ML25155B865

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NPM-20 - FW Final Safety Evaluation for NuScale TR-141299, Rev 1, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155 Prop and non-Prop
ML25155B865
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Site: 99902078
Issue date: 06/04/2025
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From:

Thomas Hayden Sent:

Wednesday, June 4, 2025 12:42 PM To:

NuScale-SDA-720DocsPEm Resource

Subject:

FW: Final Safety Evaluation for NuScale TR-141299, Rev 1, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events defined by 10 CFR 50.155 Prop and non-Prop Attachments:

Final Safety Evaluation of the NuScale Power LLC Topical Report TR-141299 Revision 1 NuScale Power Plant Design Capability to MBDBE defined by 10 CFR 50 - 051625.docx Tommy Hayden Project Manager Email: thomas.hayden@nrc.gov Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission From: Thomas Hayden Sent: Friday, May 16, 2025 9:56 AM To: Regulatory Affairs <regulatoryaffairs@nuscalepower.com>

Cc: Griffith, Thomas <tgriffith@nuscalepower.com>; Cummings, Kristopher

<kcummings@nuscalepower.com>; Shaw, Peter <pshaw@nuscalepower.com>; Mahmoud -MJ-Jardaneh <Mahmoud.Jardaneh@nrc.gov>; Getachew Tesfaye <Getachew.Tesfaye@nrc.gov>

Subject:

Final Safety Evaluation for NuScale TR-141299, Rev 1, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events defined by 10 CFR 50.155 Prop and non-Prop By letter dated June 26, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML24178A399 (Proprietary) and ML24178A398 (non-Proprietary)),

NuScale Power, LLC (NuScale), submitted Topical Report (TR) TR-141299, Revision 1, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155 to the U.S. Nuclear Regulatory Commission (NRC). The NRC staff has prepared a final safety evaluation for TR-141299, Revision 1. The non-proprietary (ML25136A034) and proprietary (ML25136A035) final safety evaluations are enclosed. The NRC staff has found that TR-141299, Revision 1, is acceptable for referencing in licensing applications for the NuScale small modular reactor design to the extent specified and under the conditions and limitations delineated in the enclosed final safety evaluation.

The NRC staff requests that NuScale publish the accepted version of this TR as soon as possible following receipt of this electronic mail. The accepted version shall incorporate this electronic mail and the enclosed final safety evaluation after the title page. It must be well indexed such that information is readily located. Also, it must contain historical review information, including

NRC requests for additional information and accepted responses. The accepted version of the TR shall include a -A (designated accepted) following the report identification number.

If the NRCs criteria or regulations change such that the NRC staffs conclusion in this electronic mail (that the TR is acceptable) is invalidated, NuScale and/or the applicant referencing the TR will be expected either to revise and resubmit its respective documentation or to submit justification for continued applicability of the TR without revision of the respective documentation.

If you have any questions or comments concerning this matter, I can be reached at (301) 415-2956 or via e-mail address at Thomas.Hayden@nrc.gov. The attached documents are both password protected. Password to follow in a separate email.

Docket No. 99902078 Sincerely, Tommy Hayden Project Manager Email: thomas.hayden@nrc.gov Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

Hearing Identifier:

NuScale_SDA720_Doc_Public Email Number:

19 Mail Envelope Properties (SA1PR09MB8478915A353EB495CEA7930BE86CA)

Subject:

FW Final Safety Evaluation for NuScale TR-141299, Rev 1, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events defined by 10 CFR 50.155 Prop and non-Prop Sent Date:

6/4/2025 12:42:22 PM Received Date:

6/4/2025 12:42:25 PM From:

Thomas Hayden Created By:

Thomas.Hayden@nrc.gov Recipients:

"NuScale-SDA-720DocsPEm Resource" <NuScale-SDA-720DocsPEm.Resource@nrc.gov>

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SA1PR09MB8478.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 3211 6/4/2025 12:42:25 PM image001.png 13623 Final Safety Evaluation of the NuScale Power LLC Topical Report TR-141299 Revision 1 NuScale Power Plant Design Capability to MBDBE defined by 10 CFR 50 - 051625.docx 134716 Options Priority:

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SAFETY EVALUATION BY THE U.S. NUCLEAR REGULATORY COMMISSION TOPICAL REPORT TR-141299, REVISION 1 NUSCALE POWER PLANT DESIGN CAPABILITY TO MITIGATE BEYOND-DESIGN-BASIS EVENTS DEFINED BY 10 CFR 50.155 NUSCALE POWER, LLC

2

1.0 INTRODUCTION

1.1 Summary By letter dated September 11, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23254A360), NuScale Power, LLC (NuScale), submitted, for U.S Nuclear Regulatory Commission (NRC) staff review and approval, Topical Report (TR)

TR-141299-P, Revision 0, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155 (Ref. 1). The NRC staff conducted an audit for TR-141299-P, Revision 0, starting on December 20, 2023 (Ref. 2) and concluding on March 30, 2024. On June 26, 2024, NuScale submitted Revision 1 of TR-141299-P (Ref. 3), hereafter referred to as the TR.

