ML25303A305

Chapter 6 - Us NRC Draft Safety Evaluation Related to the U.S. Sfr Owner, LLC Construction Permit Application for the Kemmerer Power Station Unit 1

ML25303A305
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	11/06/2025
From:	NRC/NRR/DANU
To:
References
Download: ML25303A305 (1)

Text

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.

6-1 6

SAFETY-SIGNIFICANT SSC CRITERIA AND CAPABILITIES The Kemmerer Unit 1 (KU1) PSAR chapter 6 includes both safety-related (SR) and non-safety related with special treatment (NSRST) structures, systems, and components (SSC) criteria and capabilities. This KU1 PSAR includes elements that address Regulatory Guide (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 Non-Light-Water Reactors, Revision 0 (ML23269A222), and DANU-ISG-2022-01 Review of Risk-Informed, Technology-Inclusive Advanced Reactor ApplicationsRoadmap, March 2024 (ML23277A139).

6.1 Design Requirements of Safety-Related SSCs 6.1.1 Introduction KU1 PSAR section 6.1 includes design requirements for SR SSCs and includes a discussion of the design basis hazard levels (DBHLs) in PSAR section 6.1.1, a discussion of the approach to the safety-related design criteria (SRDC) in PSAR section 6.1.2, and a summary of DBHL-related requirements for non-safety-related SSCs in PSAR section 6.1.3.

6.1.2 Regulatory Evaluation The regulatory requirements applicable to the staffs review of PSAR section 6.1 are:

Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a),

10 CFR 50.35, 10 CFR 50.150(a)(1),

10 CFR Part 50, Appendix S, and 10 CFR 100.23.

Applicable guidance for the staffs review of PSAR section 6.1 includes:

DANU-ISG-2022-01, March 2024 RG 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants, Revision 1 (ML070360253)

RG 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Revision 2 (ML21253A071)

RG 1.91 Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants, Revision 3 (ML21260A242)

RG 1.189, Fire Protection for Nuclear Power Plants, Revision 5 (ML23214A287)

RG 1.204, Guidelines for Lightning Protection for Production and Utilization Facilities, Revision 1 (ML23241A957)

RG 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, Revision 0 (ML20091L698), which endorses with clarifications, the guidance in NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1 (ML19241A472)

6-2 RG 1.253, Revision 0, which endorses with clarifications and additions NEI 21-07, Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report:

For Applicants Utilizing NEI 18-04 Methodology, Revision 1 (ML22060A190)

The principal design criteria (PDC) as defined in PSAR section 5.3 and evaluated by the staff in the corresponding section of this safety evaluation (SE), that apply to the review of PSAR section 6.1 are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 81, Reactor Building Design Basis 6.1.3 Technical Evaluation 6.1.3.1 Design Basis Hazard Levels The staff reviewed PSAR section 6.1.1, which discusses the DBHLs for SSCs designed to withstand the effects of hazards that are external to the plant and internal plant hazards, ensuring no adverse impact on their ability to perform intended safety functions. The applicant stated that DBHL-related requirements for NSRST and NST SSCs are based on preventing interactions with SR SSCs. The design parameters for DBHLs of external hazards, including wind and tornado, external flood, missile protection, seismic hazards, and extreme winter precipitation, are presented in KU1 PSAR table 2.1-1. The applicants justification for these parameters is evaluated in section 2.1 of this SE.

The staff performed its evaluation of DBHLs in PSAR section 6.1.1, using the guidance in RG 1.233 and RG 1.253. Specifically, NEI 21-07, endorsed with clarifications and additions in RG 1.253, states that CP application content for section 6.1.1 should be as complete as possible, based on the site characterization required in 10 CFR 50.34(a). NEI 21-07 further states that DBHLs extend beyond environmental hazards external to the plant and include both external and internal hazards, such as seismic events, wind (including tornadoes and wind-generated missiles), external flooding, hazards from external facilities, internal fires, internal floods, high-energy line breaks, and internally generated missiles.

USO selected a set of DBHLs as part of its design bases that the staff reviewed and for the reasons described below determined were consistent with RG 1.233, including an update to the PSAR to clarify that DBHLs are selected in accordance with approved methods and consistent with section 6.1.1 and table 6-1 as described in NEI 21-07, as well as certain subsections of chapters 3.2 through 3.7 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (ML16084A812, ML070570001, ML070570002, ML070550043, ML070570003, ML070370569, ML070380167, ML063600395, ML14190A180, ML070510635, ML100331298, ML070460362, ML070570004, ML070550032, ML16088A041, ML14198A460, ML13198A223). USO stated that a hazards screening analysis was performed for all applicable hazards and the list of DBHLs will be re-assessed and updated at the OL stage.

The staff reviewed USOs approach for identifying, selecting, and assessing DBHLs and, for the reasons described below, the staff determined it was reasonable and that the DBHLs

6-3 preliminarily identified for the CPA provide a sufficient basis for hazard evaluation. Detailed staff evaluations of individual hazards are provided in the following subsections.

High Winds The staff reviewed the applicants assessment of high wind hazards and the site-specific high wind parameters provided in PSAR section 2.4.1.3.2. The applicant stated that for the structural design of safety-significant SSCs, wind loading conditions are based on 3-second gust wind speeds of 110 and 115 miles per hour, in accordance with American Society of Civil Engineers/Structural Engineering Institute (ASCE/SEI) Standard 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures. The applicants evaluation was based on local historical data for the KU1 site area, as documented in PSAR section 2.4.1.3.2 and table 2.1-1. The staff determined that the applicant provided sufficient information to characterize the high wind hazard at the KU1 site and to support the design basis high wind hazard parameters, including the selected parameters to be reasonable for use in the KU1 design basis, as documented in section 2.4.2.1.5 of the SE.

A more detailed evaluation of the adequacy of the SR, NSRST, and NST SSCs to withstand high wind hazards without adverse impact on their safety functions is provided in section 6.1.3.3 and chapter 7 of this SE.

Tornadoes The staff reviewed the applicants assessment of tornado hazards and the site-specific tornado parameters described in KU1 PSAR section 2.4.1.3.3. The applicant stated that the design-basis tornado characteristics applicable to safety-significant SSCs are consistent with RG 1.76, Revision 1 including: maximum wind speed, translational speed, maximum rotational speed, radius of maximum rotational speed, pressure drop, and rate of pressure drop.

Based on local historical data for the KU1 site provided in PSAR section 2.4.1.3.3 and table 2.1-1, the staff determined that the applicant has provided sufficient information to characterize tornado hazard and to support the design basis tornado parameters, as documented in section 2.4.2.1.6 of the SE.

A more detailed evaluation of the ability of SR, NSRST, and NST SSCs to perform their required safety functions under tornado hazard conditions is provided in section 6.1.3.3 and chapter 7 of this SE.

External Flooding PSAR section 2.5.1 provides potential flooding information for the vicinity of KU1 site and summarizes the types and combinations of flood-producing phenomena considered in establishing the design-basis flood. The postulated flood scenarios include:

Rivers and streams Dam breaches Surges and seiches Tsunami

6-4 Snow melts and ice jams Channel diversion.

Additionally, the impacts of local intense precipitation (LIP) are described in PSAR section 2.5.1.

The staffs review of the external flooding hazard and evaluation of the site-specific external flood parameters are provided in section 2 of this SE. As described in that section, based on the estimated flood data for the KU1 site provided in PSAR section 2.5.1 and table 2.1-1, the staff determined that USO provided sufficient information to characterize the external flood hazard at the proposed site and to support its evaluation of the design-basis external flood hazard.

The staff notes that USO clarified that flooding of SR SSCs from external sources is mitigated by minimizing potential water intrusion to below-grade SR SSCs and by designing NSRST SSCs surface structure concrete levels and exterior doors above the water levels corresponding to the design-basis flood and local precipitation. Further evaluation of KU1 SR, NSRST, and NST SSCs demonstrate that the external flood hazard would not adversely affect their ability to perform their required safety-related functions, is documented in section 6.1.3.3 of this SE.

In summary, the staff reviewed the information in the KU1 PSAR and determined that the applicant provided sufficient information at the CP stage to characterize the external flooding hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

Internal Flooding PSAR section 6.1.1.3 states that, The design basis internal flooding analysis will consider the single postulated fluid system failure that results in the most limiting flood conditions in each area that contains safety-significant SSCs. The PSAR states that the analysis considers sodium, molten salt, oil, and water systems resulting from the following postulated failures:

Selected pipe failures in seismically-qualified fluid systems, as addressed in PSAR section 6.1.1.5; Circumferential breaks of piping systems not designed to retain function following seismic events (i.e., not designed to SCS1, SCS2, or SCN1 seismic standards) and piping systems not designed to be protected from the tornado DBHL; Rupture of tanks not designed to retain function following seismic events (i.e., not designed to SCS1, SCS2, or SCN1 seismic standards) and tanks not designed to be protected from the tornado DBHL; Communication from other areas, including communication through the floor drain system; Operation of the fire protection system; Flooding due to impact of postulated missiles on fluid systems (as addressed in PSAR section 6.1.1.4.1); and Flooding due to postulated heavy load drops.

6-5 In addition to the information in the PSAR, the staff conducted an audit that included supporting documentation to confirm the information in PSAR section 6.1.1.3.

PSAR section 6.1.1.3 identifies potential design features to protect safety-significant SSCs from postulated flooding, including use of isolation valves, physical barriers, and drains. The PSAR also notes that systems in the Energy Island (EI) (e.g., water, salt, and oil systems) are physically separate and at a lower elevation than the safety-significant SSCs, which are located in the NI structures.

At the preliminary design stage evaluated in the PSAR, the seismically-qualified fluid systems consist primarily of sodium systems. Due to the reactive nature of sodium, internal flooding from sodium pipe leaks is largely accounted for in the sodium leak fire protection strategy. PSAR section 5.3.1.4 states that sodium-containing SSCs external to the reactor vessel are provided with measures such as guard pipes, jackets, location within lined cells, or catch pans to collect leakage from moderate energy pipe cracks in seismically-qualified SSCs.

Based on the PSAR descriptions and confirmation via the audit, the staff determined that USO has sufficiently addressed how the design-basis internal flooding hazard will be established and necessary protective design features will be implemented.

The staff determined that the internal flooding hazard at KU1 has been reasonably characterized, and design measures are analyzed and considered to ensure that SR SSCs can maintain their safety-significant functions under postulated internal flood conditions, based on the preliminary facility design. Necessary flood protection design features will be evaluated at the OL stage, should the applicant apply for an OL.

In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the internal flooding hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

Turbine Missile The KU1 turbine missile hazard is assessed and discussed in PSAR section 2, table 2.1-2, and section 6.1.1.4. PSAR table 2.1-2 indicates that the turbine generator is located on the energy island within the turbine facility building, approximately 500 ft south of the nuclear island, with its rotor oriented in the north-south direction. Based on this location and orientation, USO concluded that the potential for damage from turbine-generated missiles to safety-significant SSCs is insignificant.

The staffs review of the turbine missile hazard and its potential impact on KU1 SSCs is provided in chapter 2 of this SE. In that section, the staff determined that USO has provided sufficient information to characterize the turbine missile hazard and to support its evaluation of the design-basis turbine missile hazard.

Additionally, based on the staffs review of the preliminary design and layout of the turbine facility, the staff determined that the rotor orientation and the separation distance between the

6-6 turbine and the nuclear island are sufficient to ensure that safety-significant SSCs are protected from potential turbine missile impacts.

In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the turbine missile hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

Internally Generated Missile The hazard associated with missiles generated from rotating equipment is assessed in PSAR section 6.1.1.4. For missiles potentially generated from pressurized systems and components, including tanks and cylinders containing compressed gas, USO states in the PSAR that a probability analysis is being performed. The applicant states in PSAR 6.1.1.4.1, if the probability of a missile occurrence, impacting a SR target and causing an adverse effect on its ability to perform an SR function, is less than 1x10 per year, the missile is not considered a design-basis missile.

As described in PSAR section 6.1.1.4.2, SR SSCs are either protected from or designed to withstand impacts from a single design-basis missile without loss of capability to perform their required safety functions. Design-basis missiles are addressed using one or more of the following methods:

Physically separating or orienting potential design-basis missile sources to prevent impacts on SR SSCs.

