ML24131A116

NPM-20 - Nonproprietary - NuScale SDAA Section 19.2 - Request for Additional Information No. 024 (RAI-10185-R1)

ML24131A116
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Site:	99902078
Issue date:	05/10/2024
From:	NRC
To:	NRC/NRR/DNRL/NRLB
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From:

Getachew Tesfaye Sent:

Friday, May 10, 2024 1:18 PM To:

Request for Additional Information Cc:

Alina Schiller; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Vance, Lindsay; NuScale-SDA-720RAIsPEm Resource

Subject:

Nonproprietary - NuScale SDAA Section 19.2 - Request for Additional Information No. 024 (RAI-10185-R1)

Attachments:

SECTION 19.2 - RAI-10185-R1-FINAL NON-PROPRIETARY.pdf Attached please find NRC staffs nonproprietary request for additional information (RAI) concerning the review of NuScale Standard Design Approval Application for its US460 standard plant design (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23306A033). The encrypted proprietary version will be submitted in a separate email.

Please submit your technically correct and complete response by the agreed upon date to the NRC Document Control Desk.

If you have any questions, please do not hesitate to contact me.

Thank you, Getachew Tesfaye (He/Him)

Senior Project Manager NRC/NRR/DNRL/NRLB 301-415-8013

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NuScale_SDA720_RAI_Public Email Number:

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Subject:

Nonproprietary - NuScale SDAA Section 19.2 - Request for Additional Information No. 024 (RAI-10185-R1)

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5/10/2024 1:18:09 PM Received Date:

5/10/2024 1:18:13 PM From:

Getachew Tesfaye Created By:

Getachew.Tesfaye@nrc.gov Recipients:

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Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>

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1 REQUEST FOR ADDITIONAL INFORMATION No. 024 (RAI-10185-R1)

BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 19, PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT SECTION 19.2, SEVERE ACCIDENT EVALUATION EVALUATION ISSUE DATE: 05/10/2024

=

Background===

By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR), Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, Revision 1 (Agencywide Documents Access and Management System Accession No.ML23304A385) and Chapter 6, Engineered Safety Features, Revision 1 (ML23304A345) of the NuScale Standard Design Approval Application (SDAA) for its US460 standard plant design. The applicant submitted the US460 standard plant SDAA in accordance with the requirements of Title 10 Code of Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, Subpart E, Standard Design Approvals. The NRC staff has reviewed the information in Chapters 6 and 19 of the SDAA, and other chapters as necessary, and has determined that additional information is needed to complete its review.

Question 19.2-1 Regulatory Basis:

10 CFR 52.137(a)(2) requires a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon which the requirements have been established, and the evaluations required to show that safety functions will be accomplished.

10 CFR 52.137(a)(4) requires analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

10 CFR 52.137(a)(9) requires, for applications for light-water cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application.

10 CFR 52.137(a)(12) requires an analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter.

2 10 CFR 52.137(a)(23) a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass.

10 CFR 52.137(b) requires, in part, an application for approval of a standard design, which differs significantly from the light-water reactor designs of plants that have been licensed and in commercial operation before April 18, 1989, or uses simplified, inherent, passive, or other innovative means to accomplish its safety functions, must meet the requirements of 10 CFR 50.43(e), as identified below.

10 CFR 50.43, Additional standards and provisions affecting class 103 licenses and certifications for commercial power, states, in part, the following:

(e) Applications for a design certification, combined license, manufacturing license, or operating license that propose nuclear reactor designs which differ significantly from light-water reactor designs that were licensed before 1997, or use simplified, inherent, passive, or other innovative means to accomplish their safety functions, will be approved only if:

(1)(i) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (ii) Interdependent effects among the safety features of the design are acceptable, as demonstrated by analysis, appropriate test programs, experience, or a combination thereof; and (iii) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions 10 CFR 50.44(c) Requirements for future water-cooled reactor applicants and licensees.The requirements in this paragraph apply to all water-cooled reactor construction permits or operating licenses under this part, and to all water-cooled reactor design approvals, design certifications, combined licenses or manufacturing licenses under part 52 of this chapter, any of which are issued after October 16, 2003.

(1) Mixed atmosphere. All containments must have a capability for ensuring a mixed atmosphere during design-basis and significant beyond design-basis accidents.

