ML25303A309

Chapter 8 - Us NRC Draft Safety Evaluation Related to the U.S. Sfr Owner, LLC Construction Permit Application for the Kemmerer Power Station Unit 1

ML25303A309
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	11/06/2025
From:	NRC/NRR/DANU
To:
References
Download: ML25303A309 (1)

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THIS NRC STAFF DRAFT SE HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ACRS. THIS DRAFT SE HAS NOT BEEN SUBJECT TO FULL NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS.

8 PLANT PROGRAMS This chapter of the safety evaluation (SE) describes the staffs review and evaluation of the Kemmerer unit 1 (KU1) PSAR chapter 8, which contains a preliminary description of plant programs, quality assurance, and fire protection programs.

The applicable regulatory requirements for the evaluation of the plant programs are as follows:

Title 10 of the Code of Federal Regulations (CFR) 50.34(a)(3)(i) 10 CFR 50.34(a)(7) 10 CFR 50.34(b)(6)(iv) 10 CFR 50.34(f)(2)(iii) 10 CFR 50.34(f)(2)(xxvi) 10 CFR 50.35 10 CFR 50.40 10 CFR 50.43(e)(1) 10 CFR 50.48(a) 10 CFR 50.49 10 CFR 50.55a 10 CFR 50.55(f) 10 CFR 50.65(a) 10 CFR 50.65(b) 10 CFR 50.69 10 CFR Part 50, Appendix A 10 CFR Part 50, Appendix B The applicable guidance for the evaluation of the plant programs are as follows RG 1.20, Revision 4, Comprehensive Vibration Assessment Program for Reactor internals During Preoperational and Startup Testing (ML16056A338)

RG 1.89, Revision 2, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants. (ML22325A263)

RG 1.100, Revision 4, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants. (ML19312C533)

RG 1.189, Revision 5, Fire Protection for Nuclear Power Plants. (ML23214A263)

RG 1.209, Revision 0, Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants.

(ML20166A001)

RG 1.246, Revision 0, Acceptability of ASME Code,Section XI, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants, for Non-Light Water Reactors. (ML22111A087)

RG 1.253, Guidance for a Technology-Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Advanced Reactors, (ML23269A222)

DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor ApplicationsRoadmap (ML23277A139)

8-2 DANU-ISG-2022-05, Advanced Reactor Content of Application Project Chapter 11, Organization and Human-System Considerations Interim Staff Guidance, March 2024 (ML23277A143)

DANU-ISG-2022-07, Risk-informed Inservice Inspection/Inservice Testing Programs for Non-LWRs (ML23277A231)

Definitions of PDC are included in PSAR Section 5.3. The following PDC were identified as applicable to the plant programs as follows:

PDC 1, Quality Standards and Records PDC 3, Fire Protection PDC 19, Control Room PDC 30, Quality of Primary Coolant Boundary PDC 32, Inspection of Primary Coolant Boundary PDC 36 Inspection of Emergency Core Cooling System PDC 37, Testing of Emergency Core Cooling System PDC 45, Inspection of Structural and Equipment Cooling System PDC 46, Testing of Structural and Equipment Cooling System PDC 73, Sodium Leakage Detection and Reaction Prevention and Mitigation PDC 77, Inspection of Intermediate Coolant Boundary 8.1 Plant Programs PSAR Section 8.0 describes plant programs initially considered as special treatments for safety-significant structures, systems, and components (SSCs). Programs that apply to safety-related (SR) or non-safety-related with special treatment (NSRST) SSCs are governed under the quality assurance program description (QAPD) that is evaluated in section 8.2 of this SER. The special treatment programs that are initiated to support the design and construction phase are referenced in table 8.1-1 of this SE. Additional special treatment programs that are associated with the operations phase are also located in table 8.1-1.

USO requested an exemption to the maintenance rule as documented in Enclosure 4, Regulatory Exemptions, of its application for the Kemmerer 1 facility. The exemption request proposes to limit the maintenance rule scope to safety-related and non-safety-related with special treatment SSCs to align with the licensing basis of the Natrium reactor. This exemption is evaluated in Appendix [X] of this SE.

8.1.1 Technical Evaluation As discussed in RG 1.253, PSAR chapter 8.0 should provide an overview of the plant programs relied upon to support the LMP-based safety analysis, addressing these programs purpose, scope, and performance objectives, as well as applicability to SR SSCs, NSRST SSCs, and operations activities. In addition, RG 1.253 states that at the CP stage applicants should include general descriptions in the SAR regarding any programs needed to implement special treatments and meet reliability and performance targets for SR SSCs and NSRST SSCs.

The staffs evaluation of the human factors engineering program and associated technical information for the facility is addressed in section 11.2. The staffs evaluation of the emergency preparedness program and associated technical information for the facility is addressed under section 11.3. The staffs evaluation of the post-construction inspection, testing, and analysis program and associated technical information for the facility is addressed in chapter 12.

8-3 8.1.1.1 Comprehensive Vibration Assessment Program With respect to the Comprehensive Vibration Assessment Program (CVAP), table 8.0-1 in PSAR chapter 8 references RG 1.20 and specifies that the CVAP is required for unique or first-of-its-kind designs. The applicant states that the results of the CVAP will be documented in a series of technical reports submitted to the NRC. The applicant will use these reports to demonstrate compliance with regulatory requirements and to ensure safe operation of the reactor. The applicant summarizes the subprograms that form the CVAP in table 8.0-1 of the PSAR.

In PSAR section 7.1.2.2 the applicant states that reactor vessel internal components are classified as prototypes, and are screened and analyzed for effects due to potential excitation mechanisms in accordance with RG 1.20. In PSAR chapter 7, the applicant identifies the systems that will be included in the CVAP to evaluate the effects of flow-induced vibration. In PSAR section 12.5.3 flow-induced system vibration testing is included as part of the power ascension testing.

The staff evaluated the CVAP program description in chapter 8 of the PSAR its proposed application in chapter 7 of the PSAR and compared it to the guidance in RG 1.253 (content of application) and RG 1.20 (CVAP program development). Additionally, as part of an audit, the staff reviewed the supporting documentation for the CVAP program to verify the associated information in the PSAR. Based on comparison of the PSAR information to the relevant guidance, and confirmed in audit, the staff determined that the proposed CVAP program is acceptable to support the CP application for KU1.