Demonstrating that a missile impact does not adversely affect the ability of SR SSCs to perform their SR functions.

Providing missile barriers. When missile barriers are employed, the design procedure for the barrier will be provided at the OL stage.

For the design-basis missile assessment, USO considered missiles generated from pressurized systems and components, turbine missiles, rotating equipment, secondary missiles generated from primary missile impacts, wind-and tornado-generated missiles, and site-proximity missiles.

The staff evaluated the descriptions of the internally generated missiles provided in PSAR section 6.1.1.4, and confirmed via audit the statements in the PSAR were accurate, and determined that USO has provided sufficient information to characterize internally generated missile hazards and to support its evaluation of the design-basis internally generated missile hazard at the CP stage. Further evaluation of the KU1 SR, NSRST, and NST SSCs to demonstrate that missile hazards generated from rotating equipment and pressurized systems will not adversely affect their ability to perform required safety-related functions can reasonably be left until the OL stage since appropriate design provisions are in place for the related safety significant SSCs.

In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the internal generated missile hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

6-7 Rupture of Piping PDC 4 requires safety-significant SSCs to be appropriately protected against dynamic effects, including those resulting from missiles, pipe whipping, and discharging fluids caused by equipment failures or external events and conditions outside the nuclear power unit. PSAR section 6.1.1.5 states that high-energy liquid piping systems are not planned for installation in safety-significant structures.

Due to the physical separation between the energy island and the nuclear island, the applicant concluded that the effects from pipe ruptures in fluid systems located in the energy island do not affect SSCs in the nuclear island. Circumferential ruptures and longitudinal splits are not postulated in moderate-energy piping.

PSAR section 6.1.1.5 states that SR SSCs are designed such that design-basis pipe ruptures will have no adverse impact on their capability to perform required SR functions. Design-basis pipe ruptures are addressed by one or more of the following methods:

Physically separating piping systems subject to postulated design-basis pipe ruptures from SR SSCs where feasible.

Providing guard piping, barriers, or fluid collection systems to mitigate the effects of design-basis pipe ruptures on SR SSCs.

Demonstrating that design-basis pipe ruptures do not adversely impact the ability of SR SSCs to perform their SR functions.

In summary, the staffs review determined that the applicant has provided sufficient information to characterize the pipe rupture hazard and to support its evaluation of the design-basis pipe rupture hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07. Further evaluation of the KU1 SR, NSRST, and NST SSCs to demonstrate that the pipe rupture hazard would not adversely affect their ability to perform required safety-related functions, will be provided and reviewed by the staff at the OL stage, should the applicant apply for an OL.

Seismic The seismic hazard at the KU1 site is addressed in several sections of the PSAR, with the detailed discussions provided in PSAR section 2.6 and the associated seismic design described in PSAR section 6.4.1. In addition to the PSAR, the staff conducted an audit that included supporting documents that define plant-level requirements to meet design-basis hazards, including seismic, to confirm information in PSAR sections 2.6 and 6.4.1.

The staffs review of the KU1 seismic hazard and evaluation of the potential impact of a seismic event on SSCs are provided in sections 2.6, 6.1.3.3, 6.4, and 7.8 of this SE. In those sections, the staff determined that USO has provided sufficient information to characterize the design-basis seismic hazard and to support its evaluation of how the design addresses the hazard.

Although not explicitly considered in the DBHLs, sufficient seismic design margin will need to be demonstrated at the OL stage to meet the LMP risk target for beyond-design-basis events (BDBEs). The staff evaluation of seismic DBHLs is documented in SE section 2.6.2, and the seismic design evaluation is provided in SE section 6.4.

6-8 In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the seismic hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

Extreme Winter Precipitation The KU1 design-basis extreme winter precipitation hazard, along with its parameters and supporting bases, is described in PSAR section 2.4.1.3.5 and table 2.1-1. Loads resulting from extreme winter precipitation have been considered in the design of SSCs at KU1. The overall mitigation strategy is to protect safety-significant SSCs by locating them within appropriately designed buildings. SSCs that cannot be fully enclosed are designed to withstand extreme winter precipitation at levels commensurate with their safety and risk significance.

In summary, the staff reviewed the KU1 extreme winter precipitation hazard and its potential impact on SSCs provided in section 2 of the PSAR and determined that the applicant provided sufficient information at the CP stage to characterize the extreme winter precipitation hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07. Further evaluation of the KU1 SR, NSRST, and NST SSCs to demonstrate that the extreme winter precipitation hazard would not adversely affect their ability to perform required safety-related functions, will be provided and reviewed by the staff at the OL stage, should the applicant apply for an OL.

Internal Fire As described in section 6.1.1.8 of the PSAR, internal fire is a hazard that will be fully evaluated at the OL stage. During the OL stage, fire protection features will be developed or refined to satisfy DBA analysis assumptions and internal fire PRA results, or DBA analysis assumptions will be updated as necessary based on the evolving design. However, at the CP stage, analyses of DBAs that could lead to sodium fires established design requirements to maintain DBA analysis and ensure that SSCs required to mitigate the DBAs are available.

As described in KU1 PSAR section 5.3.1.3, the NI will be designed in accordance with RG 1.189. Area-specific reviews are being performed by the applicant to establish design criteria for internal hazards and SSC interactions. Future fire hazard analyses and modeling will define fire areas and establish appropriate design features to suppress fires within the area and prevent fire propagation beyond the area. These design features include sodium leak detection, containment, inert environments, equipment material selection to reduce combustible quantities, fire detection, and both automatic and manual dry-gas and water suppression systems.

The staff determined that USO has provided sufficient information to characterize the internal fire hazard at the CP stage. Section 7.5.2 of the PSAR provides additional information on the preliminary design of fire protection system and section 8.3 of this SE provides additional information on the fire protection program used to manage these hazards. Further evaluation of the KU1 SR, NSRST, and NST SSCs, to demonstrate that the internal fire hazard will not adversely affect their ability to perform the required safety-related functions, will be provided and reviewed by the staff at the OL stage, should the applicant apply for an OL.

6-9 In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the internal fire hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

External Fire USO described external fire hazards in section 2.3.3.1.7 of the PSAR. The staff evaluated fire hazards in section 2.3.7 of the SE, where it was concluded that wildfires do not pose a significant hazard in the region surrounding the KU1 site and the further evaluation of potential fires from the sodium test and fill facility (TFF) and energy island could be left to the OL stage of the review, should the applicant apply for an OL. Further, as identified in table 3.1-2 of the PSAR, USO qualitatively screened out the external fire hazard. The staffs evaluation of table 3.1-2 determined that the applicant should reassess the external fire events to ensure conformance with the non-LWR PRA standard and RG 1.247 at the OL stage, should the applicant apply for an OL. Additional considerations of the external fire hazard are discussed below.

Section 5.3.2.10 describes the PDC related to control room habitability and the potential impact of external smoke on the habitability of the main control room (MCR). Additionally, PSAR section 7.5.1, 7.6.7, and 7.8.4 describe design considerations for the MCR. Collectively in these sections, USO addressed the design of the control room envelope and its associated heating, ventilation, and air conditioning (HVAC) system. The control room envelope is designed to maintain isolation capability in the event of an external fire and associated smoke.

As described in table 1.4-1 of the PSAR, the KU1 design fully conforms to RG 1.189 which ensures control room habitability, 72-hour remote shutdown complex (RSC) habitability, and proper interface with the nuclear island fire protection (NFP) system to maintain HVAC operation during fire events.

Based on comparison of RG 1.189 guidance and the preliminary design described in the PSAR, the staff determined that USO has provided sufficient information to characterize the external fire hazard at the CP stage. Further evaluation of the KU1 SR, NSRST, and NST SSCs, to demonstrate that the external fire hazard will not adversely affect their ability to perform required safety-related functions, will be provided and reviewed by the staff at the OL stage, should the applicant apply for an OL.

As described above, the staff determines that the external fire hazard has been reasonably characterized, and the site layout and design features provide assurance that safety-significant SSCs can continue to perform their required functions in the event of an external fire.

In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the external fire hazard consistent with the guidance in RG 1.253, which endorses NEI 21-07.

6-10 Transportation The KU1 PSAR provides a discussion of nearby transportation routes and incident-related hazards in chapters 2 and 7. PSAR section 2.3 describes the assessment of hazards associated with nearby industrial, transportation, and military facilities and included consideration for waterway, railway, highways, and pipelines in accordance with RG 1.78 and RG 1.91. As described in PSAR section 2.3.3.1.4, none of the examined transportation hazards present a design basis hazard to the KU1 site. The staff conducted an audit that included supporting documentation of transportation hazards and confirmed consistency the discussion in PSAR section 2.3 and with guidance in NEI 21-07.

Each transportation mode has been evaluated to identify chemicals potentially transported along these routes that could constitute a design basis event. The analysis established minimum safe distances for postulated accidents, which were then compared to the proximity of these transportation routes to KU1. The staff conducted an audit that included the supporting detailed analysis of transportation hazards, including the potential chemical release of propane to confirm the descriptions of the hazards in the PSAR.

In summary, the staff reviewed the information in the KU1 PSAR, confirmed it through an audit of the supporting documentation, and determined that the applicant provided sufficient information at the CP stage to characterize the transportation hazards consistent with the guidance in RG 1.253, which endorses NEI 21-07, and the analyses provide assurance that safety-significant SSCs can perform their required functions in the event of a transportation accident..

Volcanic Hazards Volcanic hazards in the vicinity of the KU1 site are assessed and discussed in PSAR section 2.7. The KU1 values of the site characteristics of these hazards are presented in PSAR table 2.1-2. Based on its assessment of volcanic phenomena in NAT-3226-A, An Analysis of Potential Volcanic Hazards at the Proposed Natrium Site near Kemmerer, Wyoming, Revision 0A (ML24303A409), USO concluded that the site is located at sufficient distances from potential future volcanic eruption sources to preclude the following hazard phenomena from impacting the site:

The potential for the opening of a new volcanic vent Proximal hazards Lava flows Pyroclastic density flows.

Tephra falls, including volcanic ash and debris flows, were assessed and described in sections 8 and 9 of NAT-3226-A.

The staff separately reviewed the USO characterization of tephra fall hazards in NAT-3326-A and its conclusions in the safety evaluation included with the approved version of the TR as documented in ML24198A093 are incorporated in this SE and referenced in section 2.7 of this SE.

6-11 Thunderstorm and Lightning Hazards Thunderstorm and lightning strike hazards at the KU1 site are assessed and discussed in PSAR chapters 1 and 2. Specifically, PSAR section 1.1.4.3.11 describes that the Grounding, Earthing, and Lightning Protection (NGL) system provides protection for plant personnel and equipment from transient over-voltages that can result from electrical faults or lightning strikes. The applicant states the system also provides a proper ground reference for instrumentation signals.

PSAR section 1.1.4.3.11 states the nuclear island lightning protection system is designed to protect exposed buildings, structures, stacks, substations, and associated electrical and electronic circuits from hazards associated with transient over-voltages due to lightning strikes and switching surges.

Additionally, as documented in PSAR table 1.4-1, USO committed to fully conform to RG 1.204.

Based on its review by comparing the PSAR information on thunderstorm an lightning hazards to the guidance in NEI 21-07 and RG 1.204, the staff determined that USO has provided sufficient information to characterize the thunderstorm and lightning hazards. The staff concludes that the thunderstorm and lightning hazards have been reasonably characterized at the CP stage, and the proposed design and mitigation measures provide assurance that safety-significant SSCs can perform their required functions during thunderstorm and lightning hazards.

Aircraft Impact Accident PSAR section 3.1 table 3.1-2 indicates aircraft impact accidents are quantitively screened out of the PRA. However, the staff evaluation of USOs aircraft impact accident analysis and the associated effects on the facility to demonstrate compliance with 10 CFR 50.150(a)(1), are described in SE section 11.4.