(2) Combustible gas control. All containments must have an inerted atmosphere, or must limit hydrogen concentrations in containment during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100 percent fuel clad-coolant reaction, uniformly distributed, to less than 10 percent (by volume) and maintain containment structural integrity and appropriate accident mitigating features.

(3) Equipment Survivability. Containments that do not rely upon an inerted atmosphere to control combustible gases must be able to establish and maintain safe shutdown and containment structural integrity with systems and components capable of performing their

3 functions during and after exposure to the environmental conditions created by the burning of hydrogen. Environmental conditions caused by local detonations of hydrogen must also be included, unless such detonations can be shown unlikely to occur. The amount of hydrogen to be considered must be equivalent to that generated from a fuel clad-coolant reaction involving 100 percent of the fuel cladding surrounding the active fuel region.

(4) Monitoring. (i) and (ii) Equipment must be provided for monitoring oxygen in containments that use an inerted atmosphere for combustible gas control. Equipment for monitoring oxygen and hydrogen must be functional, reliable, and capable of continuously measuring the concentration of oxygen in the containment atmosphere following a significant beyond design-basis accident for combustible gas control and accident management, including emergency planning.

(5) Structural analysis. An applicant must perform an analysis that demonstrates containment structural integrity. This demonstration must use an analytical technique that is accepted by the NRC and include sufficient supporting justification to show that the technique describes the containment response to the structural loads involved. The analysis must address an accident that releases hydrogen generated from 100 percent fuel clad-coolant reaction accompanied by hydrogen burning. Systems necessary to ensure containment integrity must also be demonstrated to perform their function under these conditions.

10 CFR 50.12, Specific exemptions, Section (a). The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are (1) Authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.

(2) The Commission will not consider granting an exemption unless special circumstances are present 10 CFR Part 50, Appendix A, General Design Criteria Criterion 1Quality standards and records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

Criterion 4Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

Criterion 41Containment atmosphere cleanup. Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to

4 the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.

Criterion 42Inspection of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components,to assure the integrity and capability of the systems.

Criterion 43Testing of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural integrity of its components, (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation.

Issue:

A primary post-accident safety function is to protect the containment integrity. The passive autocatalytic recombiner (PAR) is a new design component that was not included in the US600 DCA design. The NuScale US460 SDAA design credits a single PAR for preventing a combustible mixture in the containment vessel (CNV) for design-basis events (DBEs), including anticipated operational occurrences (AOOs) that are expected to occur at least once in the lifetime of the plant, and during a severe accident (SA). The PAR is the basis for SDAA, Part 7, Section 2 (ML23304A389), which requests an exemption to 10 CFR 50.44(c)(4) for hydrogen and oxygen monitoring in the containment and is a new exemption request compared to the DCA. NuScales analysis demonstrates that (( Based on its analysis on the presence and treatment of combustible gas in the RCS in (( }}, which includes an evaluation of the scenario described above, NuScale concludes, The results found that (( }} (emphasis added). To perform its function unimpaired during DBEs, including AOOs, and SAs, the single PAR must be designed to withstand the normal operating environment of 60 years and the post-accident environment. The operational environment for the PAR in the US460 design during normal operations, and DBE and SA conditions is markedly different from that in operating reactors. As an example, for normal operation, the PAR needs to survive the high radiation environment of both neutrons and gammas to which it is exposed, which is different from operating reactors where PARs are usually located outside the bioshield and not exposed to neutron radiation. The main impacts of neutron irradiation would be PAR material embrittlement and activation. As