The staff notes that, in PSAR table 8.0-1, the applicant references 10 CFR 50.55a(b)(3)(iii)(C),

and regulatory guidance in RG 1.20. The NRC has recently removed 10 CFR 50.55a(b)(3)(iii)(C) from the NRC regulations based on the applicability of other regulatory requirements to address flow-induced vibration in nuclear power plants, such as through the principal design criteria required by 10 CFR 50.34(a)(3)(i).

8.1.1.2 Equipment Qualification Program With respect to the Equipment Qualification (EQ) Program, table 8.0-1 in PSAR chapter 8 states that the EQ Program provides direction on establishing and performing equipment qualification, and defines how to ensure systems, components, and designs comply with applicable technical requirements, codes, standards, and regulatory requirements. The EQ program is applied to SSCs to ensure performance of design functions independent of safety classification. The EQ Program established disciplines of Electrical Equipment Qualification, Mechanical Equipment Qualification, and Instrumentation and Controls Equipment Qualification. The EQ Program includes development of testing and qualification specifications, processes, procedures, and reports. In PSAR table 8.0-1, the applicant references the 10 CFR Part 50 regulations in Appendix B, 10 CFR 50.49, 10 CFR 50.55(a)(h), 10 CFR 50.34(a)(3)(i), and regulatory guidance in RG 1.89, Rev. 2 which endorses with certain clarifications IEEE/IEC Std. 60780-323-2016 for satisfying regulatory requirements for environmental qualification of certain electric equipment important to safety for nuclear power plants, RG 1.100, Rev. 4, which endorses with certain exceptions and clarifications IEEE Std. 344-2013, ASME QME-1-2017, and IEEE Std. C37.98-2013 (IEEE/IEC Std. 60980-344-2020 and ASME QME-1-2023) for seismic qualification of electrical and active mechanical equipment and the functional qualification of active mechanical equipment for nuclear power plants, and RG 1.209, Rev. 0, which endorses with certain

8-4 enhancements and exceptions IEEE 323-2003 for satisfying the environmental qualification of safety-related computer-based instrumentation and control systems for service in mild environments at nuclear power plants. If the NRC has not generically accepted a referenced code or standard, the applicant will be expected to justify the use of that code or standard as part of the licensing process.

Mechanical Equipment Qualification The staff evaluated the EQ program description in the PSAR and compared it to the relevant guidance in RG 1.253 and RG 1.100, Rev.4. Additionally, the staff audited the applicants Mechanical Equipment Qualification Program supporting documents to better understand the proposed EQ program for mechanical equipment (see audit report MLXXXXXXXXX). Based on this review and supported by confirmation in the audit, the staff determined that the applicants approach is consistent with the guidance and there is reasonable assurance the applicants proposed mechanical EQ program is acceptable for the KU1 CP. The staff notes that the NRC has not yet endorsed the ASME QME-1-2023 Standard referenced in the Kemmerer Unit 1 PSAR. However, the NRC staff has reviewed ASME QME-1-2023 and considers the ASME standard to be acceptable for use as part of this PSAR.

Electrical Equipment Qualification The staff reviewed the program description in chapter 8 of the PSAR and compared that information to relevant guidance in RG 1.253 and RG 1.89, Rev.2. Additionally, the staff audited the applicants EQ program documents to better understand their proposed EQ program for electrical equipment (see audit report MLXXXXXXXXX). Based on this review and supported by confirmation in the audit, the staff determined that the applicants approach is consistent with the guidance and there is reasonable assurance that the applicants proposed EQ program will ensure that the EQ requirements specified in 10 CFR Part 50, Appendix B, 10 CFR 50.34(a)(3)(i), 10 CFR 50.55(a)(h), and 10 CFR 50.49 will be met.

Instrumentation and Controls Equipment Qualification The staff reviewed the program description in chapter 8 of the PSAR and compared that information to relevant guidance in RG 1.253, RG 1.209, Rev. 0 and RG 1.180, Rev. 2.

Additionally, the staff audited applicants Instrumentation and Controls Equipment Qualification preliminary documentation to better understand their proposed Equipment Qualification program for Instrumentation and Control Equipment (see audit report MLXXXXXXXXX). Based on this review and supported by confirmation in the audit, the staff determined that the applicants approach is consistent with the guidance and there is reasonable assurance that the applicants proposed Instrumentation and Controls Equipment Qualification program, when implemented, will ensure that the applicable requirements specified in 10 CFR Part 50, Appendix B, 10 CFR 50.55a(h), and 10 CFR 50.49 will be met.

8.1.1.3 Testing Program With respect to the Testing Program, PSAR table 8.0-1 indicates that the purpose of the Testing Program is to provide reasonable assurance that testing demonstrates that SR and NSRST SSCs perform satisfactorily in service. The applicant states that the Testing Program also identifies the programs and groups responsible for proof tests, preoperational tests, and operational tests needed to demonstrate compliance with the design. The applicant also states that the testing will comply with written test procedures. In PSAR table 8.0-1, the applicant

8-5 references the NRC regulations in 10 CFR 50.43(e)(1) and 10 CFR Part 50, Appendix B, and regulatory guidance in ASME NQA-1-2015, which is accepted in RG 1.28, with appropriate regulatory positions.

The staff reviewed the program description in chapter 8 of the PSAR and compared that information to relevant guidance in RG 1.253 and RG 1.28, Rev 6. Additionally, the staff audited the applicants Testing Program documents and, as part of that audit, the staff discussed the importance of early development of the Testing Program to meet 10 CFR 50.43(e)(1) to demonstrate the performance of each safety feature of the design of the Natrium reactor (see audit report MLXXXXXXXXX). Based on this review and supported by the audit, the staff determined that the applicants approach is consistent with the guidance and there is reasonable assurance that the applicants proposed Testing Program, when implemented, will ensure that the applicable requirements specified in 10 CFR Part 50, Appendix B and 10 CFR 50.43(e)(1) will be met.

8.1.1.4 Reliability and Integrity Management Program PSAR table 8.0-1 identifies the Reliability and Integrity Management (RIM) program as a special treatment program supporting the design and construction phase. The objective of the RIM program is to ensure that passive components within the RIM program scope achieve an acceptable level of reliability to support the plant probabilistic risk assessment throughout the life of the plant. The RIM program involves design interaction, performance monitoring, inspections, tests, maintenance, and replacements, as strategies to ensure the SSCs meet their reliability targets.

The applicant stated that the RIM program will be developed following the 2019 Edition of ASME BPVC Section XI, Division 2 (XI-2) and RG 1.246. The RIM program scope includes applicable SSCs whose failure could adversely affect plant safety and reliability. The RIM program requires conducting a degradation mechanism assessment (DMA) for each component in the scope of the RIM program, determining the reliability targets for the SSCs, and identifying RIM strategies to ensure reliability targets can be met.