1.2 Scope of the NRC Staffs Review The purpose of the TR, as stated in its section 1.1, Purpose, is to describe the NuScale plant design capability to mitigate beyond-design-basis events (BDBEs) as defined in Title 10 of the Code of Federal Regulations (CFR), section 50.155 Mitigation of beyond-design-basis events, specifically, (i) the plant response to the loss of all alternating current power concurrent with loss of normal access to the normal heat sink and (ii) the design capability to mitigate the loss of large plant areas due to explosions or fire.

NuScale requested NRC review and approval of the TR for NuScale Power Plant capability to mitigate beyond-design-basis events and stated in TR section 1.3, Conditions of Use, that an adopter of the TR must provide (a) plant specific design information that includes the (( described in the TR, (b) a plant specific thermal analysis demonstrating the plants capability to cope with BDBEs with (( }} equipment identified in the TR, (c) a maintenance rule program in accordance with 10 CFR 50.65, requirements for monitoring the effectiveness of maintenance at nuclear power plants, and (d) an emergency plan in accordance with 10 CFR 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities or 10 CFR 50.47(b), and appendix E to Part 50, Emergency Planning and Preparedness for Production and Utilization Facilities. The NRC staffs review of the TR was limited to the generally non-specific information provided in the TR, and the staffs approval of the TR is subject to the limitations and conditions listed in section 5.0, Limitations and Conditions, of this report. Accordingly, future applicants who wish to utilize the TR for their plants will be required to provide additional information, detailed in the limitations and conditions in section 5.0, in order to receive NRC approval of the use of the topical report in their applications.

2.0 BACKGROUND

2.1 Regulatory Requirements and Relevant Regulatory and Industry Guidance The TR was developed to describe the NuScale design capabilities to mitigate beyond-design-basis events as defined by 10 CFR 50.155.

3 Applicable Regulations 10 CFR 50.155 requires applicants and licensees subject to 10 CFR Part 50, Domestic licensing of production and utilization facilities, and all applicants and licensees for a power reactor combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants to develop, implement, and maintain strategies and guidelines to mitigate a beyond-design-basis event. Related Guidance The staff used the following guidance during the review of the TR: Regulatory Guide (RG) 1.226, Flexible Mitigation Strategies for Beyond-Design-Basis Events, Revision 0 (Ref. 4, ML19058A012), identifies methods and procedures to demonstrate compliance with 10 CFR 50.155 Nuclear Energy Institute (NEI) 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Revision 4 (Ref. 5, ML16354B421), which is endorsed by RG 1.226, provides industry guidance in meeting 10 CFR 50.155 SECY-19-0066, Staff Review of NuScale Powers Mitigation Strategy for Beyond-Design-Basis External Events (Ref. 6, ML19148A443) NUREG-0800, Standard Review Plan (SRP) Section 19.4, Strategies and Guidance to Address Loss of Large Areas of the Plant due to Explosions and Fires (ML13316B202) 2.2 Summary of Technical Information The TR contains a description of proposed NuScale design capabilities and features to mitigate BDBEs as defined by 10 CFR 50.155. Design features that provide enhanced capabilities for coping with an extended loss of electrical power, loss of normal access to the normal heat sink, and loss of large areas due to explosions or fire are discussed within the TR. These features include the use of passive safety systems capable of maintaining core cooling, containment, and spent fuel cooling functions and a large reactor pool serving as the ultimate heat sink for the facility. The TR specifies how these features enable a design to mitigate BDBEs using (( }} plant equipment for a specified extended duration without the need for alternating current power, special equipment, or additional guidelines and strategies. NuScale states that the design features described in the TR can maintain core cooling, containment, and spent fuel pool (SFP) cooling using (( }}. The TR does not provide thermal analyses; instead, the TR states that an adopter of the TR must provide plant specific thermal analysis demonstrating its capability to maintain core cooling, containment, and SFP cooling for (( }} identified in the TR.

3.0 TECHNICAL EVALUATION

NuScale states that design characteristics include the following: Containment cooling is designed to be maintained for at least (( }} without pool inventory makeup or operator action. Therefore, the capability of the containment cooling supports the ability of the design to (( 