5 such, qualification testing is needed to demonstrate the integrity of the PAR after 60 years of neutron and gamma irradiation or to inform an appropriate PAR life cycle management strategy. As another example, the PAR in the US460 design will be exposed to jet impingement forces as well as mechanical loads due to the mass and energy release to the CNV during the DBE and severe accidents, which is different from operating reactors. Qualification testing or a combination of testing and analysis would address these jet impingement loads. Post accident, the PAR must function during and after beta and gamma irradiation dose for 30 days. NuScale has proposed to address the lack of the PAR design envelope, and qualification testing and analysis information in the FSAR by using the equipment survivability (ES) approach for the SDAA, including reliance on SDAA combined license (COL) Items 3.11-1 through 3.11-3, and COL Item 19.1-8. The ES approach addresses the SA dose only and does not address the unique PAR exposure to 60 years of neutron irradiation, dynamic effects such as jet impingement on the PAR, the effect of mechanical loads on the PAR during the mass and energy release phase, nor a post-accident environment of 30 days as stipulated for the harsh environment in the CNV. NuScales ES approach as described in FSAR chapter 19.2.3 is to compare the equipment qualification (EQ) doses with the severe accident doses. Another consideration is the potential for boron deposition as the CNV is repeatedly filled with borated water during refueling. None of the existing COL Items in Section 3.11 of the FSAR reflect or include the PAR design specifications or necessary qualification testing and analysis. COL Item 19.1-8, which requires to confirm that the key assumptions used in the PRA are reflected in the as-built, as-operated PRA, does not apply to the PAR because the PAR is currently not modeled in the probabilistic risk assessment (PRA). NuScale also states that procured equipment must meet its internal design requirements, citing its Quality Assurance Program Description (QAPD). The NRC staffs review determined that the QAPD applies to safety-related structures, systems, and components (SSCs) and three specific non-safety-related SSCs. The scope of the QAPD does not include the PAR. While NuScale cites Regulatory Guide 1.7, Control of Combustible Gas in Containment, Revision 3, as providing the augmented quality requirements for the PAR, the RG does not include any specificity on these requirements (e.g., codes and standards, qualification testing, analysis, etc.). An example of the specificity needed can be found in Table B-2 of the Technical Report, Treatment of DC Power in Safety Analyses, for the DC power system (EDAS). The Statement of Considerations for 10 CFR Part 52 (72FR49352) state that the information for the SDA needs to be equivalent to that for a DCA and that the information for a DCA must include performance requirements and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the NRC, and procurement specifications and construction and installation specifications by an applicant. The FSAR does not include information on the PAR consistent with the requirements in 10 CFR 52.137 (a)(2), (a)(4), and (a)(12). Specifically, the FSAR is lacking information about: (i) the design envelope for parameters that are key to achieving the functionality of the PAR during DBEs, including AOOs and SAs, including the 60-year neutron dose during normal operations, mechanical loads, dynamic effects from jet impingement, pressure, temperature, and humidity, and (ii) the qualification testing and analysis necessary to demonstrate that the key parameters, and

6 consequently, PAR functionality are achieved. The current FSAR only points to the existence of a PAR, which is insufficient for the NRC staff to make a safety finding not only for the PAR in Sections 6.2.5, 19.2.3, and the exemption in SDAA, Part 7, Section 2, but also for other safety-related SSCs (e.g., CNV, emergency core cooling system (ECCS)). This is necessary because NuScales analysis demonstrates that the PAR is needed to maintain the containment inert during DBEs, including AOOs. The current FSAR information is also insufficient for the staff to ensure that a COL applicant demonstrates that PAR functionality is achieved via qualification testing and analysis. Information Requested: NuScale is requested to provide SDA FSAR markups that provide:

1. The values or range of values for parameters that are key to assuring the functionality of the PAR during normal operations, DBEs, and SAs, including the 60-year neutron dose, temperature, pressure, humidity, mechanical loads, seismic categorization, and dynamic effects from jet impingement.

2. The technical basis for the selected values or range of values provided in item (1).

3. A new COL item or amendment to an existing COL item to demonstrate that the values or range of values for key parameters provided in item (1) are achieved through qualification testing and analysis.

4. Add the PAR to the scope of the QAPD, which is found in the topical report MN-122626, revision 1, NuScale Power LLC, Quality Assurance Program Description.

5. Confirm that the PAR is included in the FSAR Table 3.11-1, List of EQ Equipment Located in Harsh Environments. Justify any changes to the inclusion of the PAR in Table 3.11-1.