6.1.3.2 Safety Related Design Criteria PSAR section 6.1.2 states that the safety-related (SR) design criteria (SRDC) are provided in PSAR tables 5.2-1, 5.2.-2 and 5.2-3 to show their connection to the SR functions. SR functions are described in PSAR section 5.2.1 and section 5.4 of this SE. PSAR table 5.2-1 lists the SR functions supporting control of heat generation and the associated SRDC, table 5.2-2 lists SR functions supporting control of heat removal and the associated SRDC, and table 5.2-3 lists SR functions supporting retaining radionuclides and the associated SRDC. The staff evaluation of the SR function and the associated preliminary SRDC is provided in chapter 7 of this SE, in the section(s) the SRDC applies to. Table 6.1-1 of this SE provides the SRDCs and their corresponding systems and sections within the PSAR and the SE.

6.1.3.3 DBHL-related Requirements for Non-safety Related SSCs In PSAR section 6.1.3, the applicant describes the design requirements for NSRST and NST SSCs such that the DBHLs will not cause an NSRST or NST SSC to adversely impact the ability of an SR SSC to perform an SR function. The applicant states that the design of the facility will prevent DBHL-related interactions (seismic and tornado) between NSRST or NST SSCs and SR SSCs. The focus of the staffs review of PSAR section 6.1.3 was the potential for seismic or tornado missile interactions between SR and non-safety related SSCs. Seismic design of

6-12 NSRST SSC to perform safety-significant functions and protection of NSRST SSCs from seismic interaction are covered in PSAR section 6.4.1 and evaluated by the NRC staff in section 6.4.3.1 of this SE.

PSAR section 6.1.3.1 describes how potential seismic interactions are identified through an area-based review process that treats safety-significant SSCs as seismic interaction targets and NSRST and NST SSCs as potential seismic interaction sources. USO states that the source SSCs will be designed to ensure that interactions will not lead to adverse consequences and will meet the LMP risk target (including in the BDBE region), which will be demonstrated at the OL stage by the seismic PRA (SPRA). PSAR section 6.1.3.1 states credible seismic interactions are identified if the target is within a specified physical zone of influence around the source.

USO states that if physical rearrangements and barriers cannot prevent interactions, then seismic interaction prevention design requirements are applied.

USO stated that to prevent seismic-related DBHL interactions, NSRST and NST SSCs with seismic interaction prevention design requirements will be designed in accordance with ASCE/SEI 7-16 at the SCN1 seismic classification level. The parameters for SCN1 are provided in PSAR table 6.4-1B and defined in PSAR section 6.4.1.1. While ASCE/SEI 7-16 is the code of record, the applicant performed a supplemental evaluation is performed to ensure that source SSCs are designed with a performance level greater than or equal to that of the target SSC and with inelastic deformation at or below that corresponding to Limit State B in ASCE/SEI 43-19.

PSAR section 6.1.3.1 states that if the Limit State B capacity is not initially met, then additional design requirements, beyond the base ASCE/SEI 7-16 requirements, are added for these SSCs until they meet ASCE/SEI 43-19 Limit State B in this iterative design process.

The PSAR states that these additional design requirements could include, for example: limiting structural systems to those endorsed by ASCE/SEI 43-19, including ductile detailing in design, or limiting ASCE/SEI 7-16 Response Modification Coefficients (R and Rp) based on ASCE/SEI 43-19 inelastic energy absorption factors (Fµ) for Limit State B. The applicant will also evaluate the seismic separation between NSRST or NST structures and adjacent SR SSCs using the methods in ASCE/SEI 43-19 section 7.3. This methodology contributes to meeting the requirements of PDC 81, providing a methodology to prevent the RXB superstructure from impacting the RAC and its ability to passively remove residual heat from the RES.

In addition to the physical interactions between sources and targets, USO will also treat sodium-containing SSCs as special seismic interaction source in their analysis of seismic interactions because of the potential for seismic-induced fires. Sodium-containing SSCs will be designed to retain their sodium inventory following the SSE. This seismic performance requirement for sodium retention will be achieved either through their design (a performance level equivalent to SR SSCs) or alternative means. Additional details on sodium retention seismic interaction design requirements are expected at the OL stage, should the applicant apply for an OL.

With seismic interaction prevention requirements being applied separately from SSC classification, there could be NST SSCs whose failure could lead to an event with consequences exceeding the F-C target curve based on seismic interaction with an SSC performing a safety-significant function. To address the safety concerns associated with this, seismic interaction sources have design, analysis, and appropriate programmatic controls applied at a level consistent with NSRST SSCs. Programs applied to SSCs with seismic

6-13 interaction prevention requirements include the Quality Assurance Program Description (QAPD), the Post-Construction Inspection, Testing, and Analysis Program (PITAP), and the Natrium Maintenance Program.

For tornado DBHL interaction (PSAR section 6.1.3.2), the applicant will design NSRST/NST SSCs to the provisions of ASCE/SEI 7-16, and the tornado and tornado missile parameters specified in RG 1.76 (PSAR table 1.4-1). Specifically, these parameters include design basis tornado hazard wind speed, atmospheric pressure drops, and tornado missiles for the design of NSRST/NST SSCs as presented in PSAR table 2.1-1. As discussed in PSAR section 7.8.2.2.3, USO has applied additional design requirements from American National Standards Institute/American Institute of Steel Construction (ANSI/AISC) N690, Specification for Safety-Related Steel Structures for Nuclear Facilities and American Concrete Institute (ACI) 349, Nuclear Safety-Related Concrete Structures, which ensures that the SSCs would not result in any adverse interactions with SR SSCs.

The staff finds USOs DBHL-related requirements with regard to the seismic and tornado interaction for non-safety-related SSCs that could have adverse interactions with SR SSCs acceptable for the CP application based on the following:

The process of identifying target and source pairs based on a physical zone of influence appears to be a systematic and thorough method for identifying SSCs needing seismic interaction prevention requirements to ensure their failure will not prevent a safety-significant SSC from performing its PRA safety function.

The application of ASCE/SEI 7-16, along with the supplemental evaluation to ensure that inelastic deformation will not exceed what is allowed under ASCE/SEI 43-19 Limit State B, and confirmation of adequate separations using ASCE/SEI 43-19 section 7.3 methods provides assurance at the CP stage that seismic interaction will not impact safety-significant SSCs for design basis loading. This design approach, including additional design requirements (i.e., to meet ASCE/SEI 43-19 Limit State B), should be confirmed through the seismic PRA by the applicant at the OL stage.

Appropriate programmatic control, including QAPD requirements consistent with NSRST SSCs, have been applied to ensure that the critical characteristics associated with the seismic interaction prevention requirements are maintained from design through construction and operation.

The design approach for the NSRST/NST SSCs for tornado interaction is adequate for the preliminary design because USO proposed high wind and tornado site characteristics, which are listed in PSAR table 2.1-1, based on those in table 1 (region 3) of RG 1.76. The staffs review is in section 2.4.2.4.4 of this SE.

6.1.4 Conclusion The staff reviewed the applicants information on the design-basis hazard levels for safety-related SSCs and concludes that the applicants information is sufficient to meet the requirements of PDC 2 and 10 CFR Part 50, Appendix S. Specifically, the applicant has demonstrated that nuclear power plant SSCs important to safety will be designed to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety functions. The information summarized in PSAR section 6.1 is sufficient for USO to

6-14 define the design requirements for safety-related SSCs, the DBHLs, the associated safety-related design criteria (SRDC), and DBHL requirements for non-safety-related SSCs.

1. Regarding the external hazards, USO assessed and developed DBHL parameters for high winds and tornadoes, external flooding, missile hazard, seismic hazard, extreme winter precipitation, volcanic hazard, transportation hazard, thunderstorms and lighting, external fire, and aircraft impact as described in multiple chapters of the PSAR. In addition to the staffs evaluation provided in this section, the staffs review of these hazards is also documented in chapter 2 of this SE. The staff determines that USO provided sufficient information to characterize each hazard at the CP stage and that the applicants approach for preventing adverse interactions with SSCs performing safety-significant functions is reasonable.

2. Regarding the internal flood hazard, USO stated that the SR SSCs are designed to withstand the design basis inernal flooding with no adverse impact on their capability to perform their SR functions. Internal flooding is a hazard that will be assessed at the OL stage. The staff determined that it is reasonable to address flooding hazards from design basis pipe rupture with physical separation of safety-significant SSCs, guard piping, protective barriers, and fluid collection systems, or by demonstrating that safety-significant SSCs are not adversely affected should flooding occur.

3. Regarding the internal fire hazard, USO stated that the design basis internal fire hazard will be evaluated in accordance with RG 1.189, Rev. 5, including potential fire impacts from sodium piping leakage or rupture, at the OL stage. The staff determined that the applicants approach to internal fire hazard is consistent with RG 1.253 guidance.

4. Regarding the design basis missile hazard, USOs assessment of tornado-generated missile hazards, documented in PSAR section 2.4.1.3.3, is reasonable. USO provided sufficient information to characterize the missile hazard at the KU1 site, including consideration of interactions between NSRST/NST and SR SSCs.

5. Regarding the seismic and tornado interactions, the staff reviewed USOs approach for addressing seismic and tornado interaction effects for non-safety-related SSCs. The staff determines, based on the evaluation above in section 6.1.3.3 of this SE, that the applicants approach provides assurance at the CP stage that potential interactions during the SSE and tornado DBHL will be adequately addressed. Seismic design requirements for SR and NSRST SSCs to ensure performance of safety-significant functions are documented in section 6.4.3 of this SE.

Based on the evaluations described in the previous subsections, the staff concludes that USO has reasonably characterized the DBHLs and that its proposed design approach provides assurance that SR and NSRST SSCs will withstand the effects of natural and man-made hazards without loss of safety function. Accordingly, the staff finds that the information in the PSAR satisfies the applicable requirements of 10 CFR 50.34(a) and 10 CFR 50.35 for issuance of a construction permit. The staffs evaluation of the associated preliminary design requirements is provided in section 6.4.3 of this SE.

6-15 Table 6.1-1. Safety-Related Design Criteria from the KU1 PSAR and Associated PSAR and SE Sections PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section DL3-RC1 Scram - Gravity Driven Absorber Insertion by Latch Release Upon receipt of a scram signal from RPS, the Control Rod Drive System shall release the control rod assemblies in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Control Rod Drive System (CRD) 7.2.5 Upon CRD rod release, control rod assemblies shall insert into the core to a depth and within a time limit to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition Reactor Core System (RCC) 7.1.1 Upon CRD rod release, RES shall provide structural support and position control preventing binding of CRDL and ensure control rod release is uninhibited to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Enclosure System (RES) 7.1.2 The RPS shall receive monitored signals and generate a scram signal upon exceeding a scram setpoint in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Protection System (RPS) 7.6.3 XIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Nuclear Instrumentation System (XIS) 7.6.4 RIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Instrumentation System (RIS) 7.6.5 DL3-RC2 Reactor Scram on Loss of Power Upon receipt of a scram signal from RPS, the Control Rod Drive System shall release the control rod assemblies in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Control Rod Drive System (CRD) 7.2.5

6-16 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section Upon power loss, the RPS shall fail safe by generating a scram signal in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Protection System (RPS) 7.6.3 DL3-HR1 PSP Coastdown The Primary Sodium Pump (PSP) shall coast down upon a pump trip to support transition to natural circulation, without flow reversal within the fuel channels, to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Primary Heat Transport System (PHT) 7.1.3 DL3-HR2 PSP Trip on High High Primary Sodium Temperature Upon receiving a trip signal from the RPS, the PSPs shall trip to eliminate heat generation to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Primary Heat Transport System (PHT) 7.1.3 The RPS shall receive monitored signals and generate a PSP trip signal upon exceeding a high temperature limit with a low flux signal in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Protection System (RPS) 7.6.3 XIS and RIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Nuclear Instrumentation System (XIS) 7.6.4 XIS and RIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Instrumentation System (RIS) 7.6.5 DL3-HR3 ISP Trip on High High Primary Sodium Temperature Upon receiving a trip signal from the RPS, the Intermediate Sodium Pumps (ISP) shall trip to eliminate heat generation to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Intermediate Heat Transport System (IHT) 7.1.4

6-17 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section The RPS shall receive monitored signals and generate an ISP trip signal upon exceeding a high temperature limit with a low flux signal in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Protection System (RPS) 7.6.3 XIS and RIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Nuclear Instrumentation System (XIS) 7.6.4 XIS and RIS shall send monitored signals to the RPS in time to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Instrumentation System (RIS) 7.6.5 DL3-HR4 Inherent - RAC Operation RAC shall continuously transfer heat to the atmosphere via natural circulation at a rate during accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Air Cooling System (RAC) 7.2.1 Reactor Enclosure System (RES) 7.1.2 Reactor Building (RXB) 7.8.1 DL3-HR5 Natural Circulation of Sodium in Primary System The primary flow circuit elevations and geometries shall promote natural circulation to remove heat from the core at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Primary Heat Transport System (PHT) 7.1.3

6-18 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section The primary flow circuit elevations and geometries shall promote natural circulation to remove heat from the core at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 The primary flow circuit elevations and geometries shall promote natural circulation to remove heat from the core at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Enclosure System (RES) 7.1.2 DL3-HR6 Passive Heat Removal in EVHM The Ex-Vessel Handling Machine (EVHM) shall passively remove decay heat from spent fuel assemblies at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-HR7 Passive Heat Removal in EVST The Ex-Vessel Storage Tank (EVST) shall passively remove decay heat from spent fuel assemblies at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-HR8 Passive Heat Removal in BLTC The Bottom Loading Transfer Cask (BLTC) shall passively remove decay heat from spent fuel at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2

6-19 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-HR9 Passive Heat Removal in PRC The Pin Removal Cell (PRC) shall passively remove decay heat from spent fuel at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-HR10 Passive Heat Removal in FHP The spent fuel pool shall passively remove decay heat from spent fuel at a rate to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Water Fuel Pool Handling System (FHP) 7.3.1 Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-HR11 SPS Pump Trip on Low Low Primary Sodium Level Upon receiving a trip signal from the RPS, the SPS pumps shall trip in time to maintain reactor vessel primary sodium level high enough to establish reasonable assurance that natural circulation in the primary sodium flow circuit can develop.