8.1.1.4.1 Program Scope and Degradation Mechanism Assessment The staff reviewed the applicants preliminary RIM program scope and the criteria used to determine the SSCs to be included in the RIM program as described in Section 6 of NAT-13478 (ML25274A130). The applicant states that the RIM program scope includes SR and NSRST passive mechanical components. The staff notes that SSCs in the scope of RIM form the primary coolant pressure boundary, are credited as part of the functional containment, are required to ensure a flowpath for natural circulation in the primary system for emergency core cooling, among other functions that affect plant safety and reliability. The preliminary RIM program scope is consistent with the scope requirement in ASME BPVC XI-2 which requires passive SSCs whose failure could adversely affect plant safety and reliability to be included in the RIM program. Based on the consistency of the preliminary RIM program scope with ASME BPVC XI-2, the staff finds the preliminary design information relative to RIM scope acceptable at the CP stage. The final RIM program scope will be reviewed at the OL stage.

The staff reviewed the preliminary DMA and the screening criteria for the RIM program. Section 7 of NAT-13478 describes the DMA process and the preliminary DMA screening criteria. The applicant states that the DMA process is iterative, with the criteria being updated and expanded

8-6 upon as testing completes and the design matures. The staff evaluation of the preliminary DMA for SSCs exposed to sodium or cover gas is in section 7.1.2.1.4 of this SE.

8.1.1.4.2 Reliability Targets The staff reviewed the preliminary process for allocating and confirming reliability targets for SSCs that are in the RIM program scope as described in section 6 of NAT-13478. The applicant stated that reliability targets are set as the mean value of the failure rates associated with the component type and failure mode from the failure data used in the PRA. If the particular SSC failure mode is not included in the PRA data, then the target reliability is set based on a conservative failure effect and the associated failure rates for impacted mitigation functions, or the frequencies for SSCs whose failure causes an initiating event.

The staff recognizes that, given the preliminary nature of the PRA and the plant design, it is reasonable to not establish final reliability targets at the CP stage. Accordingly, the staff did not make any findings on the acceptability of the reliability target allocation or confirmation process or the preliminary reliability targets. The staff expects the applicant to provide, during the OL stage, the final reliability targets for SSCs that are in the RIM program, including the technical basis on how the reliability targets are derived as well as the associated uncertainties. For further evaluation of the reliability and capability targets for safety-significant SSCs, see section 6.2.3 of this SE.

In preparation for the OL review, the staff identified several areas where additional information will be needed to support review of the reliability targets related to SSCs in the RIM program.

Three particular areas are (1) the treatment of passive SSCs that are not typically modeled in a PRA, (2) how uncertainties associated with the reliability targets are considered, and (3) the process for confirming reliability targets. For the first area, ASME BPVC XI-2 RIM-2.4.3(b) indicates that for a component to be in the scope of RIM it must be included in the PRA.

However, the applicant has identified some SSCs that are not modeled in the PRA but will be included in the RIM program. The applicant will need to clearly identify a deviation from the endorsed RIM process for these SSCs and justify its approach to identifying reliability targets and RIM strategies for these SSCs to ensure they reliably perform their functions through the life of the plant. The staff will review how the RIM process addresses SSCs not modeled in the PRA but included in the RIM program in light of ASME BPVC XI-2 RIM-2.4.3(b) and the LMP guidance at the OL stage.

For the second area, ASME BPVC XI-2 RIM-2.4.2(c) states: The allocation of SSC-level Reliability Targets shall consider the uncertainties inherent in the prediction of SSC reliability.

However, the staff notes that the reliability target allocation process and preliminary reliability targets provided did not appear to include consideration of uncertainty. This is particularly important given the level uncertainty associated with passive SSC performance based on the limited knowledge and operating experience in sodium fast reactor conditions. The staff will review how the RIM process addresses uncertainty associated with establishing and meeting reliability targets consistent with the requirements of ASME BPVC XI-2 (including RIM-2.4.2(c))

and the LMP guidance at the OL stage.

For the third area, ASME BPVC XI-2 RIM-2.5.2 states: The impact of each RIM strategy on the reliability of each SSC in the scope of the RIM Program shall be assessed for comparison against the SSC-level Reliability Targets. ASME BPVC XI-2 RIM-2.5.1(c) similarly indicates that RIM strategies shall be selected to achieve and maintain SSC reliability consistent with their reliability targets. However, the staff noted that the applicants RIM process did not include a

8-7 step to assess, demonstrate or confirm that the reliability targets are met by the identified RIM strategies. This aspect of the process to confirm reliability targets, consistent with the requirements of ASME BPVC XI-2 and the LMP guidance, is expected to be included at the OL stage and the staff is open to qualitative or semi-quantitative approaches based on engineering judgement and appropriate consideration of uncertainties. The staff will review how the RIM process ensures reliability targets are met consistent with the requirements of ASME BPVC XI-2 and the LMP guidance at the OL stage.

8.1.1.4.3 RIM Strategy Development The staff reviewed the applicants proposed methodology to identify preliminary RIM strategies during the design and construction phase as described in section 6 of NAT-13478. The methodology provides prioritization for developing RIM strategies using a graded risk evaluation process utilizing the results of the reliability target allocation and DMA. The RIM risk evaluation includes considerations such as the design margin, component function, the risk and consequences of a postulated failure of the SSC, susceptibility of each applicable degradation mechanism for the SSC, feasibility and efficacy of monitoring and non-destructive examination (MANDE) for detecting degradation, and the power generation impact of the mechanism.

One aspect of the RIM strategy development the staff noted is the RIM Strategy Significance (RIMSS) categories, in which RIM strategies are binned into four categories. The staff notes that the lowest bin (RIMSS I) identifies that no MANDE would be provided for components least susceptible to a degradation mechanism or with a low probability or consequence of failure. The staff expectation is that any SSC in the scope of RIM that is deemed to have potentially active degradation should be covered by either a primary or expansion (with appropriate justification and connection to the primary RIM strategy) RIM strategy that is demonstrated to achieve and maintain the specified reliability target for that SSC. The scope of SSCs with potentially active degradation mechanisms binned into RIMSS I will be closely evaluated at the OL stage if this approach to RIM strategy binning is maintained to ensure those SSCs to which no MANDE is assigned are appropriate.