Question 19.2-2 Regulatory Basis: 10 CFR 52.137(a)(2) requires a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon which the requirements have been established, and the evaluations required to show that safety functions will be accomplished. 10 CFR 52.137(a)(4) requires an analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

7 10 CFR 52.137(a)(9) requires, for applications for light-water cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. 10 CFR 52.137(a)(12) requires an analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter. 10 CFR 52.137(a)(23) requires a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass. 10 CFR 52.137(a)(25) requires that the application must contain a final safety analysis report that describes the design-specific probabilistic risk assessment and its results. 10 CFR 52.137(b) requires, in part, an application for approval of a standard design, which differs significantly from the light-water reactor designs of plants that have been licensed and in commercial operation before April 18, 1989, or uses simplified, inherent, passive, or other innovative means to accomplish its safety functions, must meet the requirements of 10 CFR 50.43(e), as identified below. 10 CFR 50.43, Additional standards and provisions affecting class 103 licenses and certifications for commercial power, states, in part, the following: (e) Applications for a design certification, combined license, manufacturing license, or operating license that propose nuclear reactor designs which differ significantly from light-water reactor designs that were licensed before 1997, or use simplified, inherent, passive, or other innovative means to accomplish their safety functions, will be approved only if: (1)(i) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (ii) Interdependent effects among the safety features of the design are acceptable, as demonstrated by analysis, appropriate test programs, experience, or a combination thereof; and (iii) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions 10 CFR 50.44(c) Requirements for future water-cooled reactor applicants and licensees. The requirements in this paragraph apply to all water-cooled reactor construction permits or operating licenses under this part, and to all water-cooled reactor design approvals, design certifications, combined licenses or manufacturing licenses under part 52 of this chapter, any of which are issued after October 16, 2003. (1) Mixed atmosphere. All containments must have a capability for ensuring a mixed atmosphere during design-basis and significant beyond design-basis accidents.

8 (2) Combustible gas control. All containments must have an inerted atmosphere, or must limit hydrogen concentrations in containment during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100 percent fuel clad-coolant reaction, uniformly distributed, to less than 10 percent (by volume) and maintain containment structural integrity and appropriate accident mitigating features. (3) Equipment Survivability. Containments that do not rely upon an inerted atmosphere to control combustible gases must be able to establish and maintain safe shutdown and containment structural integrity with systems and components capable of performing their functions during and after exposure to the environmental conditions created by the burning of hydrogen. Environmental conditions caused by local detonations of hydrogen must also be included, unless such detonations can be shown unlikely to occur. The amount of hydrogen to be considered must be equivalent to that generated from a fuel clad-coolant reaction involving 100 percent of the fuel cladding surrounding the active fuel region. (4) Monitoring. (i) and (ii) Equipment must be provided for monitoring oxygen in containments that use an inerted atmosphere for combustible gas control. Equipment for monitoring oxygen and hydrogen must be functional, reliable, and capable of continuously measuring the concentration of oxygen in the containment atmosphere following a significant beyond design-basis accident for combustible gas control and accident management, including emergency planning. (5) Structural analysis. An applicant must perform an analysis that demonstrates containment structural integrity. This demonstration must use an analytical technique that is accepted by the NRC and include sufficient supporting justification to show that the technique describes the containment response to the structural loads involved. The analysis must address an accident that releases hydrogen generated from 100 percent fuel clad-coolant reaction accompanied by hydrogen burning. Systems necessary to ensure containment integrity must also be demonstrated to perform their function under these conditions. 10 CFR 50.12, Specific exemptions, Section (a). The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are-- (1) Authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. (2) The Commission will not consider granting an exemption unless special circumstances are present 10 CFR Part 50, Appendix A, General Design Criteria Criterion 1Quality standards and records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

9 Criterion 4Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. Criterion 41Containment atmosphere cleanup. Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. Criterion 42Inspection of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, to assure the integrity and capability of the systems. Criterion 43Testing of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural integrity of its components, (3) the operability of the systems as a whole and, under conditions as close to design as practical. Issue: To address the issue of combustible mixture within the reactor coolant system (RCS), NuScale has identified the ECCS valves (reactor vent valves) opening automatically at 8 hours after design basis events to prevent a combustible gas mixture in the RCS. These valves vent H2 and O2 into the CNV, promptly creating a combustible gas mixture in the CNV. NuScale then relies on the PAR with preventing combustion in the CNV, thereby protecting the CNV integrity and avoiding a direct path of radioactive release to the environment. NuScales supporting analysis for AOOs, events which are expected to occur at least once in the life of the plant, demonstrates release of a combustible mixture in the CNV within 24 hours and reliance on the PAR to maintain containment integrity. Based on its analysis on the presence and treatment of combustible gas in the RCS in (( }}, which includes an evaluation of the scenario described above, NuScale concludes, The results found that (( }} (emphasis added). The PAR is currently not modeled in the NuScale SDAA PRA based on the results of SA sequence simulations that demonstrate that a combustible mixture is not present in the