}}

4 (( }} without predetermined supplemental actions. The installed equipment is listed in TR table 4-1, (( }} and is described in TR section 4.0, Plant Systems and Responses to a Loss of All Alternating Current Power Event. In TR section 4.0, NuScale also states that (( }} those described in the TR. This statement is consistent with the docketed response for audit issue, MBDBE.LTR-8 (Ref. 7), which clarified that (( }}. The staff understands these statements to mean (( }} in the design bases for plants adopting this TR. The TR does not contain any supporting analyses or references to such analysis showing the basis for the above determination. In response to staffs audit questions, NuScale revised the TR to include four conditions of its use as stated in TR section 1.3: To adopt the described methodology for the response to a beyond-design-basis external event (BDBEE), an adopter of the topical report must provide: (1) A plant specific design that (( }} described within this report. (2) A plant specific thermal analysis demonstrating (( }}. This analysis addresses site specific conditions, including configuration of the plant with respect to the selected number of modules and spent fuel pool capacity, for all modes of operation (normal and refueling). (3) A maintenance rule program in accordance with 10 CFR 50.65. (4) An emergency plan in accordance with 10 CFR 50.160 or 50.47(b) and 10 CFR Part 50 Appendix E describing communications and coordination with local, state, federal, and tribal agencies. As written in Limitation and Condition no. 5.1, the staff determined that these same conditions of use are applicable. Furthermore, in Limitation and Condition no. 5.1 the staff clarifies that in order to satisfy Condition of Use #1 an applicant or licensee must provide a plant specific design that includes (( }} with the described system functions listed in TR Table 4-1 and the system design features and equipment classifications as described in TR section 4.0 - 4.15 for each system In TR section 3.2.4, Monitoring, NuScale states that (( }}. NuScale design capabilities and features to mitigate BDBEs, as defined by 10 CFR 50.155, are based on a NuScale Power Modules (NPM) ability to (( }}. The SFP is part of the UHS. The TR is intended to be generically applicable to NuScale reactor designs with the capabilities and features discussed in the TR. NuScale has not provided supporting analyses to demonstrate the NPMs ability to withstand beyond design basis events in the TR, and a supporting analysis to demonstrate these capabilities is therefore required to be provided by an applicant or a licensee who adopts the TR.

5 The potential for the future submittal of various design features as well as the need for analysis to be provided by a COL applicant has caused the staff to impose limitations and conditions for using the TR, as provided in section 5.0. The NRC staffs evaluation of TR sections 3.0, Plant Baseline Coping Criteria for Loss of all AC Power, through 9.0,Spent fuel pool monitoring after final fuel removal from the reactor vessel, as required by 10 CFR 50.155(e), is provided in sections 3.1, Plant Baseline Coping Criteria for Loss of all AC Power, through 3.7, SFP monitoring after final fuel removal from the reactor vessel, as required by 10 CFR 50.155(e), respectively, of this report. 3.1 Plant Baseline Coping Criteria for Loss of all AC Power 3.1.1 Assessment of Electrical Power In the TR, section 1.2, Scope, NuScale states that the report is applicable to NuScale small modular reactor designs that have structures, systems and components (SSCs) capable of performing their safety functions without off-site electric power or operator actions following a BDBE. In the TR, section 3.1.5, Initial Event Conditions and Assumptions, NuScale states that for the baseline coping capability, 1) station batteries and associated direct current (DC) buses remain available for the designed operating time of the station batteries and 2) installed electrical distribution system, including inverters and battery chargers, remain available provided they are seismic Category I. The initial conditions and assumptions in NEI 12-06, Revision 4 (Ref. 5), and RG 1.226, Revision 0 (Ref. 4), assume that station batteries would remain available following a BDBE since they are considered robust. While the augmented direct current system (EDAS) batteries are not safety related, NuScale states, in section 4.3.2, Equipment Qualification, of the TR, that the EDAS SSCs are located in Seismic Class I structures. Seismic category I buildings are designed to withstand design basis external events and to provide safety margins and conservatism to make structural failure unlikely. By placing station batteries inside seismic category I structures, they will be reasonably protected against natural hazards. Therefore, the staff finds that the initial assumptions described in TR following BDBEE are consistent with the NEI 12-06, Revision 4, guidance, which is endorsed by RG 1.226. In section 3.2, Plant Design Capabilities of the TR, NuScale states that the first 72 hours of a loss of all alternate current (AC) power is identical to a station blackout (SBO) and no AC power is relied upon for performing safety functions. Following a loss of all AC power event and following the automatic response of safety-related equipment, NuScale states, in section 3.2 of the TR, that (( }}. In section 4.3, Augmented Direct Current Power System of the TR, NuScale states that EDAS provides power to plant loads including the module protection system (MPS), plant protection system (PPS), and safety display and indication system (SDIS) and augmented direct current power system - common (EDAS-C) services plant common loads, including main control room (MCR) emergency lighting and post-accident monitoring (PAM) variable indications displayed in