Sodium Processing System (SPS) 7.2.4 RIS shall send monitored signals to the RPS in time to maintain reactor vessel primary sodium level necessary to establish reasonable assurance that natural circulation in the primary sodium flow circuit can develop.

Reactor Instrumentation System (RIS) 7.6.5

6-20 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section The RPS shall monitor received signals, and if primary sodium level exceeds the setpoint after a scram shutdown, the RPS shall trip the SPS pumps in time to maintain reactor vessel primary sodium level high enough to establish reasonable assurance that natural circulation in the primary sodium flow circuit can develop.

Reactor Protection System (RPS) 7.6.3 DL3-HR12 ISP Pump Trip on High High Primary Sodium Level Upon receiving a trip signal from the RPS, the ISPs shall trip in time to prevent over-pressurization and failure of the primary boundary, to establish reasonable assurance primary inventory is maintained, and establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Intermediate Heat Transport System (IHT) 7.1.4 XIS and RIS shall send monitored signals to the RPS in time for the plant to respond to prevent over-pressurization and failure of the primary boundary, resulting in a safe shutdown condition.

The functions will provide reasonable assurance primary inventory is maintained and any associated radiological dose calculation results are maintained under the 10 CFR 50.34 dose criteria.

Reactor Instrumentation System (RIS) 7.6.5 XIS and RIS shall send monitored signals to the RPS in time for the plant to respond to prevent over-pressurization and failure of the primary boundary, resulting in a safe shutdown condition.

The functions will provide reasonable assurance primary inventory is maintained and any associated radiological dose calculation results are maintained under the 10 CFR 50.34 dose criteria.

Nuclear Instrumentation System (XIS) 7.6.4 DL3-HR12 (cont.)

ISP Pump Trip on High High Primary Sodium Level The RPS shall monitor received parameters, and if primary level exceeds the setpoint after a scram shutdown, the RPS shall trip the ISPs. This function shall occur in time to prevent over-pressurization and failure of the primary boundary, which provides reasonable assurance the primary inventory is maintained, and radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Protection System (RPS) 7.6.3

6-21 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section DL3-HR13 Passive Heat Removal in the Failed Fuel Canister The Failed Fuel Canister shall passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that the radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Water Fuel Pool Handling System (FHP) 7.3.1 Fuel assemblies shall be designed to passively remove decay heat at a rate through natural circulation and conduction to establish reasonable assurance that radionuclide release results in calculated radiological dose under the10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-RR1 Primary Coolant Boundary including RES Barrier The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Enclosure System (RES) 7.1.2 The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Primary Heat Transport System (PHT) 7.1.3 The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Sodium Processing System (SPS) 7.2.4 DL3-RR1 (Cont.)

Primary Coolant Boundary including RES Barrier The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Sodium Cover Gas System (SCG) 7.2.3 The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Control Rod Drive System (CRD) 7.2.5

6-22 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section The primary system boundary shall have low leakage in postulated accident conditions to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

In-Vessel Fuel Handling System (FHI) 7.3.3 DL3-RR2 Cladding barrier During accident conditions, the fuel cladding shall retain radionuclide fission products to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Core System (RCC) 7.1.1 DL3-RR3 EVHM Cask barrier During postulated accidents, the EVHM, in a standalone state, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents, when the EVHM is joined with the reactor head, the combined barrier shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents, when the EVHM is joined with the EVST, the combined barrier shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 DL3-RR4 EVST Barrier During postulated accidents the EVST, in a standalone state, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents, when the EVST is joined with the BLTC, the combined barrier shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2

6-23 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section The EVST Auxiliary Systems, up to and including the isolation valves, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents the EVST, in a standalone state, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Sodium Processing System (SPS) 7.2.4 The EVST Auxiliary Systems, up to and including the isolation valves, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Sodium Processing System (SPS) 7.2.4 DL3-RR5 BLTC barrier During postulated accidents the BLTC, in a standalone state, shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents, when the BLTC is joined with the Pool Immersion Cell (PIC), the combined barrier shall contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 DL3-RR5 (Cont.)

BLTC barrier During postulated accidents, during all fuel movements in and out of the BLTC, the installed equipment shall temporarily contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 During postulated accidents, during all fuel movements in and out of the BLTC, the installed equipment shall temporarily contain radionuclides from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Water Pool Fuel Handling System (FHP)

6-24 PRA Safety Function Function Description SRDC from PSAR Table 5.2-1 through 5.2-3 SSC PSAR/SE Section DL3-RR6 PRC Cell barrier The PRC shall contain radionuclide releases from the fuel to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Ex-Vessel Fuel Handling System (FHE) 7.3.2 DL3-RR7 RES Pressure Relief Prior to reaching the RES and supporting system functional containment design pressure, SCG shall actuate the primary pressure relief valve to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Sodium Cover Gas System (SCG) 7.2.3 Prior to reaching the RES and supporting system functional containment design pressure, SCG shall actuate the primary pressure relief valve to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Reactor Enclosure System (RES) 7.1.2 DL3-RR8 Failed Fuel Canister Barrier During accident conditions, the Failed Fuel Canister containing failed fuel shall retain radionuclide fission products to establish reasonable assurance radionuclide release results in calculated radiological dose under the 10 CFR 50.34 dose criteria at a safe shutdown condition.

Water Pool Fuel Handling System (FHP) 7.3.1

6-25 6.2 Reliability and Capability Targets for Safety-Significant SSCs 6.2.1 Introduction The staff reviewed the USOs assessment of reliability and capability targets for safety-significant structures, systems, and components (SSCs) provided in PSAR section 6.2. In addition, the staff audited the specific reliability targets supporting documentation which present reliability targets for a range of safety-significant SSCs identified through the applicants probabilistic risk assessment (PRA) and design processes.

The design reliability assurance program (D-RAP), as described by USO, is intended to ensure that safety-significant SSCs are designed, procured, constructed, operated, and maintained to meet reliability and availability levels commensurate with their safety importance, consistent with NRC guidance mentioned in the previous section.

For each SSC identified as safety-significant, the applicant established preliminary quantitative reliability targets intended to reflect the level of performance necessary to ensure the successful completion of the associated safety functions under both design-basis and beyond-design-basis conditions. These targets are documented in the D-RAP tables and are intended to guide design decisions, procurement specifications, equipment qualification, testing and inspection programs, and maintenance activities.

Regarding capability targets, in response to the staffs audit question, USO stated that these targets are derived from the PRAs success criteria analysis, the integrated plant analysis of licensing basis events (LBEs), and by source term analyses that influence event consequence evaluations. The success criteria and other LBE analyses serve as the starting point for capability target development, as they establish the SSC performance necessary to meet the applicable risk metrics. Where appropriate, the capability targets also account for the conditions under which the SSCs must perform their functions during the LBEs, as well as the timeframes within which active SSCs must operate to mitigate the LBEs.

6.2.2 Regulatory Evaluation Applicable guidance for the review of PSAR section 6.2 includes:

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR EditionSevere Accidents, section 17.4, Reliability Assurance Program, Revision 1 (ML13296A435)

RG 1.233, Revision 0 RG 1.253, Revision 0 6.2.3 Technical Evaluation The staff audited supporting documentation to confirm the description of reliability and capability targets in PSAR section 6.2.

6-26 Based on its review, the staff determined that the general approach described in the D-RAP, identifying safety-significant SSCs, establishing reliability targets, and integrating these targets into the design control and assurance process, is consistent with NRC principles for a graded approach to reliability assurance. This integrated approach provides a framework for ensuring that performance expectations for SSCs are defined early in the design process and maintained throughout subsequent design, procurement, construction, and operational phases.

However, the staff identified several areas where additional information is needed at the OL stage. Specifically, the staff noted that certain reliability targets were set below the typical values established for comparable safety functions in other reactor designs. At the CP stage, there are not detailed technical justifications to fully substantiate the adequacy of these lower targets. In addition, because the current PRA is preliminary and the SSC designs are not yet finalized, the determination on the proposed targets and if they are technically achievable or provide sufficient margin to ensure reliable performance of the associated safety functions will occur at the OL stage.

Accordingly, at the CP stage, the staff evaluated the adequacy of the process to determine frequency and probability values for reliability targets were acceptable. The finding on the acceptability of the specific frequency or probability values for reliability targets will be subject to further review at the OL stage, when more detailed design information, including final SSC configurations, failure modes analyses, and supporting analyses, will be available.

With respect to capability targets, the KU1 PSAR indicates that the specific operational performance criteria are included as part of the profile of the success of a PRA safety function.

These quantitative targets are based on analytical models used to demonstrate the capability of the plant to respond to hazards. The success criteria define the physical characteristics of systems or design features that are required to meet the safety functions defined. At the preliminary design stage reflected in the PSAR, these may still be at the system level with characteristics like total leakage rate specified, but as the design progresses specific characteristics for each SSC that contributes to the success of the PSF will need to be determined. These become the capability targets for those SSCs, relied upon to meet risk metrics. The staff confirmed in audit space that the process USO is using to develop the capability targets from the PRA success criteria, safety analysis, and source term analysis as described in the PSAR, is reasonable.

The staff expects the applicant to provide, during the OL stage, the specific reliability, availability, and capability targets for safety-significant SSCs, including:

The technical basis for each target; How the targets are derived from PRA and deterministic analyses; The verification and validation methods; and The programs and processes to ensure these targets are maintained throughout the plants operational life.

The staff recognizes that, given the preliminary level of design detail at the CP stage, it is reasonable to leave the finalization of the reliability, availability, and capability targets to the OL stage when sufficient design, procurement, and system integration information is available.

6-27 6.2.4 Conclusion Based on the review described above, the staff determined that USOs general approach to establishing reliability and capability targets for safety-significant SSCs, and implemented through the D-RAP, is reasonable for the CP stage. As discussed above, the applicant will need to fully demonstrate the target justification and quantitative capability criteria during the OL stage, based on the full-scope PRA developed in conformance with the non-LWR PRA standard.

The staff recognizes that, given the preliminary level of design detail at the CP stage, it is reasonable to leave the finalization of the reliability, availability, and capability targets to the OL stage when sufficient design, procurement, and system integration information is available. At that time, the staff will verify whether the reliability, availability, and capability targets derived from the PRA and other deterministic approaches are appropriately justified and acceptable to ensure continued compliance throughout the facilitys operational life.