10 containment for 72 hours for those sequences. These are the so-called failure sequences in the PRA which lead to core damage. However, NuScales analyses demonstrate that the PAR is necessary for containment integrity during DBE scenarios. The DBE sequences represent the so-called success sequences in the PRA which do not lead to core damage. Therefore, failure of the PAR during success sequences would result in a direct release of radioactivity (i.e., the radioactive steam transferred from the RCS to the containment via the ECCS) to the environment. NuScale has not provided any quantitative evaluation that demonstrates a different conclusion. Consequently, the current PRA does not reflect the US460 design, and the PAR needs to be included in the SDAA PRA for the success sequences which do not go to core damage but can result in large release due to loss of containment integrity. Information Requested: To support the staffs finding against 10 CFR 52.137(a)(25) that the PRA represents the as-designed plant:

1. Include the PAR in the PRA for the NuScale SDAA and provide (i) FSAR markups with corresponding event trees, risk insights, dominant large release frequency sequences, and risk quantification results, and (ii) a discussion of the modeling of the PAR in the PRA, including the sequences that are impacted, the basis for the selected reliability, and any sensitivities performed.

2. Demonstrate that the human reliability analysis (HRA) and resulting human error probability (HEP) for the manual bypass of the 8-hour ECCS timer includes consideration of the need to confirm the critical hydrogen concentration in the RCS by the operators in addition to subcriticality at cold conditions. Include discussion of how the operators will make this determination identifying the instrumentation that will be used by the operators for confirmation and its reliability. If the HRA or the HEP needs to be changed, justify the new value, and provide FSAR markups resulting from HEP change.

3. Provide FSAR markups in Chapter 6 to include details related to when and how the manual bypass of the 8-hour ECCS timer will be performed by the operators to confirm critical hydrogen concentration in the RCS.

4. Provide a summary of NuScale report ((

}}, on the docket that describes (i) the representative SA scenarios selected for evaluation, (ii) the modeling of H2 from clad oxidation and O2 from radiolysis, and (iii) the results that demonstrate that for 72 hours, a combustible mixture in the CNV is not produced, including Table O-2, and Figures O-1 through O-13 from Appendix O of the report. Question 19.2-3 Regulatory Basis: 10 CFR 52.137(a)(2) requires a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon

11 which the requirements have been established, and the evaluations required to show that safety functions will be accomplished. 10 CFR 52.137(a)(4) An analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents. 10 CFR 52.137(a)(9) For applications for light-water cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. 10 CFR 52.137(a)(12) An analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter. 10 CFR 52.137(a)(23) a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass. 10 CFR 50.44(c) Requirements for future water-cooled reactor applicants and licensees. The requirements in this paragraph apply to all water-cooled reactor construction permits or operating licenses under this part, and to all water-cooled reactor design approvals, design certifications, combined licenses or manufacturing licenses under part 52 of this chapter, any of which are issued after October 16, 2003. (1) Mixed atmosphere. All containments must have a capability for ensuring a mixed atmosphere during design-basis and significant beyond design-basis accidents. (2) Combustible gas control. All containments must have an inerted atmosphere, or must limit hydrogen concentrations in containment during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100 percent fuel clad-coolant reaction, uniformly distributed, to less than 10 percent (by volume) and maintain containment structural integrity and appropriate accident mitigating features. (3) Equipment Survivability. Containments that do not rely upon an inerted atmosphere to control combustible gases must be able to establish and maintain safe shutdown and containment structural integrity with systems and components capable of performing their functions during and after exposure to the environmental conditions created by the burning of hydrogen. Environmental conditions caused by local detonations of hydrogen must also be included, unless such detonations can be shown unlikely to occur. The amount of hydrogen to be considered must be equivalent to that generated from a fuel clad-coolant reaction involving 100 percent of the fuel cladding surrounding the active fuel region.