6 the MCR. An applicant must satisfy Limitation and Condition no. 5.1, which states that an applicant using the TR must meet TR Section 1.3, Conditions of Use. In this regard, Condition of Use #1 requires an adopter of the TR to provide a plant specific design that (( }} described within the TR. Limitation and Condition no. 5.1 expands on this requirement by specifying that the adopter must also ensure the plant design includes the described system functions, system design features, and equipment qualification as described in the TR. Further, Condition of Use #2 states that an adopter of the TR must provide a plant specific thermal analysis demonstrating maintenance of core cooling, containment and spent fuel cooling (( }}. By satisfying this limitation and condition the augmented quality, capacity, and capability of the EDAS to provide power for a BDBE would be demonstrated. In section 3.2.4 of the TR, NuScale states that (( }}. The staff notes that UHS and SFP level instruments can be powered by EDAS, (( }} and therefore, an applicant must satisfy Limitation and Condition no. 5.2. By satisfying this condition, an applicant would be able to address (( }}. Accordingly, the staff has determined that the description of the electrical power supply for the SFP level instrumentation is consistent with the guidance provided by RG 1.227, Wide-Range Spent Fuel Pool Level Instrumentation, Revision 0 (Ref. 8). Therefore, an applicant or licensee following this TR and satisfying Limitation and Condition no. 5.2, would be able to demonstrate that its use of the NuScale design meets 10 CFR 50.155(e) as it relates to power supplies for the first 72 hours following a BDBE. In section 4.6.1, [SDIS] System Design, of the TR, NuScale states that the SDIS provides accident monitoring functions, and that electrical power is provided to the SDIS from two separate and independent divisions of EDAS-C. In section 4.3.1, [EDAS] System Design, of the TR, NuScale states that EDAS-C power divisions have a specified minimum battery duty cycle of 72 hours. The NRC staff concludes that the approach in the TR, which outlines the NuScale electrical power system design, would be acceptable to ensure that electrical power is supplied to electrical equipment (e.g., instrumentation, lighting, emergency core cooling system (ECCS) solenoid valves) for a BDBE leading to a loss of all AC power, provided the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(e). 3.1.2 Plant Design Capabilities The TR describes plant design capabilities for core cooling, containment, SFP cooling, and monitoring following a loss of all AC power event. TR section 3.2 states the following:

7 Following a loss of all AC power event, automatic responses of safety-related equipment establish and maintain the key safety functions of core cooling, containment, and SFP cooling by placing the reactor modules and spent fuel into a safe, stable, shutdown state with passive cooling. (( }}. Core coolingduring a loss of all AC power event, containment isolation within (( }} of the event preserves the reactor coolant inventory. The decay heat removal system (DHRS) passively removes decay heat for up to the first 24 hours following a loss of all AC power event. By 24 hours, the ECCS valves automatically open and the ECCS maintains core cooling for the extended loss of all AC power event. ContainmentAs stated in TR section 3.2, the safety-related containment isolation valves (CIVs) and the containment vessel (CNV) provide passive containment isolation function without operator action or electrical power. Heat removal to the UHS passively controls temperature and pressure to ensure containment integrity. Peak pressure and temperature conditions for the CNV are designed to occur early in the event when the ECCS valves open and prevent a challenge to containment integrity. Containment cooling is designed (( }}. SFP CoolingAs stated in TR section 3.2, the SFP, as part of the UHS, communicates with the refueling pool and reactor pool above the SFP weir wall. As such, the pools respond as a single volume during a loss of all AC power event until the UHS level lowers to the weir wall. The UHS inventory is designed to maintain passive cooling of the spent fuel in the SFP (( }}. MonitoringAs stated in TR section 3.2, (( }}. However, post-accident monitoring variable indications are maintained in the main control room for at least 72 hours to provide additional assurance that systems respond as designed. The NRC staff concludes that the approach outlined in the TR would be acceptable to ensure plant design capabilities for core cooling, containment, SFP cooling, and monitoring following a loss of all AC power event, provided the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(1) for core cooling, containment, and SFP cooling and 10 CFR 50.155(b)(2)(e) for spent fuel pool monitoring. 3.2 Plant Systems and Responses to a Loss of All Alternating Current Power Event Section 4.0, Plant Systems and Responses to a Loss of All Alternating Current Power Event, of the TR describes individual system responses to the event in order to provide an overview of the integrated plant response. The staff reviewed the descriptions of systems in accordance with 10 CFR 50.155.

8 10 CFR 50.155(c) states the following: (c) Equipment. (1) The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must have sufficient capacity and capability to perform the functions required by paragraph (b)(1) of this section. (2) The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must be reasonably protected from the effects of natural phenomena that are equivalent in magnitude to the phenomena assumed for developing the design basis of the facility. The system qualification and availability for the NuScale design are laid out in TR section 4.0. As described in the TR, systems used to support the mitigating strategies are protected in seismic Category I structures and include protection from other applicable design basis hazards (such as wind and flood events) and are assumed to survive the BDBEE. Table 4-1 in the TR provides a summary of the functions of the (( }} and locations that are relied upon for mitigation of the BDBEE. The NRC staff concludes that the approach in the TR, which outlines the NuScale plant systems and responses to the BDBEE, would be acceptable to ensure that the plant systems are protected from BDBEE and should continue to perform to support their safety functions, provided the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(c). 3.3 Safety Functions during a Loss of All Alternating Current Power 3.3.1 Systems and Safety Functions Section 5.0, Safety Functions during a Loss of All Alternating Current Power, of the TR describes the safety functions for a loss of all AC power event. Section 5.0 of the TR describes each system and the safety function that it performs. The safety functions include Core Cooling, Containment, and SFP cooling. 10 CFR 50.155(b)(1) states, in part, the following: (b) Strategies and guidelines. Each applicant or licensee shall develop, implement, and maintain; (1) Mitigation strategies for [BDBEEs] These strategies and guidelines must be capable of being implemented site-wide and must include the following: (i) Maintaining or restoring core cooling, containment, and spent fuel pool cooling capabilities; In Section 5.1, Integrated Plant Response, the TR explains how the NuScale design is based on passive systems whose safety functions can be performed without operator intervention when initiated from 100 percent power. The TR describes how the safety functions are met during non-power operation modes as well. Initial coping aligns with the typical plant performance during a loss of all AC power. Indefinite coping is covered in section 3.3.2 of this SE.