In addition, the staff reviewed the preliminary screening criteria and the preliminary RIM strategies and identified several concerns including consideration of long-term environmental effects on material properties in the design analysis, management of SSCs needed for safety-related core cooling, the combined effects of multiple degradation mechanisms, and the need for testing to characterize materials performance (see section 7.1.2.1.4 of this SE and audit report MLXXXXXXXXX). The applicant stated that the development of RIM strategies based on uncertainties under ASME BPVC XI-2 RIM-2.6 is a planned future activity and likely to lead to additional MANDE being identified.

The staff did not make any findings on the preliminary RIM strategies. The staff will review the final RIM program that will include final RIM strategies covering SSCs that are in the RIM program and technical basis on how the reliability targets are met with appropriate consideration of uncertainties during the OL stage.

8.1.1.4.4 Research and Development (R&D) Item on Assuring Materials Performance PSAR chapter 13 and section 12.1 of NAT-13478 describe a research and development (R&D) item associated with the development of information needed to provide assurance of adequate structural material performance for safety-significant SSCs included in the RIM program. The identified activities include research to improve understanding of the effects of high temperature,

8-8 chemistry exposure, and irradiation on materials; determination of requirements for further materials testing for environmental compatibility and activities to support the RIM program implementation. In order to provide assurance that the materials will perform as designed and to confirm that the applicant will develop and implement appropriate RIM strategies to address degradation mechanisms in safety-significant SSCs, the staff is proposing two permit conditions in the KU1 construction permit related to providing updates associated with this R&D item.

These permit conditions are discussed in greater detail in chapter 13 of this SE.

Based on the endorsement of ASME BPVC XI-2 in RG 1.246 and review of the preliminary RIM implementation documents and the associated permit conditions, the staff finds the applicants preliminary RIM program acceptable to support the design and construction stage. The staff will review final information at the OL stage. The staff expects the applicant to continue to update and develop the RIM program as design matures and detailed design information become available. The staff will review the final RIM program including the RIM scope, DMA, reliability targets, and the associated RIM strategies at the OL stage.

8.1.1.5 Design Reliability Assurance Program The objective of the D-RAP is to ensure that the plant is designed and constructed in a manner consistent with the risk insights and key assumptions (e.g., SSC design, reliability, and availability) derived from probabilistic, deterministic, and other analyses used to identify and quantify risk. This objective is achieved through implementation of the essential elements of the D-RAP, including organization, design control, procedures and instructions, records, and corrective actions during design and construction activities.

The purpose of the RAP is to provide reasonable assurance that:

The plant is designed, constructed, and operated consistent with the risk insights and key assumptions (e.g., SSC design, reliability, and availability)

RAP SSCs do not degrade to an unacceptable level of reliability, availability, or condition during plant operations Transients that challenge RAP SSCs are minimized RAP SSCs will function reliably when challenged.

The applicant confirmed during the audit (MLXXXXXXXXX) that the quality assurance (QA) programs (e.g., design, procurement, fabrication, construction, inspection, and testing activities) will provide control over activities affecting the quality of the RAP SSCs. QA controls for safety-related SSCs are established under 10 CFR Part 50, Appendix B. SRP Section 17.5, Part V, Nonsafety-Related SSC Quality Controls, addresses QA controls for non-safety-related RAP SSCs.

The KU1 D-RAP applies NEI 18-04 guidance, which directs inclusion of SR and NSRST SSCs in the RAP. In accordance with NEI 18-04 Section 4.1 (Tasks 4A and 4B), each non-safety-related SSC is evaluated for its risk significance. An SSC is considered risk significant if it is necessary to keep one or more LBEs within the F&C target or is significant in relation to one of the cumulative risk metric limits. Also, an SSC is risk significant when the total frequency of all LBEs with failure of the SSC exceeds 1 percent of the cumulative risk metric limit, based on mean estimates of frequencies and consequences. The cumulative risk metrics and limits, consistent with NEI 18-04, are as follows:

8-9 The total mean frequency of exceeding a site boundary dose of 100 mrem is less than 1/plant year.

The average individual risk of early fatality within 1 mile of the EAB is less than 5x10 per plant year.

The average individual risk of latent cancer fatality within 10 miles of the EAB is less than 2x10 per plant year.

The applicant clarified during the audit that the PRA is used to establish reliability targets for SR and NSRST SSCs. KU1 D-RAP uses the absolute risk metric approach of NEI 18-04 without deviation. For SSCs identified based on the basis of defense-in-depth or deterministic considerations, and not modeled in the PRA, the reliability target is a conservative value established by the system designer.

Regarding the availability targets, the applicant clarified during the audit that most passive SSCs are expected to remain in service for the plant operating states for which they provide a mitigation function. However, the availability targets cannot be finalized until the design is completed and testing, maintenance, and inspection requirements are identified. Once the final design is complete, any anticipated out-of-service times for SSCs required during operating states will be incorporated into the PRA to confirm acceptability with the LMP results.

The applicant clarified that prior to initial fuel load, the reliability of D-RAP SSCs will be ensured through implementation of the QA program, the RIM program, the equipment qualification program, and the testing program. These programs ensure that reliability targets are incorporated into the design, procurement, fabrication, qualification, inspection, and testing of RAP SSCs. The QA program for D-RAP and implementing quality controls to ensure that relevant inputs (e.g., RAP SSC lists, PRA models, risk insights, and key assumptions) will meet QA requirements and specifications. These activities will be carried out under the KU1 QAPD and associated implementing procedures. The QAPD provides direction for procurement, fabrication, inspection, and testing activities.

The staff reviewed the applicants submittal and supporting documents and finds that the D-RAP, including the described process and identified RAP SSCs, is reasonable for the construction permit stage, given the limited scope of PRA and design information currently available. The interim staff guidance DC/COL-ISG-018 states:

The scope of RAP should not be limited to risk-significant SSCs modeled in the PRA. SSCs that are not modeled in the PRA should also be evaluated for inclusion (e.g., by deterministic or other methods). The scope of RAP should include safety-related and non-safety-related SSCs identified as risk-significant (or significant contributors to plant safety).

For passive system designs, RAP should also include all SSCs subject to regulatory treatment of non-safety systems (RTNSS).

Therefore, the staff expects the applicant to continue implementing and updating the KU1 D-RAP during the construction phase using the RAP expert panel, consistent with ISG guidance, to enhance the RAP process and SSC list.

Based on the review of the applicants D-RAP description, methodology, and QA implementation, the staff determines that the KU1 D-RAP provides reasonable assurance, for the purposes of the construction permit stage, that the design and construction of RAP SSCs

8-10 will be consistent with the key assumptions and risk insights derived from the PRA and other supporting analyses.