12 (4) Monitoring. (i) and (ii) Equipment must be provided for monitoring oxygen in containments that use an inerted atmosphere for combustible gas control. Equipment for monitoring oxygen and hydrogen must be functional, reliable, and capable of continuously measuring the concentration of oxygen in the containment atmosphere following a significant beyond design-basis accident for combustible gas control and accident management, including emergency planning. (5) Structural analysis. An applicant must perform an analysis that demonstrates containment structural integrity. This demonstration must use an analytical technique that is accepted by the NRC and include sufficient supporting justification to show that the technique describes the containment response to the structural loads involved. The analysis must address an accident that releases hydrogen generated from 100 percent fuel clad-coolant reaction accompanied by hydrogen burning. Systems necessary to ensure containment integrity must also be demonstrated to perform their function under these conditions. 10 CFR 50.12, Specific exemptions, Section (a) The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are-- (1) Authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. (2) The Commission will not consider granting an exemption unless special circumstances are present 10 CFR Part 50, Appendix A, General Design Criteria Criterion 1Quality standards and records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Criterion 4Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. Criterion 41Containment atmosphere cleanup. Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for

13 offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. Criterion 42Inspection of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, to assure the integrity and capability of the systems. Criterion 43Testing of containment atmosphere cleanup systems. The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural integrity of its components, (3) the operability of the systems as a whole and, under conditions as close to design as practical. Issue: Based on its analysis on the presence and treatment of combustible gas in the RCS in (( }}, which includes an evaluation of an AOO, NuScale concludes, The results found that (( }} (emphasis added). Therefore, the PAR is necessary to prevent a combustible mixture in the CNV and maintain CNV integrity for DBEs, including AOOs. The CNV is a safety-related SSC, which is a key fission product barrier and reactor coolant pressure boundary, especially after ECCS activation for DBEs. The integrity of the CNV is essential for both preventing the release of radioactive material and for effective core cooling. Therefore, the staff believes that the current categorization of the PARs function as non-risk significant does not reflect its risk significance demonstrated by NuScales analyses. The PAR is currently designated as nonsafety-related. NuScale evaluated the PAR against the regulatory treatment of nonsafety systems (RTNSS) and determined that the PAR did not meet any of the five RTNSS criteria (A - E). Per the Design Reliability Assurance Program (D-RAP), an expert panel evaluated the PAR for risk significance and determined that the PAR did not perform any risk-significant functions, other than providing defense in depth for maintaining containment integrity following a SA. Consequently, the PAR is currently designated as non-risk significant. The SDAA FSAR, Revision 1, cites Regulatory Guide (RG) 1.7 for the augmented quality requirements for the PAR. Per SRP Chapter 17.4, Revision 1, Reliability Assurance Program, the Reliability Assurance Program (RAP) provides reasonable assurance of the following: The RAP SSCs do not degrade to an unacceptable level of reliability, availability, or condition during plant operations; These SSCs will function reliably when challenged; Quality assurance (QA) programs related to design and construction activities (e.g., design, procurement, fabrication, construction, inspection, and testing activities) to provide control over activities affecting the quality of the RAP SSCs.