9 The core cooling safety function was discussed in TR section 5.2, Core Cooling. NuScale states that reactor coolant inventory is maintained as the NuScale Power Module (NPM) design would be isolated and inventory would be maintained within the containment vessel. The NPM passive plant design does not include reactor coolant pumps and therefore there is no Reactor Coolant Pump (RCP) seal leakage. The isolation of the CNV allows the water level to remain above the top of the active fuel. The potential exists for the addition of water when necessary. NuScale states that reactivity control is maintained to continue safe shutdown conditions during the event, and the NPM design allows for shutdown through reactor trip and the control rod insertion. (( }}. NuScale states in the TR that [d]uring the loss of all AC power (( }}. In SECY-19-0066, the staff stated that if the staff determines that there are no credible transient phenomena (e.g., return to power) that could challenge core cooling, containment, or SFP cooling beyond 72 hours, then no additional review or approval of these capabilities would be required at the COL stage. In Limitation and Condition no. 5.1, the staff incorporates TR section 3, Conditions of Use; TR Condition of Use #2 states that an adopter of the TR must provide a plant specific thermal analysis demonstrating its capability to (( }}. Therefore, consistent with SECY-19-0066, the staff determines that subject to Limitation and Condition no. 5.1, an applicant or licensee referencing this TR would be required to address any credible transient phenomena (e.g., return to power) that could challenge core cooling, containment, or SFP cooling. NuScale states that decay heat removal is passively accomplished in the NPM. NuScale explains that the DHRS allows the UHS inventory to remove heat from the CNV. The DHRS works in conjunction with the ECCS to remove the decay heat. Eventually, the UHS will begin to boil. The water level will then lower in the UHS. Eventually, water can be added to the UHS, if and when needed, through protected piping. NuScale states that SFP cooling is accomplished through heat transfer to the UHS. The heat from the spent fuel is transferred to the SFP which is initially connected to the UHS. As the UHS boils, the SFP can become disconnected from the UHS by the SFP weir wall. Eventually, if and when needed, water can be added to the UHS which also adds water to the SFP. The NRC staff concludes that the approach outlined in TR section 5.0, which includes a summary of the core cooling, containment, and SFP cooling functions in TR Table 5-3, Baseline Coping Capability Summary, would be acceptable to ensure the NPM would be able to perform extended coping, provided the applicant satisfies Limitation and Condition 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition, providing a plant specific design and analysis to support the expected coping period, including all modes of operation, when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(1). 3.3.2 Indefinite maintenance of core cooling, containment, and spent fuel pool cooling capabilities

10 The regulation at 10 CFR 50.155(b)(1) states that each applicant or licensee shall develop, implement, and maintain: Mitigation strategies for beyond-design basis external eventsStrategies and guidelines to mitigate beyond-design-basis external events from natural phenomena that are developed assuming a loss of all ac power concurrent with either a loss of normal access to the ultimate heat sink or, for passive reactor designs, a loss of normal access to the normal heat sink. These strategies and guidelines must be capable of being implemented site-wide and must include the following: (i) Maintaining or restoring core cooling, containment, and spent fuel pool cooling capabilities; and (ii) The acquisition and use of offsite assistance and resources to support the functions required by paragraph (b)(1)(i) of this section indefinitely, or until sufficient site functional capabilities can be maintained without the need for the mitigation strategies. The statements of consideration for 10 CFR 50.155 in the Federal Register (84 FR 39684; August 9, 2019) (Ref. 9) state, in part (at 39709): The requirement to enable the acquisition and use of offsite assistance and resources to support the functions required by § 50.155(b)(1)(i) of this section indefinitely, or until sufficient site functional capabilities can be maintained without the need for the mitigation strategies means that licensees need to plan for obtaining sufficient resources (e.g., fuel for generators and pumps, cooling and makeup water) to continue removing decay heat from the irradiated fuel in the reactor vessel and SFP as well as to remove heat from containment as necessary until an alternate means of removing heat is established. ...More detailed planning for offsite assistance and resources is necessary for the initial period following the event; less detailed planning is necessary as the event progresses and the licensee can mobilize additional support for recovery. Section 3.3, Considerations in Utilizing Off-Site Resources, of NEI 1206, Revision 4, states, in part: Since site access is considered to be restored to near-normal within 24 hours, by 72 hours from the event initiation, outside resources should be able to be mobilized by that time such that a continuous supply of needed resources will be able to be provided to the site. Within these first 72 hours a site will have deployed its FLEX strategies which should result in a stable plant condition on the FLEX equipment and plans will have been established to maintain the key safety functions for the long term. Therefore, FLEX strategies and/or resources are not required to be explicitly planned in advance for the period beyond 72 hours.