Accordingly, the staff determined that the applicants derivation of reliability and capability targets, through the D-RAP and success criteria analysis, is an acceptable approach for the CP stage and expects the applicant to address the following at the OL stage:

The preliminary reliability targets provided are reassessed, verified, and validated; Technical justifications for certain less stringent reliability targets are fully developed; Quantitative capability targets for safety-significant SSCs are defined and documented; The RAP integrated decision process panel will reevaluate and confirm the established targets, and The applicant demonstrates that these targets are achievable and will be maintained throughout the design, construction, and operational phases of the plant.

6.3 Special Treatment Requirements for Safety-Significant SSCs 6.3.1 Introduction PSAR section 6.3 describes the special treatment requirements for safety-significant SSCs (i.e.,

SR and NSRST SSCs). Special treatments are those requirements that provide increased assurance beyond normal industrial practices that safety-significant structures, systems, or components (SSCs) perform their design basis functions. Special treatments are developed to provide reasonable confidence that safety-significant SSCs meet the reliability and capability targets.

6.3.2 Regulatory Evaluation Applicable guidance for the review of PSAR section 6.3 includes:

DANU-ISG-2022-01, March 2024 RG 1.233, Revision 0

6-28 6.3.3 Technical Evaluation PSAR section 6.3 indicates that preliminary special treatments applied to safety-significant SSCs are identified in PSAR chapter 7 considering the PRA safety functions of the SSC, as well as the safety-significance, risk significance, and equipment type of the SSC. The PSAR further states that special treatments will be finalized and applied to safety-significant SSCs once the reliability and capability targets are finalized. The final special treatments will be provided at the OL stage of the review.

The staff reviewed the information in PSAR section 6.3 and audited supporting documentation related to the process for selecting and applying special treatments. The staff compared the information in PSAR section 6.3, as well as the audited reports, with the guidance in NEI 21-07 as endorsed by RG 1.253 at determined that process for establishing special treatments are consistent.

6.3.4 Conclusion Based on its review described above, the staff determined the preliminary information regarding the process of identifying special treatments requirements is consistent with RG 1.253 and acceptable for the CP review.

6.4 Design of Safety-Significant SSCs 6.4.1 Introduction PSAR section 6.4 provides design attributes for safety-significant SSCs (i.e., SR and NSRST SSCs). Subsections within this chapter include:

6.4.1 Seismic Design 6.4.2 Design of Safety-Significant Structures 6.4.3 Mechanical Systems and Components 6.4.4 ASME BPVC,Section III Piping Systems, Piping Components, and Associated Supports 6.4.5 Application of Industrial Codes 6.4.2 Regulatory Evaluation The applicable regulatory requirements for the design of safety-significant SSCs are:

10 CFR 50.34(a),

10 CFR 50.35, and 10 CFR 50 Appendix S.

The applicable guidance for the review of PSAR section 6.4 includes:

DANU-ISG-2022-01, March 2024 DC/COL ISG-017, Interim Staff Guidance on Ensuring Hazard-Consistent Seismic Input for Site Response and Soil Structure Interaction Analyses, March 2010 (ML100570203)

RG 1.12, Nuclear Power Plant Instrumentation for Earthquakes, Revision 3 (ML17094A831)

RG 1.29 Seismic Design Classification for Nuclear Power Plants, Revision 6 (ML21155A003)

6-29 RG 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Revision 2 (ML2328/ML23284A272)

RG 1.87, Acceptability of ASME Section III, Division 5, High Temperature Reactors, Revision 2 (ML21091A276)

RG 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Revision 3 (ML12220A043)

RG 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Revision 1 (ML003739367)

RG 1.142, Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments), Revision 3 (ML20141L613)

RG 1.166, Pre-Earthquake Planning, Shutdown, and Restart of a Nuclear Power Plant Following an Earthquake, Revision 1 (ML19266A616)

RG 1.199, Anchoring Components and Structural Supports in Concrete, Revision 1 (ML19336A079)

RG 1.221, Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants, Revision 0 (ML110940300)

RG 1.243, Safety-Related Steel Structures and Steel-Plate Composite (SC) Walls for other than Reactor Vessels and Containments, Revision 0 (ML21089A032)

NUREG-0800 o Section 3.5.1.1, Internally Generated Missiles (Outside Containment), Revision 3 (ML070370569) o Section 3.5.1.2, Internally Generated Missiles (Inside Containment), Revision 3 (ML070380167) o Section 3.5.1.3, Turbine Missiles, Revision 3 (ML063600395) o Section 3.5.3, Barrier Design Procedures, Revision 3 (ML070570004) o Section 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Revision 3 (ML070550032) o Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Revision 3 (ML16088A041) o Section 3.6.3, Leak-Before-Break Evaluation Procedures, Revision 1 (ML063600396) o Section 3.7.2, Seismic System Analysis, Revision 4 (ML13198A223) o Section 3.9.3 ASME Code Class 1, 2, and 3 Components, and Component Supports, and Core Support Structures Revision 3 (ML14043A231)

The applicable principal design criteria (PDC) that apply to the review of this section are:

PDC 1 Quality Standards and Records PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 14, Primary Coolant Boundary PDC 31, Fracture Prevention of Primary Coolant Boundary PDC 80, Reactor Vessel and Reactor System Structural Design Basis PDC 81, Reactor Building Design Basis

6-30 6.4.3 Technical Evaluation 6.4.3.1 Seismic Design PDC 2 requires that safety-significant SSCs are designed to withstand the effects of natural phenomena including earthquakes. USO is using a graded approach to seismic design considering the safety significance of SSCs. The performance targets based on the seismic classification and corresponding design approach will be modeled within a seismic probabilistic risk assessment (SPRA). An iterative process will then be used by USO to determine if requirements should be raised or can be lowered. Initial seismic classifications are assigned at the CP stage and will be confirmed or adjusted for the OL stage based on the SPRA. The final performance targets and special treatments are determined by iterating between the design of SSCs and the SPRA until the overall seismic risk for the facility is acceptable.

External hazard PRAs, including the SPRA, will be developed by USO for the OLA, but the external hazard PRAs are not required at the CP stage. Review of the USO seismic approach focused on understanding the basis for selection of initial seismic classifications, codes and standards, and special treatments to provide reasonable assurance that the USOs approach provides an adequate level of safety at the CP stage. The final SPRA developed at the OL stage will confirm that the facility can be constructed and operated safely.

With respect to seismic instrumentation, PSAR section 6.4.1 refers to section 1.1.4.3.16, which states that the seismic monitoring system (SMS) equipment is designed, built, and tested in full conformance with RG 1.12. The PSAR further states that the SMS provides time history acceleration data in the free-field and safety-related structures and indicated that acceleration sensors are located in the NCB, RXB, FHB, and NI yard areas. In addition, the SMS provides the data necessary to directly compare the measured motion with both site-specific ground motion response spectra (GMRS) and the design in-structure response spectra to evaluate the seismic response of nuclear power plant features important to safety promptly after an earthquake as required by 10 CFR 50 Appendix S.

The staff reviewed PSAR section 1.1.4.3.16 and determined that more specificity will be needed at the OL stage regarding the actual locations of all proposed instrumentation as well as the bases for these locations. In addition to the instrument type and locations, a discussion regarding instrument operability, characteristics, installation, remote indication, and maintenance (per RG. 1.12) should be provided at the OL stage. The staff also determined that more specificity regarding the proposed shutdown criteria (per RG 1.166), including the OBE and SSE exceedance criteria, will be needed at the OL stage. The staff determined that this information can reasonably be left to the OL stage of the review, and the staff will review USOs seismic instrumentation program at that time. Additionally, as described in section 6.4.3.3 of this SE, justification for whether additional seismic monitoring is needed on the isolated portion of the facility will also be reviewed at the OL stage.

6.4.3.1.1 Seismic Classification of SSCs The staff reviewed PSAR section 6.4.1 and tables 6.4-1A and B and 6.4-2 to understand the preliminary approach to seismic classification of SSCs and corresponding seismic design requirements. Safety-related SSCs are classified by USO as either SCS1 or SCS2. USOs preliminary seismic classification approach for SR SSCs is to assign SCS1 to all SR SSCs

6-31 unless failure of the SR SSC in a seismic event does not adversely affect the safety significant function, in which case the SR SSC is assigned an SCS2 classification.

USO will design SR SSCs to meet the requirements in Appendix S to 10 CFR Part 50. An SCS1 classification imposes seismic requirements consistent with the Seismic Category I classification from RG 1.29. USO applies the following codes and standards to SCS1 and SCS2 SSCs, which are consistent with historical SR SSC design for light water reactors:

ASCE/SEI 4-16, Seismic Analysis of Safety-Related Steel Structures for Nuclear Facilities.

ACI 349-13, Building Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary.

American National Standards Institute/American Institute of Steel Construction (ANSI/AISC) N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities.

The difference between SCS1 and SCS2 SR SSCs is the qualification criteria. SCS1 qualification requires reasonable assurance of function under seismic loads consistent with SC-1 SSCs described in RG 1.29, while SCS2 qualification requires reasonable confidence of function under seismic loads consistent with ASCE/SEI 7-16.

According to USOs methodology and classification scheme as summarized in PSAR table 6.4-1B and section 6.4.1.1, non-safety-related with special treatment (NSRST) SSCs are classified as either SCN1, SCN2, or SCN3. The initial seismic classification approach for NSRST SSCs assigns SCN3 as a minimum. The seismic classification is then increased to SCN1 or SCN2 as appropriate based on factors such as SSC function, importance to life safety, and potential contribution to events with dose consequences. SCN1 will be assigned as the seismic classification for active components and for SSCs identified as being seismic risk significant. Seismic risk significance is determined by evaluating the role of the SSC in preventing or mitigating LBEs and the potential unmitigated dose consequences from a common cause seismic failure. SCN2 or SCN1 are assigned as appropriate to meet minimum building code requirements. SCN3 will not be applicable to structures but is reserved for non-structural systems or components.

USO performed the seismic design of NSRST structures in accordance with the following commercial industrial codes and standards:

International Building Code (IBC) 2021, International Building Code ASCE/SEI 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures ACI 318-19, Building Code Requirements for Structural Concrete ANSI/AISC 360-16, Specification for Structural Steel Buildings The staff determined that the approach to initial seismic classification is reasonable, and that the design codes associated with each seismic classification are reasonable based on the risk-informed approach described in NEI 18-04, as endorsed in RG 1.233. USO committed to confirm or adjust the initial seismic classifications and associated design criteria at the OL stage based on the seismic PRA, which provides reasonable confidence that the SSCs will meet the intended performance targets when evaluated at the OL stage using the SPRA.

6-32 6.4.3.1.2 Seismic Design Basis PSAR section 6.4.1.2.1 states that the seismic design for SR SSCs is based on the SSE. The development of the SSE design response spectra is described in PSAR section 2.6.2. The site horizontal and vertical SSE response spectra are shown in PSAR figure 2.6-81 and the spectral acceleration data is provided in PSAR table 2.6-15. The applicant stated in PSAR section 2.6.2.4.2 that the site GMRS, generated at 15 ft below the ground surface, is selected as the SSE. The staff reviewed and determined the applicants information was acceptable regarding the development of the GMRS in SE section 2.6.3.2.4. USO did not develop a separate seismic analysis involving the OBE in accordance with RG 1.61, because the OBE is one-third of the SSE.

For seismic design, USO stated in PSAR section 6.4.1.2.2 that distinct foundation input response spectra (FIRS) are developed for the RXB, FHB, and NCB. The full column method, which includes soil layers above the respective foundation levels, was used to develop the FIRS. USO stated in PSAR section 2.6.4.10.1 that the shallow foundations are constructed on compacted backfill and weathered rock and the deeper foundations are constructed on fresh rock. The dynamic properties of soil layers, shear wave velocities, and strain-compatible shear moduli and damping values required for the full column method are discussed in PSAR section 2.6.4.2.9 and evaluated by the staff in SE section 2.6.4.

To implement the full column method, a soil column analysis is typically performed using one-dimensional equivalent-linear wave propagation analysis. This approach is accepted as standard practice in geotechnical engineering for developing FIRS, in accordance with ASCE 4-16, and consistent with staff guidance in NUREG-0800 (Section 3.7.2) and DC/COL ISG-017 for the seismic design of nuclear power plants. According to PSAR section 6.4.1.2.2, the FIRS are developed at distinct foundation levels of below grade SR structures. The details of where the FIRS are developed for each of the substructures will be provided and reviewed at the OL stage. The staff will confirm at the OL stage that the final design is based on the methodology described and that the applicant used site specific geotechnical data for developing seismic design input.