8.1.1.6 Inservice Testing Program With respect to the IST Program, PSAR Table 8.0-1 indicates that the IST Program provides the requirements for IST activities for SR and NSRST pumps, valves, and dynamic restraints to assess their operational ability to perform their specified functions. The applicant states that the IST program describes baseline testing, IST activities, examination, and monitoring needed to provide reasonable assurance of the operational readiness of the components. In PSAR Table 8.0-1, the applicant references the NRC regulations in 10 CFR 50.34(b)(6)(iv), and guidance in the draft ASME OM-2 Code.

The staff compared the guidance to the description of the IST program in the PSAR.

Additionally, the staff audited the Natrium IST Program documents (see audit report MLXXXXXXXXX). Based on the review and confirmed during the audit, the staff determined the IST program is acceptable to support the CP application for KU1.

8.1.2 Conclusion Based on the review described above, the staff concludes that the information in PSAR section 8.0 is sufficient and meets applicable guidance and regulatory requirements identified in this chapter for the issuance of a CP. Further information as may be required to complete the review can reasonably be left for later consideration when the final design with the final plant programs review completed at the OL stage.

Table 8.1-1: Special Treatment Program Applicable Regulations and Guidance Program Title SSC Applicability Applicable Regulation Applicable Guidance Special Treatment Programs Initiated to Support Design and Construction Human Factors Engineering SR, NSRST, non-safety related with no special treatment (NST) 10 CFR 50.34(f)(2)(iii)

DANU-ISG-2022-05, Advanced Reactor Content of Application Project Chapter 11, Organization and Human-System Considerations Interim Staff Guidance Comprehensive Vibration Assessment Program SR, NSRST, NST 10 CFR 50.34(a)(3)(i);

NATD-LIC-RPRT-0002-A, Principal Design Criteria for the Natrium Advanced Reactor, Rev 1

RG 1.20, Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Startup Testing Equipment Qualification Program SR and NSRST 10 CFR Part 50, Appendix B; 10 CFR 50.34(a)(3)(i); 10 CFR RG 1.89, Environmental Qualification of Certain Electric Equipment

8-11 50.55(a)(h); and 10 CFR 50.49 Important to Safety for Nuclear Power Plants.

RG 1.100, Seismic Qualification of Electrical and Active Mechanical Equipment for Nuclear Power Plants. RG 1.209, Guideline for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants.

Testing Program SR, NSRST 10 CFR Part 50, Appendix B, and 10 CFR 50.43(e)(1)

ASME NQA-1, Quality Assurance Requirements for Nuclear Facility Applications.

Reliability and Integrity Management Program SR, NSRST, NST 10 CFR 50.34(b)(6)(iv)

RG 1.246, Acceptability of ASME Code,Section XI, Division 2, Requirements for Reliability and Integrity Management (RIM)

Programs for Nuclear Power Plants, for Non-Light Water Reactors.

Design Reliability Assurance Program SR, NSRST 10 CFR Part 50, Appendix B DANU-ISG-2022-01 Section 4, Quality Assurance Plan Inservice Testing Program SR, NSRST 10 CFR 50.34(b)(6)(iv)

No final guidance currently available for the ASME OM-2 Code referenced in the PSAR.

Post-Construction Inspection, Testing and Analysis Program SR, NSRST 10 CFR 50.34(a)(7); 10 CFR 50.35; 10 CFR 50.40; 10 CFR Part 50, Appendix B DANU-ISG-2022-06, Advanced Reactor Content of Application Project Chapter 12 -

Post-construction Inspection, Testing, and Analysis Program Quality Assurance Program SR, NSRST 10 CFR 50.34(a)(7);

10 CFR 50.35; 10 CFR 50.40; 10 CFR

8-12 50.55(f); 10 CFR Part 50, Appendix B Additional Special Treatment Programs for Operations Fire Protection Program SR, NSRST 10 CFR 50.48 Not Applicable (NA) at this time because USO is only seeking a CP at this time that does not include a request for byproduct, source or special nuclear material. As such a fire protection program is not required.

Leakage Program SR, NSRST 10 CFR 50.34(f)(2)(xxvi)

NA - program not required at the CP stage Maintenance Rule Program SR, NSRST 10 CFR 50.34(b)(6)(iv),

10 CFR 50.65(a) and (b),

10 CFR 50.69 NA - program not required at the CP stage.

Emergency Preparedness Program N/A 10 CFR 50.33(g)(2), 10 CFR 50.34(a)(10)(ii), and 10 CFR 50.160.

NA - program not required at the CP stage.

8-19 8.3 Fire Protection PSAR Section 8.2 describes general fire protection design considerations, sodium leakage concepts, fire analyses, and the fire protection program. This section also states that the design maintains adequate fire protection defense in depth in all fire areas and takes into account prevention, detection, containment, suppression, and the ability to maintain safe shutdown in accordance with 10 CFR 50.48(a), PDC 3, and Regulatory Guide (RG) 1.189, Revision 5.

PSAR Section 8.2 further states that fire protection requirements are considered in the design of the Natrium reactor plant for both conventional and sodium fires and that the conventional fire protection design uses approved guidance, methodologies, and codes and standards. For sodium fire protection, the KU1 design uses sodium fast reactor industry information to develop Natrium-specific guidance, requirements, and criteria.

8.3.1 Technical Evaluation As discussed in RG 1.189, the CP application should provide assurance of that the fire protection systems will meet the relevant requirements in 10 CFR 50.48(a).

8.3.1.1 Cells and Guard Vessels PSAR section 8.2.1 discusses various methods of limiting sodium interactions with air and concrete including seals and inert gases, cells, guard pipes, guard enclosures, clamshells, catch pans, inerted guard vessels, leakage detection, materials, and fire areas and barriers. The applicant states that seals and inert gases are used as part of individual sodium-containing system design and that inerting is used in areas or cells containing radiological sodium to mitigate sodium-air reactions and the airborne release of radionuclides. In addition, cells are rooms with a limited leak rate and known oxygen volume which is either actively inerted or becomes de facto inerted upon consumption of the limited available oxygen after a sodium leak.

Finally, PSAR section 8.2.1.7 describes how inerted guard vessels are used to prevent sodium-air and sodium-concrete reactions in the event a leak from another vessel containing sodium (e.g., the reactor vessel-guard annulus) and that leak detection is provided within the inerted space.