11 The site will need to identify staging area(s) for receipt of the off-site FLEX equipment and a means to transport the off-site equipment to the deployment location. It is expected that the licensee will ensure the off-site resource organization will be able to provide the resources that will be necessary to support the extended coping duration. In addition, the licensee will need to ensure standard connectors for electrical and mechanical FLEX equipment compatible with the site connections are provided. Section 1.1.1.3, Final Phase, of RG 1.226, Revision 0, states: The final phase will be accomplished using the onsite equipment augmented with additional equipment and consumables obtained from off-site until power, water, and coolant injection systems are restored or commissioned. Staff Position: NEI 12-06, Revision 4, Section 3.0, provides an acceptable method for determining the baseline coping capabilities for the final phase. NEI 12-06, Revision 4, Section 12.2, provides an acceptable method for establishing the capability to obtain equipment and consumables from off-site until power, water, and coolant injection systems are restored or commissioned. This provides an acceptable method to sustain the listed functions indefinitely when coupled with the restoration or commissioning of power, water, and coolant injection systems. The NRC-endorsed guidance in NEI 12-06, Revision 4, section 3.3, recognizes that site access is expected to be restored within 24 hours from the event initiation and that off-site resources should be able to be mobilized by 72 hours such that FLEX strategies and/or resources are not required to be explicitly planned for the period beyond 72 hours. However, this guidance also presumes that the licensee will identify staging areas to receive off-site resources and the means to transport the equipment to areas where it is to be deployed, that the licensee will ensure the ability of an offsite organization to provide the necessary resources to support the extended coping duration, and that standard connectors for electrical and mechanical FLEX equipment that are compatible with the site connections are obtained. The staff position in RG 1.226, section 1.1.1.3 assumes that various mitigating strategies have been implemented prior to long-term coping such that the site has established the capability to receive and utilize future, unplanned resources. Therefore, some degree of planning and preparation will be needed (( }} to ensure that unplanned off-site resources can be identified, obtained, and implemented at the site (( }}. 3.3.2.1 UHS Make Up 10 CFR 50.155(b)(1)(ii) requires the capability to acquire and use off-site assistance and resources to support the functions described in 10 CFR 50.155(b)(1)(i) indefinitely, or until mitigation strategies are no longer needed. The statements of consideration for 10 CFR 50.155

12 make it clear that licensees need to plan for obtaining sufficient resources to maintain these mitigating capabilities until alternate means of heat removal are established. The TR executive summary states the following: The indefinite core cooling, containment, and spent fuel capabilities are supported by the design of the UHS [ultimate heat sink]. A plant operator has the specified timeframe after the initiation of a beyond-design-basis external event (BDBEE) to provide replenishment of the UHS water level. The LTR describes the specified timeframe as (( }}. Therefore, the TR submittal acknowledges that operator actions will be required (( }}. Therefore, the conditions described in the TR are not consistent with the conditions assumed to be in place at the site by the statements of consideration for 10 CFR 50.155 and the staff position in RG 1.226, Section 1.1.1.3 related to the establishment of the capability to receive and implement future, unplanned resources. The NRC staff concludes that the TR does not resolve this issue and, accordingly, an applicant or licensee that references this TR must satisfy Limitation and Condition no. 5.2 to resolve this issue. Limitation and Condition 5.2 requires an applicant to provide preplanned mitigating actions to ensure that long term coping requirements regarding UHS make up are satisfied or provide a satisfactory justification showing that such actions are not required. 3.3.2.2 Control Room Egress Per section 4.12.1, System Design, in the TR, breathable air is credited to be available to the control room operators for 72 hours. No other provisions are established to sustain control room operators during this initial 72-hour period. Once this breathable air is depleted, the control room will become unhabitable, and the operators will be required to egress the control room. Debris may block egress of the control room due to the BDBEE. However, no provision is described for preplanning of debris removal. Therefore, the NRC staff concludes that an applicant or licensee that references this TR must satisfy Limitation and Condition 5.3 to resolve this issue. Limitation and Condition 5.3 requires that preplanned mitigating actions are established to ensure the sustainability of control room operators for 72 hours post-BDBEE, as described in the TR, in compliance with 10 CFR 50.155(b)(1)(i), and to then egress the control room when breathable air is depleted. Alternatively, a justification must be provided that supports no preplanning for these mitigating actions. 3.3.2.3 Determining Plant Conditions Post-BDBEE Section 5.3.2, Containment Process Variables, of the TR states, in part:

13 Per baseline coping capability of Section 3.1.2 [,Baseline Coping Capability Criteria, Conditions, and Assumptions,] the instrumentation associated with each process variable is assumed to survive the BDBEE and remain fully available for a duration beyond the time necessary for the associated mitigation function to be established and monitored. Section 5.3.2 indicates that the available duration is limited to confirming that safety systems have actuated to their passive operating configuration during the initial coping phase. Once offsite support arrives, knowledge of current plant conditions will be necessary to determine appropriate mitigating actions. Therefore, the NRC staff concludes that an applicant or licensee that references this TR must satisfy Limitation and Condition no. 5.4 to resolve this issue. Limitation and Condition no. 5.4 requires that preplanned mitigating actions are established to ensure that site support personnel can ascertain plant conditions to determine necessary plant coping requirements once on-site instrumentation power systems are depleted. Alternatively, a justification must be provided that supports not preplanning for these mitigating actions. 3.3.2.4 SFP Level Monitoring Section 3.2.4, Monitoring, of the TR states the following, in part: (( }} Depending on the nature of the BDBEE, (( }}. Therefore, the NRC staff concludes that an applicant or licensee that references this TR must satisfy Limitation and Condition no. 5.5 to resolve this issue. Limitation and Condition no. 5.5 requires that an applicant or licensee referencing this TR must address (e.g., in plant procedures) how plant operators will ensure that any required debris removal will be accomplished to allow plant access to support replacement of SFP level monitoring instrumentation power supply equipment, as described in Section 9.0 of the TR, and that a preplanned source is established to provide the power supply equipment when required. Alternatively, a justification must be provided that supports no preplanning for these mitigating actions. 3.3.2.5 Conclusion The NRC staff concludes that the approach outlined in the TR would be acceptable to ensure long-term coping capability after a beyond-design-basis event in conjunction with the implementation of the conditions detailed above, provided the applicant satisfies the limitations and conditions established herein. An applicant or licensee using a NuScale design will need to implement the limitations and conditions mentioned in the above sections when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(1). 3.4 Capability to respond to a loss of large areas (LOLA) due to explosions or fire, as required by 10 CFR 50.155(b)(2)

14 10 CFR 50.155(b)(2) requires licensees to develop and implement strategies and guidelines to maintain or restore core cooling, containment, and SFP cooling capabilities under the circumstances associated with LOLA of the plant due to explosions or fire. Strategies and guidelines must address in a three-phase approach: Phase I - Enhanced firefighting capabilities Phase II - Measures to mitigate damage to fuel in the SFP, and Phase III - Measures to mitigate damage to fuel in the reactor vessel and to minimize radiological release. NuScale stated that the TR follows guidance in NUREG-0800, section 19.4, Strategies and Guidance to Address Loss of Large Areas of the Plant due to Explosions and Fires, (ML13316B202), which directs new plants to implement guidance in the February 5, 2005 Temporary Instruction 2515/168, Developing Mitigating Strategies/Guidance for Nuclear Power Plants to Respond to Loss of Large Areas of the Plant in Accordance with B.5.b of the February 25, 2002, Order, and Nuclear Energy Institute (NEI) 06-12, B.5.b Phase 2 & 3 Submittal Guideline, in addressing beyond-design-basis events (e.g., LOLA). 3.4.1 Phase 1 - Enhanced firefighting capabilities NuScale stated that NEI 06-12 guidance for firefighting response to a LOLA event includes operational aspects of responding to explosions or fire including prearranging for involvement of outside organizations, planning and preparation activities (e.g., pre-positioning equipment, personnel, and materials to be used for mitigating the event), and developing procedures and training for managing the event. NuScale also stated that it incorporates the following design features to cope with potential fires that could affect module or plant safety: Redundant safety systems to perform safety-related functions, such as reactor shutdown and core cooling Physical separation between redundant trains of safety-related equipment used to mitigate the consequences of a design-basis accident Passive design that minimizes the need for support systems and the potential effects of "hot shorts Annunciation of fire indication in the main control room to facilitate personnel response No electrical power requirement for mitigating design-basis events as safety systems are fail-safe on loss of power The NRC staff concludes that the approach outlined in the TR would be acceptable to ensure enhanced firefighting capabilities to respond to a LOLA due to explosions or fire, provided the applicant satisfies Limitation and Condition no. 5.6. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(2), 10 CFR 50.48, and GDC 3. 3.4.2 Phase 2 - Measures to mitigate damage to fuel in the SFP NuScale stated that the SFP is below grade and the walls are designed to seismic Category I requirements and are completely contained within the seismic Category I Reactor Building (RXB). NuScale also stated that all pipe connections to and from the pool are at an elevation below the normal operating level but above the minimum pool level required for SFP radiation