According to PSAR section 6.4.1.2.3, USO will develop the seismic response spectra for NSRST SSCs in accordance with ASCE/SEI 7-16 considering site-specific ground motion parameters. USO also stated that the response spectra for SCN1, SCN2, and SCN3 will envelope both the SSE and the risk-targeted maximum considered earthquake (MCER). The specific factors used to modify the SSE per ASCE/SEI 7-16, including importance factors, are listed in PSAR table 6.4-1B. Response Modification Coefficients depend on the type of SSC being analyzed, as defined in Tables 12.2-1, 12.14-1, 15.4-1, and 15.4-2 of ASCE/SEI 7-16.

The staff determined that the development of the design ground motions for SR and NSRST SSCs is reasonable because, i) they rely on the seismic hazard analysis that was reviewed and accepted by the staff in SE section 2.6.2.2, and ii) USO is following guidance in RG 1.61 to define the OBE based on the SSE. Furthermore, USO will develop the FIRS at the respective foundation levels for substructures based on the SSE. The specific FIRS will be developed and evaluated at the OL stage. The staff are evaluating the MCER with respect to the review of the

6-33 specific SSCs that these are applied to as described in sections 6.4.3.1.5 and 6.4.3.2.2 of this SE.

6.4.3.1.3 Seismic Design and Analysis of SR SSCs USO's methodologies for seismic analysis and design of SR SSCs are detailed in PSAR section 6.4.1.3 with application of the methodology to the preliminary design of the SR buildings and structures detailed in PSAR section 7.8. The seismic analyses for SR SSCs are based on the equivalent static method or dynamic analysis.

Dynamic analysis methods include response spectrum analysis, time history analysis, or random vibration theory. Response spectrum and time history analysis methods are performed in accordance with guidance in RG 1.92. For time history analysis, earthquake components are applied in three directions, either in separate analyses or concurrently in a single analysis. The random vibration theory is based on ANSI/ASCE 4-16 for soil structure interaction (SSI) analysis.

USO will apply the equivalent static load method described in the PSAR for SR SSCs per ASCE/SEI 4-16 if i) the SSC can be represented by a simple model and it produces conservative results, ii) the method accounts for relative motion between points of support, and iii) the dynamic amplification factor, per section 4.5.1 of ASCE/SEI 4-16, is applied to obtain an equivalent static load for the SSC that can be represented by a simple model.

The approaches to SSI analyses, described in PSAR section 6.4.1.3.1 address (i) a range of soil and rock properties by considering three site profiles, (ii) effects of ground motion amplification, (iii) kinematic effects of embedded structures including inertial effects, and (iv) flexibility and impedance of embedded foundations. The applicant stated in PSAR section 7.8 that SSI for RXB and FHB substructures is modeled using SASSI.

USO considered the effects of structure-soil-structure-interactions (SSSI) because the proximity of the structures to each other may influence their foundation input motions. SSSI effects are addressed by modeling both structures in SSI analyses, evaluating interactions between RXB, RAB, and FHB structures.

Damping values are consistent with RG 1.61. Material damping is developed using site-specific geotechnical data and limited to 15% of the critical damping.

The staff determined that the methodologies for dynamic analysis, SSI, and SSSI are adequate since they are consistent with ASCE/SEI 4-16 and NUREG-0800 (section 3.7.2). The considerations for dynamic analysis provided in PSAR section 6.4.1.3.1 are reasonable and consistent with traditional approaches for dynamic analysis of SR SSCs. The staff also determines that the damping values for reinforced concrete and foundation soils are acceptable because they are consistent with the guidance in RG 1.61 and NUREG-0800 (section 3.7.1).

The information provided gives reasonable assurance that the plant can be constructed and operated safely. During the OL stage, the staff will confirm that the inputs and methodologies used for the analysis and the final design follow the methodology described in the PSAR and are consistent with PDC 2.

According to PSAR section 6.4.1.3.2, USO will develop three-dimensional finite element models for linear seismic analyses of SR structures. The seismic modeling will follow ASCE 4-16 provisions, including effective stiffness of reinforced concrete, mass inertia properties, dynamic

6-34 mass determination, mass discretization, model discretization, one-step response analysis, two-step analysis for less refined models, and lumped mass stick models.

USO will model dynamic coupling of the primary SR structure with secondary equipment per ASCE/SEI 4-16. USO will use seismic anchor motions and differential movement for design of systems that span multiple levels or transition between seismically isolated and non-seismically isolated portions of the structure. These details will be reviewed at the OL stage.

The staff determined the applicants methodology for analytical modeling of SR SSCs, as discussed in PSAR section 6.4.1.3.2, is adequate because the approach aligns with ASCE/SEI 4-16, section 3, and standard engineering practices accepted in NUREG-0800 (section 3.7.2).

The final design of specific SR SSCs will be reviewed at the OL stage.

PSAR section 6.4.1.3.3 states that modal responses are combined using the square root of the sum of the squares (SRSS) method. For closely spaced and high-frequency modes, the complete quadratic combination (CQC) method will be used. The staff determines this approach acceptable because it is consistent with RG 1.92.

USO will use SSI results that consider the geometry and embedment of structures, to develop natural frequencies and responses for SR structures embedded below the finished grade with multiple foundation levels. The staff determined this approach is reasonable as it aligns with ASCE/SEI 4-16.

PSAR section 6.4.1.3.5 addresses the development of in-structure response spectra (ISRS) at equipment locations for dynamic analyses and equipment support design. USOs approach is consistent with ASCE/SEI 4-16 and includes evaluating spectral damping values, identifying SSC floor locations in the SSI model, performing seismic analysis of structures for each soil cases, combining responses, ensuring that soil cases are enveloped and gaps are filled, smoothing and broadening is performed to produce the final ISRS results, and generating time histories per ASCE/SEI 4-16 and ASCE/SEI 43-19. The staff determined this method is acceptable as it follows NUREG-0800, section 3.7.2, and RG 1.122. The staff determined that the process that will be used to generate time histories for the ISRS is consistent with ASCE/SEI 4-16 and ASCE/SEI 43-19. Review of the ISRS and its use in equipment and systems design will be performed at the OL stage.

In accordance with ASCE/SEI 4-16 section 3.1, USO will consider accidental torsional effects in the seismic analysis and design of embedded SR structures. USO plans to increase shear forces in shear walls and other lateral load-carrying members by 5% to account for these effects. The staff determined that this approach is acceptable because it meets the requirements in ASCE/SEI 4-16 and NUREG-0800, section 3.7.2.

PSAR section 6.4.1.3.7 states that the seismic stability of SR substructures is not evaluated when the center of gravity is below grade level. The staff verified that this is consistent with guidance in ASCE 43-19 section 7.2.1 and determined it is a reasonable approach.

6.4.3.1.4 Modeling and Analysis of Underground SR Structures Directly Supporting NSRST Superstructures USO discussed the modeling and analysis of underground SR structures directly supporting NSRST superstructures in PSAR section 6.4.1.4. For SR concrete substructures supporting NSRST surface steel-framed structures, dynamic interaction effects are considered through

6-35 coupled dynamic analysis. The NSRST superstructures are modeled as a lumped mass connected to the detailed finite element model of the substructure. Sensitivity analyses are used to verify that the coupling effects adequately represent the lumped mass. The staff determined this approach is acceptable as it aligns with ASCE/SEI 4-16 section 3.7.

6.4.3.1.5 Seismic Interaction The methodology to evaluate seismic interactions of NSRST and NST SSCs with SR SSCs is discussed in PSAR section 6.1.3.1 and evaluated in section 6.1.3.3 of this SE. In addition to those considerations, PSAR section 6.4.1.5 states that USOs design process ensures that potential seismic interactions will not adversely impact the ability of SR SSCs to perform their required safety functions. Additionally, interactions of NSRST and NST SSCs with NSRST SSCs will not adversely impact the ability of a seismic risk significant NSRST SSC to perform its safety function. Seismic capabilities of NSRST SSCs are associated with their seismic classification as described in PSAR section 6.4.1.1 and evaluated in section 6.4.3.1.1 of this SE.

The staff determined USOs approach for addressing seismic interactions is acceptable as it provides reasonable assurance at the CP stage that seismic interactions will not impact safety-significant SSCs for design basis loading. This approach should be confirmed through the seismic PRA at the OL stage.

6.4.3.1.6 Seismic Analysis Methods and Procedures Used for Analytical Modeling for NSRST SSCs PSAR section 6.4.1.6 states that the seismic analysis and design of NSRST SSCs is performed using commercial design codes and standards, based on ASCE/SEI 7-16, including: i) seismic load effects in ASCE/SEI 7-16 section 12.4.2, ii) strength design and allowable stress approaches per ASCE/SEI 7-16 sections 2.3.6 and 2.4.5, respectively, and iii) seismic loading applied in two orthogonal horizontal directions independently, per ASCE/SEI 7-16 section 12.5.

Structural damping of 5% of the critical damping is considered in the analysis.

For seismic analysis based on equivalent lateral force, USO will use ASCE/SEI 7-16 section 12.8. For response spectrum analysis, the number of modes will either have a combined mass participation factor of 100% or achieve 90% in each orthogonal horizontal direction. The design response spectrum is modified by applying the R/Ie factor (where Ie is the importance factor and R is the response modification factor) and the calculated displacement is modified by the Cd/Ie factor to obtain inelastic displacement, per ASCE/SEI 7-16. The square root of the sum of the squares or quadratic combination methods are used to combine response parameters.

Accidental torsion effects are addressed using ASCE/SEI 7-16 sections 12.8.4.2 and 12.9.2.2.2.

When using the equivalent lateral force procedure, the amplification of accidental torsion is considered per section 12.8.4.3 of ASCE/SEI 7-16. Staff confirmed during the audit that the fragility parameters used in the design of NSRST SSCs exceed the minimum values in ASCE/SEI 7-16 and are sufficient to provide reasonable assurance that these SSCs, which are SCN1 or SCN2, will maintain their safety functions at the DBHL.

SSCs that are not safety-significant and do not have seismic interaction requirements are outside the scope of the staffs review for seismic design considerations. For NSRST SSCs requiring special treatment, the staffs review of the seismic analysis methods is described in section 6.4.3.2.2 of this SE.

6-36 6.4.3.2 Design of Safety-Significant Structures PSAR section 6.4.2 states that the RXB, FHB, RAB, and NCB buildings include below-grade reinforced concrete substructures and above-grade steel-framed superstructures. USO classified the RXB, FHB, and NCB substructures as SR and the RAB substructure as NSRST.

The steel superstructures of the FHB, NCB, and RAB are classified as NSRST, while the RXB superstructure is classified as NST. The above-ground steel structures of the FHB, RAB, and NCB are supported on the grade slab of their respective concrete substructures. The RXB superstructure foundation is independent and not structurally connected to the RXB substructure. PSAR chapter 1 illustrates the proposed structural configurations.

Preliminary design information for these buildings is discussed in PSAR section 7.8 and reviewed in section 7.8 of this SE. Detailed design information, including building dimensions and details about the structural members and foundations, will be reviewed by the staff at the OL stage.

For clarity, this section is divided into subsections 6.4.3.2.1, Design of SR Structures, and 6.4.3.2.2, Design of NSRST Structures, differing from the PSAR's topic-based organization.

For SR structures, the staff reviewed USOs design methodology per 10 CFR Part 50 (50.34(a)(3)(i), 50.34(a)(3)(ii), and Appendix S), including seismic design inputs, analysis methods, modeling methods, applicable codes and standards, loads and load combinations, and materials.

The NSRST classification of SSCs based on the LMP framework is being used for the first time under 10 CFR Part 50. For NSRST structures, the staff reviewed USOs proposed seismic design methodology and analysis. USO aims to establish performance targets for NSRST structures through the SPRA and the LMP framework, modifying the design if needed to meet performance targets and safety criteria. This review at the CP stage provides reasonable assurance that the final design developed with the proposed methodology will meet the intended performance targets when evaluated at the OL stage using the SPRA.