The staff reviewed the information provided with respect to inerted or limited oxygen spaces, such as cells and guard vessels, and notes that these are generally effective means of preventing or mitigating sodium-air reactions and are reasonable to use in various areas of the plant. Based on the effectiveness of well-designed and monitored cells and guard vessels, the staff finds that the preliminary design of cells and guard vessels is consistent with PDC 3 and 73.

At the preliminary design stage, there are limited details of how these SSCs will be designed to perform their function to prevent or mitigate sodium fire. Given that sodium fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final design of these SSCs and their ability to perform their functions reliably in the event of postulated sodium leakage.

8.3.1.2 Guard pipes and enclosures The applicant described guard pipes, which are currently applied for systems containing radiological sodium and act as a full secondary pressure boundary in the event of a leak

8-20 designed to the same conditions as the protected system pipe. In addition, guard pipes may be inerted or not inerted. If not inerted, the conditions within the annulus are considered for the design rating of the guard pipe. Guard enclosures are similar to guard pipes, with the addition of a drainage to an acceptable location such as a drain tank cell via NNA drain line or within the process cell to catch pans. For cells which have an inerted environment, the catch pans will not have a suppression deck. For cells without an inerted environment, the cell will either be evaluated for combustion or a suppression deck will be included. Guard enclosures are qualified to withstand design-basis sodium leaks and are applied to high stress piping, nozzles, and components in locations where sodium-leak interactions with safety-significant components cannot be tolerated, such as the Head Access Area (HAA).

The applicant further describes a clamshell as a form of guard enclosure placed specifically where piping inspections are expected to be needed. These will be used for IHT SSCs located within the HAA and form the fire area and boundary between the RAB pipe chase room and the HAA. The applicant states that this feature of the clamshell enclosures to allow leaked sodium to flow into the RAB pipe chase helps ensure pressurized sodium in the clamshell cannot spray into the HAA. PSAR Section 7.5.2.3.4 states that first-of-a-kind aspects of the clamshell design will be tested to ensure it can perform its required function with testing methods and acceptance criteria to be identified in the OL application. Pre-service and in-service testing as required by the RIM program will be performed. During audit discussions, the applicant indicated that testing will be performed on the clamshell enclosures to ensure they can maintain effectiveness when exposed to design temperatures and pressures and is expected to include leak-tight testing and pressure testing after installation.

The staff reviewed the information provided with respect to guard pipes and enclosures, including clamshell enclosures. The staff notes that guard pipes and enclosures are generally an effective means of preventing or mitigating sodium-air reactions and are reasonable to use in various areas of the plant based on the risk of sodium leakage and fire. In particular, the staff notes that the proposed use of clamshell enclosures to prevent sodium leakage from IHT piping from entering the HAA and causing a sodium fire in a critical area of the plant is very important.

The staff notes that the applicant plans to perform testing of the clamshell design as well as leak-tightness and pressure testing of these SSCs after installation. Based on the information noted above, the staff finds that the preliminary design of guard pipes and enclosures, including clamshell enclosures, is consistent with PDC 3 and 73.

At the preliminary design stage, there are limited details of how these SSCs will be designed to perform their function to prevent or mitigate sodium fire. Given that sodium fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final design of the guard pipes and enclosures, including the clamshell enclosures, to ensure their ability to perform their functions reliably in the event of postulated sodium leakage (which is discussed in SE sections 8.3.1.5 and 8.3.1.6 below).

8.3.1.3 Catch Pans, Leakage Detection, and Materials PSAR section 8.2.1.6 states that catch pans are steel-lined volumes designed to retain the design-basis sodium leak volume plus margin and are used as drain locations that collect leaked sodium from guard enclosures or sloped floors. Catch pans are also placed in areas where operational leakage may occur (e.g., a sodium containing heat exchanger). The

8-21 applicant also states that suppression decks are applied to catch pans where the environmental oxygen volume cannot be limited.

PSAR section 8.2.1.8 states that leakage detection equipment for liquid sodium, sodium aerosols, and reaction products, is provided to inform plant operators in the event of a leak and that leakage detectors meet PDC 73 and the applicable guidance in RG 1.189. The PSAR further clarifies that leak detection is provided at multiple points between pipe supports along all normally sodium containing systems (i.e., drain lines excepted), within either the guard pipe, guard enclosure, or pipe insulation and that the leak detectors are capable of detecting leaks within the entire length of pipe it covers. In addition, conventional fire detection is installed in non-inerted areas to provide supplemental sodium leak detection.

PSAR section 8.2.1.9 states that the design of SSCs containing sodium considers the use of materials compatible with sodium to minimize the effects of sodium interactions, including using insulation that has low chemical reactivity with sodium. The applicant also states that in consideration of conventional fires, less combustible materials (e.g. Spectrasyn 10 for ISP and PSP lubrication and cooling fluid due to its high flash point) are chosen in the design of SSCs and that additional information regarding the materials being used for the final design will be further described at the OL stage.

The staff reviewed the information provided with respect to catch pans, leakage detection, and materials. The staff notes that catch pans and suppression decks have been used in past sodium reactor designs to mitigate and manage sodium-air reactions but require careful consideration of sizing and design to address heat loads from leaked sodium and the ensuing sodium-air reaction. The staff notes that the use of less combustible materials is a prudent approach to reduce the hazard from conventional and sodium fires. In addition, the staff notes the proposed approach to leakage detection for guard vessels, pipes, enclosures, and pipe insulation is thorough and important to enable early detection and response to sodium leaks.

Based on the information noted above, the staff finds that the preliminary design of catch pans, leakage detection, and materials is consistent with PDC 3 and 73.

At the preliminary design stage, there are limited details of how these SSCs will be designed to perform their function to prevent or mitigate sodium fire. Given that sodium fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final design of the catch pans, leakage detection, and materials to ensure their ability to perform their functions reliably in the event of postulated sodium leakage.

8.3.1.4 Fire Areas and Barriers PSAR section 8.2.1.10 states that fire areas are established and that redundant trains of safety-significant and safe shutdown SSCs are located in separate fire areas per the guidance from RG 1.189 section C.8.2. Separation by physical barriers and selective positioning of safety-significant SSCs ensure a fire is contained within a single fire area and a safe shutdown path is maintained regardless of the location of fire initiation. The applicant further states that fire barriers are designed to withstand the most restrictive hazard conditions (temperatures, pressures, etc.) postulated in the established fire areas with fire scenarios determined and fire area separation design requirements for each fire area validated through modeling and analysis.