14 Chapter 17.4 of the NuScale SDAA states that the implementation of the RAP provides reasonable assurance that, among other things, the plant is designed, constructed, and operated in a manner that is consistent with the risk insights and key assumptions (e.g., SSC design, reliability, and availability) from the probabilistic, deterministic, and other methods of analysis used to identify and quantify risk (emphasis added). Because the PAR is currently not identified as a risk significant SSC, it is not currently part of the SDAAs D-RAP. In accordance with the Staff Requirements Memorandum to the SECY-95-132, the staff verifies the future implementation of the D-RAP using the inspections, tests, analyses, and acceptance criteria (ITAAC) process. The ITAAC ensures that the design bases and other requirements have been correctly translated into the detailed design documents used for procurement and construction of every RAP SSC. The D-RAP ITAAC provides assurance to the staff that appropriate controls were imposed during the development of design products for RAP SSCs. SRP 14.3.11, Containment Systems ITAAC, contains the guidance for the ITAAC relevant to the CNV, including hydrogen generation and control. Currently, there is no ITAAC for the PAR. NuScales analyses performed in response to RAI-10081 6.3-1 demonstrates that the PAR is necessary to prevent a combustible mixture in the CNV and maintain CNV integrity for DBEs, including AOOs. The CNV is a safety-related fission product barrier and reactor coolant pressure boundary, especially after ECCS activation. The integrity of the CNV is essential for both preventing the release of radioactive material and for effective core cooling. Therefore, the PAR meets Criterion 3 of 10 CFR 50.36(c)(2)(ii)(c) because it is part of the primary success path and functions to mitigate a DBA or transient that presents a challenge to the integrity of a fission product barrier. A technical specification limiting condition for operation should be established for the PAR. The 2003 rulemaking for 10 CFR 50.44 was undertaken based on evaluations that demonstrated that combustible gases were not risk-significant for design basis events for large light-water reactors (LWRs). In contrast, as discussed above, NuScales analysis demonstrates that combustible gas mixture is risk significant for the US460 design. NuScale has not provided any quantitative evaluation that demonstrates a different conclusion. Based on the above, the PAR in Sections 6.2.5, 19.2.3, the exemption in SDAA, Part 7, Section 2, and for safety-related SSCs, such as the CNV, the staff believes that the PAR in the US460 design should be designated as a risk significant SSC and that regulatory controls are necessary to ensure that the PARs risk significant function is achieved through design, procurement, qualification testing and analyses, and during plant operations. Information Requested: To support the staffs safety findings on the PAR and safety-related SSCs, such as the CNV, against the regulatory bases identified above, NuScale is requested to provide following FSAR markups to reflect the PARs risk significant function demonstrated by NuScales analyses:

1. Identify the PAR as a risk significant SSC and add it to the SDAA D-RAP program.

2. Provide the specific augmented quality requirements for the PAR.

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3. Provide ITAAC(s) for the PAR.

4. Provide a Technical Specification for the PAR with justification for the corresponding action statement(s) and their completion time(s), and surveillance requirement(s).

Question 19.2-4 Regulatory Basis: 10 CFR 50.2, Definition, Safety-related structures, systems and components means those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 100.11 of this chapter, as applicable. 10 CFR 52.137(a)(2) requires a description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon which the requirements have been established, and the evaluations required to show that safety functions will be accomplished. 10 CFR 52.137(a)(4) An analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents. 10 CFR 52.137(a)(9) For applications for light-water cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. 10 CFR 52.137(a)(12) An analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter. 10 CFR 52.137(a)(23) a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass. Issue: 10 CFR 50.2 defines safety-related SSCs as those that are relied upon to remain functional during and following DBEs to assure: (1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (3)

16 The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 100.11 of this chapter, as applicable. Based on its analysis on the presence and treatment of combustible gas in the RCS in (( }}, which includes an evaluation of an AOO, NuScale concludes, The results found that (( }} (emphasis added). Therefore, the PAR is necessary to maintain the reactor in a safe shutdown condition following a DBE. Further, safety-related SSCs in the US460 design are not designed to withstand the dynamic effects and loading from a combustion event. Consequently, without the PAR, the containment integrity is not maintained and a direct path of radioactive release to the environment is available during DBEs. Due to ECCS operation during such events and the release of radioactive steam from the RCS to the CNV, loss of containment integrity will result in a radioactive release to the environment. The 2003 rulemaking for 10 CFR 50.44 was undertaken based on evaluations that demonstrated that combustible gas releases were not risk-significant for design basis events for large light-water reactors (LWRs). In contrast, as discussed above, NuScales analysis demonstrates that combustible gas mixture is risk significant for the US460 design. NuScale has not provided any quantitative evaluation that demonstrates a different conclusion. Therefore, the PAR is relied upon to remain functional during DBEs to maintain the US460 design in safe shut down condition and to prevent the consequences of accidents which could result in potential offsite exposures exceeding applicable regulatory limits. Information Requested: NuScale is requested to identify the PAR as a safety-related SSC or justify why the PAR can remain classified as non-safety related. If justification is provided, it should address: (1) the necessity of the PARs function to maintain containment integrity and safe shutdown conditions during DBEs, (2) the ability of safety-related SSCs to continue to perform their function under dynamic effects and loading from a combustion event with the frequency of an AOO, and (3) the difference in the quality and performance of a PAR that is designated as safety-related and one that is not (see Information Requested on risk significance of the PAR).}}