6.4.3.2.1 Design of SR Structures Methodology for Design and Analysis USO discussed applicable codes and standards (including specifications) and regulatory guidance in PSAR section 6.4.2.1. USO provided a list of codes and standards for design, fabrication, construction, testing, and inspection of safety-significant structures in PSAR tables 6.4-2, 6.4-3, and 6.4-4. The design codes for SR substructures are ACI 349-13 for reinforced concrete and ANSI/AISC N690-18 for structural steel. USO stated in PSAR section 6.4.2.3 that the concrete substructures are designed using the strength-based limit state design method with load and resistance factors from ACI 349-13 and RG 1.142.

The staff determined that USOs use of the referenced codes and standards is appropriate for the design of SR structures. ACI 349-13 and ANSI/AISC N690-18 are acceptable consensus codes and standards for concrete and structural steel design and construction of nuclear power plants in accordance with the guidance in NUREG-0800 (section 3.8.4), and as endorsed in RG 1.142 and RG 1.243.

6-37 PSAR section 6.4.2.2 describes loads and load combinations for structural design, including normal, severe environmental, extreme environmental, and abnormal loads. The load combinations for SR concrete substructures are in PSAR table 6.4-5, and for structural steel in table 6.4-6.

The staff determined that USOs information on loads and load combinations is appropriate for the design of SR structures because the load combinations and the load factors are consistent with ACI 349-13, RG 1.142, ANSI/AISC N690-18, and RG 1.243. The staff will review the application of these loads and load combinations at the OL stage.

USO stated in PSAR section 6.4.2.3.2 that structural analysis and design for SR concrete substructures is performed using GT STRUDL Version 40 to model walls and slabs with shell elements. The dynamic mass includes the dead weight of the structure plus 25% live load, superstructure load (dead weight plus 25 percent live load and 75 percent snow load), and equipment load. The superstructure is modeled separately, with its load represented as a dynamic mass in the substructure model.

The applicant used best-estimate stiffness properties (e.g., effective stiffness) for concrete per ASCE/SEI 4-16 table 3-2. Concrete sections are assumed cracked in flexure for out-of-plane responses and uncracked in axial and shear. Shear walls supported by soil are not considered cracked in flexure.

The staff determined that USOs information on the analysis of SR substructures is adequate for preliminary design because GTSTRUDL is an industry-standard The dynamic mass is consistent with ASCE/SEI 4-16 section 3.4.2.

The stiffness properties for cracked and uncracked concrete are consistent with ASCE/SEI 4-16 (section 3.3.2).

The mesh sizes for the finite element models are adequate to transmit the entire frequency range of interest.

The superstructure reactions on the substructure are adequately considered.

The staff will review the final design at the OL stage.

Materials In PSAR section 6.4.2.4, USO discussed material standards for concrete and steel used in SR and NSRST structures. Concrete material properties are in PSAR table 6.4-10, with a compressive strength of 5000 psi for safety-significant structures. Concrete works follow ACI 301-16. Reinforcing bars are Grade 60, per ASTM A615-09b and ASTM 706-09b, and headed reinforcing bars per ASTM A970-07. Structural steel conforms to ANSI/AISC N690-18 for SR structures, and the stainless-steel liner plate for the FHB spent fuel pool follows ASTM A240-16a.

The staff determined that the applicants information on the concrete and steel material properties is acceptable because it conforms to applicable design and ASTM codes and standards. The staff will verify materials used in the final design at the OL stage.

6-38 Summary of the staffs determinations The staff reviewed USOs methodology for SR substructure design and determined that the information provided gives reasonable assurance that the plant can be constructed and operated safely and that the final design will meet PDC 2, 80, 81, and 10 CFR Part 50, Appendix S requirements for withstanding earthquakes without impairing structural integrity and performance. This conclusion is based on:

1. USO provided sufficient information on applicable codes, standards, and specifications; loads and loading combinations; design and analysis procedures; structural acceptance criteria; and materials, quality control, and special construction techniques that will be used in the final design.

2. USO will design the structure to withstand the most severe earthquake established for the site and includes adequate consideration of the combinations of the effects of normal and accident conditions with environmental loadings such as earthquakes and other natural phenomena.

3. USOs seismic analysis method is based on appropriate procedures for structural modeling, SSI, development of floor response spectra, inclusion of torsional effects, consideration of foundation stability, and adequate description of geotechnical and structural material properties.

4. USOs structural analysis and design conform with established criteria, codes, standards, and specifications acceptable to the staff, including RGs 1.61, 1.122, 1.92, 1.142, 1.243, and 1.199, and industry standards ASCE 4-16, ACI 349-13, and ANSI/AISC N690-18.

5. USOs information is consistent with DANU-ISG-2022-01 section 1.1.2 which references NUREG-0800.

The staff determines the PSAR information meets 10 CFR 50.34(a) and 50.35 requirements for issuing a construction permit. The final design will be confirmed during the OL stage.

6.4.3.2.2 Design of NSRST Structures Methodology for Design and Analysis USO will use the provisions in ASCE/SEI 7-16, chapter 12 for the design of the FHB and NCB superstructures, and chapter 15 for the RAB superstructure, which is categorized under non-building structures that are similar to buildings. As discussed in section 6.4.3.1.2 in this SE, the MCER is derived from the SSE, with seismic classifications of NSRST SSCs as evaluated in 6.4.3.1.1. The seismic design parameters for NSRST structures classified as SCN1 or SCN2 are provided in PSAR table 6.4-1B. The primary design codes and standards for concrete and steel structures for NSRST substructures and superstructures are discussed in PSAR sections 6.4.2.1. For NSRST substructures and superstructures, the applicant plans to use ACI 318-19 for concrete and ANSI/AISC 360-16 for structural steel. Concrete structures will use the ultimate strength design method, and structural steel frames will use the LRFD design method.

PSAR section 6.4.2.2.5 discusses loads and loading combinations for NSRST structures. USO will use loads and load combinations from IBC-2021 for NSRST structures or those in ACI 318-

6-39 19 and ANSI/AISC 360-16 for more stringent loading. For extraordinary events, load combinations and factors from ACI 349-13 for concrete and ANSI/AISC N690 for steel will be used, as detailed in PSAR tables 6.4-7, 6.4-8, 6.4-11, and 6.4-12.

Design of NSRST Superstructures PSAR section 6.4.2.3.3 describes the design of NSRST superstructures. Reinforced concrete structural components included in NSRST superstructures will be designed using ACI 318-19, and structural steel frames using AISC 360-16 LRFD methodology, complying with AISC 325-17, AISC 341-16, ASCE/SEI-7, and IBC 2021. USO will design bolt connections following ANSI/AISC 360-16 and AISC 348-14, and weld connections following ANSI/AISC 360-16. USO will design concrete anchors and steel embedment following ACI 318-08. Per PSAR section 6.4.2.3.3, NSRST superstructures will be designed to withstand tornado missile impacts, but the building envelope will only be designed for normal wind loading and will not be relied on to protect the building interiors from tornado wind and missile impacts.

Analysis of NSRST Superstructures PSAR section 6.4.2.3.4 states that structural modeling of NSRST steel superstructures will use GT STRUDL. The lateral force resisting system will be represented in a three-dimensional finite element model, with loads, load combinations, and seismic analysis methodology in accordance with ASCE/SEI 7-16. The model will provide structural member design, deflection checks, support reactions, dynamic responses, and structural steel quantities.

Summary Although NSRST structures are safety significant, the NRC has not established detailed regulatory positions for them as it has for SR structures in NUREG-0800. Consequently, applicants should propose their own approach to designing safety-significant NSRST structures, including the design codes and standards that could be different from those approved by the staff for SR structures. USOs proposed approach uses ASCE/SEI 7-16 and associated codes and standards, which are well-established for commercial construction. However, these codes and standards have not been used in a license application for the design of safety-significant structures, nor has the NRC independently reviewed ASCE/SEI 7-16 for generic application to nuclear power plant licensing.

The staffs review in this SE is not evaluating or endorsing ASCE/SEI 7 for general use. Instead, the staff focused on USOs design approach, NSRST SSC classification, and the basis for the codes and standards used for NSRST structure design. USOs classification of NSRST structures at the CP stage will be confirmed or adjusted at the OL stage based on a risk-informed process and seismic PRA. The staff determines the overall design approach acceptable because its initial classification of NSRST SSCs at the CP stage aligns with RG 1.253 (which endorses NEI 21-07) and RG 1.233 (which endorses NEI 18-04). USOs commitment to confirm or adjust classifications and associated design criteria at the OL stage based on seismic PRA provides reasonable confidence that the performance target will be achieved.

Given the iterative design process inherent in the LMP framework, and as this is the first application of the RIPB design framework, continued engagement on the seismic design and SPRA as part of pre-application engagement prior to OLA submittal would be beneficial.

6-40 6.4.3.3 Mechanical Systems and Components PDC 1, 2, and 4 require, in part, that SSCs important to safety be constructed and tested to quality standards commensurate with the importance of the safety functions to be performed and designed with appropriate margins to withstand the effects of anticipated normal occurrences, natural phenomena such as earthquakes, and postulated accidents.

PSAR section 6.4.3 describes the design and analysis of ASME Code Section III, Division 5 mechanical components, supports, and core support structures. PSAR section 6.4.3.1.1 describes the plant transients used in fatigue analyses. The service level loadings considered in the identification of plant transients include:

Level A Service Loadings - Loading arising from system startup, hot standby operation and system shutdown Level B Service Loadings - Loadings that are anticipated to occur often enough that the design should include a capability to withstand the loadings without operational impairment Level C Service Loadings - Loadings require a shutdown for correction or repair of damage in the system Level D Service Loadings - Loadings associated with postulated events of extremely low probability whose consequences are such that the integrity and operability of the nuclear system may be impaired Testing Loadings - Loadings that occur during hydrostatic tests, pneumatic tests, and leak tests The transients used in the design of ASME Section III components are listed in PSAR table 6.4-

16. The design transients cover all normal, off-normal, and test operations that are expected to occur during the design life of the plant. The events account for filling and draining of the sodium, heat up and cool down of the systems, startup and shutdown operations, power maneuvering operations, and plant responses to upset, emergency, and faulted initiating events. During the audit, the staff confirmed the design transients and the preliminary number of occurrences for each design transient. Based on the staffs review, the staff determines the identified plant transients and the number of transients is appropriate for the design life.

PSAR section 6.4.3.1.2 identifies the computer programs used in the static and dynamic analysis of mechanical equipment. The staff has familiarity with these codes and their capabilities and determines the identified computer programs appropriate for use in the design and analysis of mechanical equipment. The staff will review the application of these codes for detailed analyses of mechanical equipment for the final design at the OL stage.

PSAR section 6.4.3.2 describes the seismic analysis of mechanical systems and components.

The seismic analysis of mechanical systems and components follows the methodology described in PSAR section 6.4.1. The staff evaluation of the seismic design and analysis methods for safety-significant SSCs is documented in section 6.4.3.1 of this SE. PSAR section 6.4.3.2 also describes the methods for analysis of the seismically isolated Reactor Enclosure System (RES). As discussed in section 7.1.2.3, the RES is supported inside the RXB substructure with a seismic isolation system (SIS). The design and qualification methodology for the SIS will be in accordance with TerraPowers SIS topical report (TR) [NAT-8922, Revision 2, (ML25195A156)]. The staff reviewed the methodology for the seismic analysis of the RES and found that the approach includes: (i) selection of frequency; (ii) determination of number of

6-41 earthquake cycles; (iii) consideration of damping for modal response; and (iv) consideration of static vertical load factors acting simultaneously with two horizontal dynamic loadings. The staff determined that the approach is acceptable because it is consistent with guidance on seismic analysis of mechanical components and systems in NUREG-0800, sections 3.7.2 and 3.7.3.

The staff will review the applications of these analyses to mechanical equipment in the final design at the OL stage.

In the SE for NAT-8922 (ML25296A226), the staff determined that the methodology provides an acceptable approach for future applicants using the Natrium design to establish the design criteria and qualification requirements for the 3D isolation system described, subject to limitations and conditions (L&Cs). The staff evaluated these L&Cs to confirm they were evaluated as part of the PSAR and either met or could be reasonably left to the OL stage when the design is finalized. The staffs SE for NAT-8922 imposed the following L&Cs:

1. The conclusions reached in this SE only address the content provided in section 7 of the TR. Thus, any licensee or applicant referencing this TR must evaluate the other aspects of the information described in the remaining six sections of the TR for any site-specific application.