Fire modeling will be used to run sensitivity analyses to determine which leak fire scenarios

8-22 create the most stringent design conditions for each area and ensure the mitigation features provide adequate defense-in-depth adequate for the hazards in that fire area.

The staff reviewed the information provided with respect to fire areas and barriers and notes the planned use of RG 1.189 to establish appropriate separation by physical barriers and physical separation ensure a safe shutdown path is maintained. The staff notes that fire scenarios will be modeled and analyzed with sensitivity analyses to validate fire area separation design requirements are met for each fire area. Based on conformance with RG 1.189 and the information noted above, the staff finds that the preliminary design information related to fire areas and barriers is consistent with PDC 3 and 73.

At the preliminary design stage, there are limited details of how the plant and specific fire protection SSCs will be designed to perform their function to prevent or mitigate sodium fire.

Given that fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final design of the fire areas and barriers to ensure their ability to perform their functions reliably in the event of postulated sodium leakage.

8.3.1.5 Sodium Leakage Concepts PSAR Section 8.2.2 discusses sodium leakage concepts and design provisions for sodium fire protection including design basis leakage into a guard enclosure, design basis leakage within a process cell, beyond design basis leakage, and operational leakage. PSAR section 8.2.2.2 states that design basis sodium leakage within a guard enclosure is designed to be contained within the enclosure and then directed into a catch pan through a drain line. The PSAR states that sodium-air reactions may occur within guard enclosures as they are not inerted but the guard enclosure will be designed based on analyses of potential reactions within the enclosure.

In addition, the drain lines will be heat traced, where needed, to minimize sodium freezing and sized to accommodate the design basis leak rate.

PSAR section 8.2.2.3 describes design basis leakage within process cell and states that each cell is designed to contain or drain design basis leakage through sloped floors and steel plates to a catch pan. For process cells that have an inerted environment, the catch pan does not include a suppression deck. For cells without an inerted environment, the cell is either evaluated for sodium-air reactions based on available oxygen, or a suppression deck is included. The staff notes that PSAR section 8.2.1.6 states that this type of catch pan will include suppression decks where the environmental oxygen volume cannot be limited.

PSAR section 8.2.2.4 describes beyond design basis leakage. Leaks postulated during beyond design basis events will be evaluated for acceptable consequences with respect to the Quantitative Health Objectives as part of the integrated risk assessment performed as described in Chapter 3 of the PSAR, considering worst case, maximum leak sizes as applicable. The PSAR states that additional leak mitigation features may be added to address unacceptable consequences.

The staff reviewed the information provided by the applicant with respect to sodium leakage concepts. At a conceptual level, the staff notes that the approach to addressing sodium leakage within guard enclosures and process cells appears reasonable. However, this conceptual approach is preliminary with many details of how the design, modeling, and analysis of sodium leakage and the associated fire hazard to be determined in the OL application. Similarly, the

8-23 staff notes the limited description of how design basis and beyond design basis leakage from sodium-containing SSCs (piping or vessels) will be established and analyzed, which will also need to be justified at the OL stage.

PSAR section 8.2.2.1 states that catch pans are provided for operational leakage only for asset protection and ease of cleanup. Since this concept and these features are not credited for safety purposes, the staff did not evaluate this information further. The staff notes that the description of operational leakage identifies it as being assumed to be less than the design basis leak rate.

Therefore, the staff focused its review on design basis and beyond design basis leakage and the features credited to detect, prevent, and mitigate design basis and beyond design basis leakage.

8.3.1.6 Sodium Leak Postulation PSAR section 8.2.2 states that leaks are postulated for all piping that normally contains sodium.

PSAR Section 8.2.3.3 further describes the approach to sodium leak postulation. Postulated design basis leakage crack sizes from piping areas subject to high stress are established as a Dt/4 (diameter multiplied by thickness divided by 4) crack for seismically qualified piping and a full-circumferential break for non-seismically qualified piping. This approach is adapted from the guidance found in Branch Technical Positions (BTPs) 3-3 and 3-4 from NUREG-0800 for moderate energy piping for light water reactors. The PSAR further states that design basis leak locations are determined using additional criteria for high-temperature creep considerations from ASME BPVC Section III Division 5 in addition to the guidance in BTP 3-4.

PSAR section 8.2.3.3 also describes the approach taken to postulating leakage in areas identified as lower stress. The PSAR identifies two exceptions based on stress criteria derived from BTP 3-4 with a third exception tied to not assuming leakage from sealed guard piping simultaneously with leakage from the system piping. PSAR section 8.2.3.3 indicates that lower stress areas meeting any of the three exceptions are excluded from considering a Dt/4 or full circumferential break as their design basis leakage size. However, design basis leakage cracks will be postulated in these lower stress locations with the postulated crack sizes developed using a methodology to be provided at the OL stage.

The staff reviewed the information provided by the applicant with respect to sodium leakage and sodium leak postulation. The staff notes that postulating leaks for all piping that normally contains sodium is a reasonable approach. However, the staff also notes that the PSAR does not describe an approach to postulation of sodium leaks from vessels. The specific justification for sodium leak postulation will need to be provided at the OL stage; alternatively, without an accepted approach, the staff expects leaks to be postulated from any sodium-containing SSC with a single barrier, including vessels.

The staff notes the use of a graded approach to sodium leak postulation for design basis and beyond design basis leakage based on the probability of leakage is generally reasonable.

However, the staff also observes that BTP 3-3 and 3-4 are part of NUREG-0800 guidance that is specific to LWRs. In particular, BTP 3-3 and 3-4 are focused on large leaks and pipe ruptures in LWRs creating dynamic effects that could damage other SSCs. However, in a sodium system even relatively small leaks can create a significant fire hazard that needs to be managed effectively. Therefore, the staffs general observation with applying BTP 3-3 and 3-4 to KU1 is that the criteria and basis of leak postulation are for a fundamentally different purpose than in LWRs and additional justification will need to be provided at the OL stage. In addition, there are significant differences between KU1 and LWRs, such as operating pressure and temperature,

8-24 coolant type and potential degradation modes that must be considered in applying these LWR criteria to a sodium reactor design. Therefore, the staff observes these conceptual approaches to be a reasonable way to analyze and address sodium leakage which is acceptable at the CP stage, but the specific criteria and basis have not yet been developed and justified for the staff to make a finding on the use of BTP 3-3 and 3-4, adapted to the Natrium design at this time.

Further, the staff notes that applicant proposes to develop a methodology to postulate design basis leakage from lower stress areas of piping. This methodology will need to be developed with appropriate technical basis and justified for how it is applied to the KU1 design.

