision 1.
Specifically, the breathing rates, occupancy factors, and omission of credit for dose mitigation equipment and prophylaxis are consistent with NRC-approved methodologies.
Sections 4.5.2.1.5 and 4.5.2.2 of the LTR describe the calculations used to determine immersion, inhalation, and shine dose at the MCR and TSC. In the case of immersion and inhalation doses, RADTRAD is used to determine the doses. In the case of shine dose, the RADTRAD output is used to determine the radioactive concentration in various compartments, and these values are used as inputs to further analyses to determine radiation shine dose.
These assumptions and methods are acceptable because they are consistent with the NRC-approved methodologies.
LIMITATIONS AND CONDITIONS The NRC staff imposes the following limitations and conditions with regard to the use and approval of the subject LTR:
1.
The NRC staffs approval of this LTR is specific to the SMR (Holtecs) SMR-300 design as described in the LTR. 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.
Certain aspects of the analyses described in the LTR may depend upon the establishment of plant-specific technical specifications. An NRC issuance of a plant-specific license whose application depends on this LTR would require that technical specifications be established to support evaluations that meet the criteria in 10 CFR 50.36(c)(ii) for limiting conditions for operation.
CONCLUSION The NRC staff reviewed the methodology for evaluating the radiological consequences of design basis accidents as proposed in the Licensing Topical Report HI-2241098, Revision 0, SMR-300 Design Basis Accident Radiological Consequences Analysis Methodology Licensing Topical Report, and finds it acceptable and consistent with current regulations. 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 this LTR, subject to the conditions and limitations discussed above, by SMR (Holtec) in support of licensing applications.
REFERENCES 1.
E.C. Beahm, C.F. Weber, T.S. Kress, and G. W. Parker, "Iodine Chemical Forms in LWR Severe Accidents," NUREG/CR-5732 (ORNLTM-11861), prepared for NRC by Oak Ridge National Laboratory, April 1992, (ML003726825).
2.
NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Revision 5, Volume 1, Specifications.
3.
RG 1.23, Revision 1, Meteorological Monitoring Programs for Nuclear Power Plants."
4.
NUREG/CR-6331, Atmospheric Relative Concentrations in Building Wakes.
Principal Contributors: David Garmon, NRR John Parillo, NRR Jason White, NRR Victoria Huckabay, NRR
ML25345A481 via Econcurrence NRR-106 OFFICE NRR/DNRL/NLIB:PM NRR/DANU/UAL1: LA NRR/DRA/ARCB NAME VHuckabay DGreene DGarmon DATE 12/11/2025 12/18/2025 12/15/2025 OFFICE NRR/DRA/ARCB: BC NRR/DNRL/NLIB: BC NAME SMeighan MHayes DATE 12/17/2025