2. An applicant or licensee referencing this TR must use the Natrium design, as summarized in sections 5.1 and 6 of the TR, or justify that any departures from these design features do not affect the conclusions of the TR and this SE.

3. The methodology described in the TR and the conclusions reached in this SE are based on a component 3D isolation system using ISUs and IDUs which limit displacement and are arranged to ensure an even distribution of loads within the SIS and limits the seismic demands exerted on the reactor. The details of the methodology in TR section 7 that were reviewed by the NRC staff are limited to this specific component 3D isolation technology.

4. If an applicant or licensee referencing this TR chooses to follow a generic qualification process as described in TR section 6.1, they must perform seismic analyses to confirm that the site-specific motion (based on the site and design of RXB, SIS, and supported subsystems) is enveloped by the generic ground motion. For conditions in which site-specific ground motion spectra are not enveloped by the bounding generic ground motion spectra, a new bounding spectra must be generated and the qualification process must be repeated.

5. The conclusions reached in this SE are limited to a design that does not include a dynamic stop. If impact between the IDU piston and housing under extreme earthquake loading is deemed possible, then a sensitivity analysis should be conducted to bound the impact loads and response of critical SIS supported SSCs to determine the potential risks.

6. The conclusions reached in this SE are based on TerraPowers methodology for verification and validation of numerical models capable of predicting results of dynamic testing of the prototype isolators consisting of linear springs and viscoelastic dampers.

To use this methodology at a specific site, an applicant or licensee shall confirm that the range of applicability for the numerical models encompass the site-specific conditions.

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7. The conclusions reached in this SE are based on TerraPowers multi-step or single-step methodology for plant seismic response analysis. To use this methodology at a specific site, an applicant or licensee shall develop a validation plan for the single-step or multi-step response analysis (including transitions between analysis codes), as applicable, and a verification plan for the codes used in the response analysis.

8. The conclusions reached in this SE are based on the specific codes and standards used to implement the methodology, listed in TR section 7. Applicants or licensees referencing this TR should justify any deviations to these codes and standards. An application using the TR methodology at any site requires a peer review as described in TR table 7-1.

9. Applicants or licensees referencing this TR must provide a basis for the adequacy of seismic monitoring equipment for the site-specific application that addresses the unique considerations for seismically isolated systems and the recommendations provided in NUREG/CR-7253, including justification for the location of instrumentation relative to the location of the seismic isolators.

The staff reviewed and determines that the PSARs use of NAT-8922 meets TR L&C 1 and 2 because the preliminary Natrium design described in the PSAR is consistent with the design information provided in the TR, including the descriptions in sections 5.1 and 6 of the TR. PSAR section 7.1.2.3 provides a description of the SIS, which uses isolation technology that is consistent with what is described in the TR, namely a 3D isolation system using ISUs and IDUs.

This is sufficient to adequately address L&C 3 at the preliminary design stage. While the PSAR does not explicitly state that the generic bounding ground motion spectra envelops the site-specific spectra, PSAR section 7.1.2.3 states that testing will be done to the full range of loading conditions experienced during a seismic event or per ASME QME-1, scaled units qualified as parent restraints will be used with design similarity and analysis to qualify the full scale units.

This information is sufficient to address L&C 4 at the preliminary design stage. At the OL stage, USO should verify explicitly that the generic bounding ground motion spectra envelops the site-specific spectra. USO adequately addressed L&C 5 by confirming, in PSAR section 7.1.2.3, that a dynamic stop is not used. L&Cs 6 and 7 are related to the verification and validation (V&V) of numerical models and analysis codes. The information provided in the TR on the approach to V&V is sufficient for the CP stage. It is reasonable to leave the review of final V&V methods and execution, addressing L&C 6 and 7, to the OL stage. USO did not identify any deviations from the codes and standards used in the methodology described in the TR. If deviations are identified as part of further design, justifications for those deviations will need to be provided at the OL stage to meet L&C 8.

A basis for the adequacy of seismic monitoring equipment that addresses the unique considerations for seismically isolated systems was not provided in the PSAR to address L&C 9.

This justification along with additional details on the seismic monitoring system are expected at the OL stage. It is reasonable to leave this to the OL stage because the justification is likely to rely partly on the level of uncertainty present in the numerical models and analysis codes, which are determined through the V&V to be performed between now and the OL stage.

As discussed in the SE for the SIS TR, the methodologies provided in the TR contribute to meeting the requirements of PDC 1, 2, and 80 for the SIS.

6-43 PSAR section 6.4.3.3.1 provides the load combinations and service stress limits for mechanical components and associated supports. The loads considered for the design of ASME Code Class A and B mechanical components are summarized in PSAR table 6.4-17 and the loading combinations and the applicable stress limits are provided in PSAR table 6.4-18. The loading combinations and the corresponding stress limits for ASME Code design are defined for the design condition, service levels A, B, C, and D (i.e. normal, upset, emergency, and faulted conditions, respectively), and the test conditions. The staff reviewed the proposed loads, load combinations, and stress limits and concludes that appropriate combinations of operating design transients and accident loadings have been specified to provide a conservative design envelope for the design of mechanical systems and components. The staff will review the application of these loads and load combinations at the OL stage.

PSAR section 6.4.3.3.2 describes the design and installation of pressure relief devices. The applicant states that the reaction loading due to discharge of a relief device is analyzed using the guidance of ASME BPVC Section III, Appendix O. The staff determines this approach acceptable because it is consistent with the guidance in NUREG-0800, section 3.9.3.

6.4.3.4 ASME BPVC,Section III Piping Systems, Piping Components, and Associated Supports PSAR section 6.4.4 addresses the design and analysis of piping systems and piping supports.

The applicant clarified during the audit that this section is only applicable to piping systems and piping supports that select ASME Section III as their design code.

PSAR section 6.4.4.2 refers to ASME Section III, Division 1 for the design of piping and support in low temperature applications and ASME BPVC Section III, Division 5 for the design of piping systems in elevated temperature applications. RG 1.87, Revision 2 endorses the 2017 Edition of ASME BPVC Section III, Division 5 as acceptable to use for the design and construction of mechanical components that operate in elevated temperature environments. ASME Section III, Division 5 refers to Section III, Division 1 for low temperature applications. Therefore, the staff determines the use of ASME Section III, Division 1 and Division 5 acceptable.

6.4.3.5 Application of Industrial Codes PSAR section 6.4.5 addresses the application of ASME BPVC Section VIII, ASME B31.1, and ASME B31.3 industrial codes for the design and analysis of safety-significant SSCs. These industrial codes are generally selected for the design and construction of NSRST mechanical systems and components as well as selected SR SSCs. The applicant stated that supplemental requirements, in addition to the industrial code requirements, are applied as special treatments as necessary to provide enhanced SSC quality for safety-significant SSCs. The application of the supplemental requirements for major NSRST SSCs is addressed in Appendix 14.2 of NAT-13478 (ML25274A130) and the staff evaluation of these special treatments for major NSRST SSCs can be found in chapter 7 of the SE. The staff will review the special treatments for other NSRST SSCs at the final design stage in the OL application.

The applicant noted that ASME BPVC Section VIII, Division 1 and Division 2 provide rules for the design and construction of pressure vessels. Division 1 provides a design-by-rule approach that is utilized for noncomplex vessel designs. Division 2 allows for a design-by-analysis approach which provides more flexibility through detailed analysis.

6-44 For safety-significant SSCs designed to ASME BPVC Section VIII, Division 1, the combination of loads follows the methodology outlined in section UG-23. For safety significant SSCs designed to Section VIII, Division 2, table 4.1.2 of the code provides the minimum required load combinations. PSAR Section 6.4.5.1 states that the design loads from the ASME Section III transients are considered as loading input for ASME Section VIII component design to ensure the SSC will perform its safety-significant function during each transient event. The staff notes that ASME Section VIII does not specify different allowable stresses for different service level loads but does permit an increase in allowable stress when earthquake or wind loading is considered in combination with other loads.

PSAR section 6.4.5.1 also indicates that the rules in ASME Section VIII will be supplemented with additional analysis to account for fatigue and creep effects. Since ASME Section VIII, Division 1 does not address fatigue failure explicitly, the fatigue evaluation procedure from Section VIII, Division 2 will be applied. Furthermore, because Section VIII does not explicitly account for creep effects in long-term service, components subject to non-negligible creep, as determined by screening per Section III, Division 5, Mandatory Appendix HCB-III, will be evaluated using the Class A rules in ASME Section III, Division 5 or Code Case 2843-3. Code Case 2843-3 provides elevated temperature rules that closely mirror Section III, Division 5 Class A rules, including fatigue and creep-fatigue assessment. The staff reviewed the design approach for pressure vessels constructed to ASME Section VIII with the additional supplemental requirements described above and found the approach acceptable because the design rules in ASME Section VIII and the supplemental requirements provide assurance that the ASME Section III design transients and the effects of creep and fatigue are considered in the design and analysis of these SSCs.

PSAR section 6.4.5.2 discusses the design and analysis of piping systems in accordance with ASME B31.1 Power Piping and B31.3 Process Piping. ASME B31.1 and B31.3 provide rules for the design and construction of industrial power piping systems and process piping systems respectively. These piping codes are generally selected for the design and analysis of NSRST piping, piping components, and associated supports as well as selected SR SSCs.

ASME B31.1 and B31.3 require defining the loads that the component will experience in service, but do not relate these loads to service levels, or specify off-normal conditions. The PSAR states that the ASME Section III design transients are considered as loading input for the design of ASME B31.1 and B31.3 piping systems and components to ensure the SCCs will perform their safety-significant functions. The staff notes that ASME B31.1 and B31.3 do not specify different allowable stresses for different service level loads but do permit an increase in allowable stress when occasional loads are considered in combination with other loads.

ASME B31.1 and B31.3 piping codes are supplemented with rules from ASME BPVC Section III, Division 5 for SSCs in elevated temperature service in the creep regime on a case-by-case basis. For piping determined to have non-negligible creep effects by ASME BPVC Section III, Division 5 Appendix HCB-III, an analysis per ASME Section III, Division 5 Subsection HCB-3634 will be applied. Section HCB-3634 provides more conservative allowable stresses for thermal expansion stresses and stress criteria for thermal cycling to account for creep effects. The staff reviewed the design approach for piping constructed to ASME B31.1 or B31.3 with the supplemental requirements described above and found the approach acceptable because the design rules in ASME B31.1 or B31.3 and the supplement requirements provide assurance that

6-45 the ASME Section III design transients and the effects of creep are considered in the design and analysis of these SSCs.

6.4.4 Conclusion Based on its review and findings documented in the preceding subsections, the staff determines the preliminary information regarding the seismic, structural, and mechanical design of safety-significant SSCs is consistent with relevant guidance and acceptable for the CP review. Further information related to the design of safety-significant SSCs can be left for consideration at the OL stage.

6.5 References ACI 318-19, Building Code Requirements for Structural Concrete, ACI, 2019 Edition.

ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, ACI, 2013 Edition.

ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC, 2018 Edition.

ANSI/AISC 360-16, Specification for Structural Steel Buildings, ANSI/AISC, 2016 Edition.

ASCE/SEI 43-19, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE, 2020 Edition.

ASCE/SEI 4-16, Seismic Analysis of Safety-Related Steel Structures for Nuclear Facilities, ASCE, 2016 Edition.

ASCE/SEI 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, ASCE, 2016 Edition.

ASME, Boiler and Pressure Vessel Code (BPVC),Section III, Division 1 Rules for Construction of Nuclear Facility Components, ASME, 2017 Edition.

ASME, BPVC,Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, ASME, 2017 Edition.

ASME, BPVC,Section VIII, Rules for Construction of Pressure Vessels, ASME, 2021 Edition.

ASME Standard QME-1, Qualification of Active Mechanical Equipment Used in Nuclear Facilities, ASME, 2023 Edition.

ASME Standard B31.1, Power Piping, ASME, 2022 Edition.

ASME Standard B31.3, Process Piping, ASME, 2020 Edition.

IBC 2021, International Building Code, IBC, 2021 Edition.