Based on the discussion above, the staff determined that the preliminary approach related to sodium leak postulation to be consistent with PDC 3 and 73. However, at the preliminary design stage, there are limited details of how sodium leakage will be managed and analyzed to ensure SSCs will be designed appropriately to manage the sodium fire hazard. Given that fire hazards are highly dependent on the final design details and further information related to sodium leak postulation and analysis, as well as a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final design for appropriate sodium leak postulation approaches and analysis to ensure the plant and specific SSCs can perform their functions reliably consistent with PDC 3 and 73.

8.3.1.7 Fire Analyses PSAR Section 8.2.3 discusses the fire analyses and indicates that fire modeling, fire PRA, fire hazards analysis, and fire safe shutdown analysis will be performed as part of the OL application. The PSAR states that as the Natrium design progresses, strategies to manage the hazards identified will be developed and may include manual firefighting, automatic or manual suppression with agents compatible with sodium, inerting, purging, using incombustible materials, and SSC specific design features. As discussed above, the KU1 fire protection design strategy follows the guidance in RG 1.189, Rev. 5 and focuses on the designation of fire areas supported by the design of fire barriers such that a protected path of safe shutdown is maintained while assuming all equipment in any one fire area is rendered inoperable by fire.

PSAR section 8.2.3.1 states that the KU1 design uses NUREG/CR-6850 EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities to provide a structure for developing a fire PRA, including to develop the sodium frequency assessment, scenario development, and fire modeling. The PSAR states that a sodium fire modeling methodology with complete verification and validation and inputs to support the fire PRA will be developed at the OL stage. PSAR section 8.2.3.2 states that the design ensures safe shutdown capability is maintained in the event of a fire by limiting fire damage to a single train of safety-significant SSCs so that the capability to shut down the plant safely is ensured. The PSAR also states that the design uses methodologies approved by the NRC to demonstrate adequate defense-in-depth and safe-shutdown margin is available.

The staff reviewed the information provided with respect to the fire analyses, including the fire modeling, fire PRA, fire hazards analysis and safe shutdown analyses. The staff notes that the fire analysis methods will follow RG 1.189, Revision, 5 and NUREG/CR-6850. Additionally, the staff notes that a sodium fire modeling methodology with complete verification and validation for inputs to support the fire PRA will be developed to meet the ASME/ANS RA-S-1.4-2021, Probabilistic Risk Assessment Standard for Advanced Non-Light Water Nuclear Power Plants.

Based on conformance with RG 1.189, NUREG/CR-6850 and the plan to develop a sodium fire

8-25 modeling methodology with complete verification and validation for inputs to support the fire PRA, the staff finds that the preliminary design information related to fire analyses is consistent with PDC 3 and 73.

At the preliminary design stage, there are limited details of how these SSCs will be designed to perform their function to prevent or mitigate sodium fire. Given that sodium fire hazards and conventional fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration. At the final design stage in the OL review, the staff will evaluate the final methodology development, verification and validation effort, fire hazards analysis and the safe shutdown analysis to ensure their ability to perform their functions reliably in the event of postulated sodium leakage and conventional fire risk. This will include a comprehensive review of the inputs used for the scoping fire models that will be used to inform design decisions and fire conditions expected for each postulated sodium leak scenario.

8.3.1.8 Fire Protection Program PSAR section 8.2.4 indicates that a fire protection program will be developed at the operating license stage to meet 10 CFR 50.48(a) using the guidance in RG 1.189, Revision 5. PSAR section 8.2.4.1 states that there is no active or manual fire suppression for large sodium fires and that sodium leaks and any resulting sodium fires are mitigated by passive means through NNA design features.

USO states that firefighting strategies for conventional deep-seated fires will be described in the fire protection plan submitted at the OL stage. USO also states that deep-seated sodium fire are contained within the NNA features and are passively self-suppressing. USO also states that open sodium leaks in the HAA are prevented by containing leakage from the primary Sodium Processing System with guard piping of the system and containing leakage from the Intermediate Heat Transport System within a leak guard enclosure that drains to the RAB pipe chase.

USO states that the RXB and RAB do not contain water-based fire suppression systems. Water-based fire suppression systems are contained in the FHB, which also has sodium-containing SSCs, but USO also states that the FHB establishes fire barriers or other mitigative features to separate areas containing sodium from areas protected by a water-based suppression system.

The staff reviewed the information provided with respect to the fire protection program and notes that the fire protection program will meet 10 CFR 50.48(a) and the guidance in RG 1.189, Revision 5. The staff review noted that some aspects of the conventional firefighting strategies cannot be utilized in the RXB, RAB and portions of the FHB due to the limitations surrounding the use of water for concerns surrounding sodium-water interactions. The firefighting strategies for both incipient and deep-seated fire conditions will be further developed during the OL license stage. The staff notes that incipient firefighting strategies considered may include one or some combination of manual firefighting, automatic or manual suppression with agents compatible with sodium, inerting, purging, using incombustible materials, and SSC specific design features.

Based on plans to meet 10 CFR 50.48(a), follow the guidance in RG 1.189, and identification of various firefighting strategies, the staff finds that the preliminary design information related to the fire protection program is consistent with PDC 3 and 73. The staff particularly notes that deep-seated firefighting strategies may require offsite response with specialized training, which will require significant description and justification at the OL stage. Additional discussion regarding Fire Protection can be found in SE section 7.5.2.

8-26 At the preliminary design stage, there are limited details of how the fire protection program will be designed to perform its design function to prevent or mitigate conventional and sodium fires in areas where water cannot be utilized. Given that sodium fire hazards and conventional fire hazards are highly dependent on the final design configuration and a fire PRA will be performed as part of the OL application, further information can reasonably be left for later consideration.

At the final design stage in the OL review, the staff will evaluate the fire protection program to ensure the ability to perform its functions reliably in the event of postulated sodium leak and/or conventional fire. This will include a review of the installed fire protection suppression systems, the firefighting strategies which will be used for the various fire conditions, the design features which will enable passive and self-suppressing sodium fires and any reliance on off-site fire brigade support to ensure the ability to achieve safe shutdown.

8.3.2 Conclusion Based on the review described above, the staff concludes that the information in PSAR section 8.2 is sufficient and meets applicable guidance and regulatory requirements identified in this chapter for the issuance of a CP. Further information as may be required to complete the review can reasonably be left for later consideration when the final design with the full scope fire protection plan, the fire PRA, and the safe shutdown analysis have been completed at the OL stage.