15 shielding and heat removal or are protected by a siphon break which prevents inadvertent lowering of the pool level below safety limits. The NRC staff concludes that the approach in the TR, which outlines that the SFP is below grade and designed such that the SFP cannot be drained below safety limits, would be acceptable to ensure enhanced SFP capability to mitigate damage to fuel in the SFP in response to a LOLA due to explosions or fire, or demonstrate this enhanced capability is not required, provided the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(2). 3.4.3 Phase 3 - Measures to mitigate damage to fuel in the reactor vessel and to minimize radiological release NEI 06-12 guidance for extensive damage mitigation was developed based on pressurized water reactor plant key safety functions, which includes reactor coolant system (RCS) inventory control, RCS heat removal, containment isolation, containment integrity, and release mitigation. In section 6.4.1 of the TR, Assessment of Key Safety Functions, NuScale described the key design features to achieve the above safety functions. The assessment for each safety function is summarized as follows: RCS Inventory Control - NuScale stated that the purpose of this key safety function is to ensure that the core is covered with water. NuScale stated that containment system is utilized as the primary means for RCS inventory control. NuScale further stated that the design does not have RCPs, and therefore, there is no potential for loss of inventory through RCP seals due to lack of seal cooling. NuScale also stated that leakage rate through CIVs is small, and makeup is not required. RCS Heat Removal - NuScale stated that the purpose of this key safety function is to remove the decay heat from the core and transfer it the UHS. NuScale stated that the primary means for heat removal during steady state, startup and hot shutdown operations is through the steam generators. NuScale further stated that the alternate means for RCS heat removal is the passive DHRS or the ECCS and that during DHRS or ECCS operations, no electrical AC power or external feedwater injection is required. Containment Isolation - NuScale stated that the purpose of this key safety function is to ensure no leakage paths exist that would allow gaseous and particulate radiation to escape containment. NuScale further stated that CIVs and the CNV are utilized to accomplish this function, and that the CIVs are energized open so a loss of DC power to those valves will result in their repositioning to their safe or accident response position. Containment Integrity - NuScale stated that the purpose of this key safety function is to ensure the containment fission product barrier is maintained to minimize or prevent radiological release outside containment. NuScale further stated that passive heat removal to the UHS controls temperature and pressure to ensure containment integrity. Release Mitigation - NuScale stated that the purpose of this key safety function is to minimize radiological release assuming severe core damage occurs, and a radiological release is imminent or in progress. NuScale stated that the CNV is an American Society of Mechanical Engineer (ASME) Boiler and Pressure Vessel (B&PV) Code Section III Class I pressure vessel forming a barrier to prevent uncontrolled release of radiological materials and radiological

16 contaminants. NuScale further stated that the reactor pressure vessel is located within the CNV, and the CNV is partially immersed in the UHS, and that the UHS is the primary means to perform the release mitigation function. The NRC staff concludes that the approach outlined in the TR would be acceptable to ensure that key safety functions to mitigate potential fuel damage and radiological release are accomplished, provided that the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(b)(2). 3.4.4 Conclusion The NRC staff concludes that the approach outlined in the TR would be acceptable to ensure adequate LOLA coping capability (( }} per NEI 06-12 guidance, provided the applicant satisfies the limitations and conditions detailed above. An applicant or licensee using a NuScale design will need to implement these limitations and conditions to demonstrate compliance with 10 CFR 50.155(b)(2). 3.5 Capacity, capability, and protection of equipment associated with mitigation of events described in the rule, as required by 10 CFR 50.155(c) 10 CFR 50.155(c) states the following: (c) Equipment. (1) The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must have sufficient capacity and capability to perform the functions required by paragraph (b)(1) of this section. (2) The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must be reasonably protected from the effects of natural phenomena that are equivalent in magnitude to the phenomena assumed for developing the design basis of the facility. TR section 7.0, Capacity, capability, and protection of equipment associated with mitigation of events described in the rule, as required by 10 CFR 50.155(c), states the following: (( }}

17 The staff has determined that future applicants that reference this TR would need to show that their proposed strategy for NPM designs provides for extended coping using (( }} that is reasonably protected from the effects of natural phenomena. As stated in Limitation and Condition no. 5.1, an applicant or licensee referencing the TR must meet the Conditions of Use listed in Section 1.3 of the TR, which details, in part, that (1), an applicant or licensee referencing the TR must provide a plant specific design that includes the (( }} [and] (2) provide a plant specific thermal analysis demonstrating (( }}. Therefore, the NRC staff concludes that the approach in the TR would be acceptable, provided the applicant satisfies Limitation and Condition no. 5.1. An applicant or licensee using a NuScale design will need to implement this limitation and condition when adopting this TR to demonstrate compliance with 10 CFR 50.155(c). 3.6 Training requirements as defined by 10 CFR 50.155(d) 10 CFR 50.155(d) states Each licensee shall provide for the training of personnel that perform activities in accordance with the capabilities required by paragraphs (b)(1) and (2) of this section. In section 8.0 Training requirements as defined by 10 CFR 50.155(d) of the TR, NuScale states the following: (( }} As stated in Conditions of Use no. 2 of TR section 1.3, which is adopted in Limitation and Condition no. 5.1, an applicant or licensee that references the topical report must provide a plant specific thermal analysis demonstrating (( }} Limitation and Condition no. 5.7 states that an applicant or licensee referencing this TR must ensure that the training program for site staff that covers (( }} as described in section 9.0 of the TR. Therefore, the NRC staff concludes that the approach outlined in the TR would be acceptable to ensure adequate training requirements, provided the applicant satisfies Limitation and Condition nos. 5.1 and 5.7. An applicant or licensee using the NuScale design will need to implement these limitations and conditions when adopting this TR to demonstrate compliance with 10 CFR 50.155(d). 3.7 SFP monitoring after final fuel removal from the reactor vessel, as required by 10 CFR 50.155(e) 10 CFR 50.155(e) states:

18 (e) Spent fuel pool monitoring. In order to support effective prioritization of event mitigation and recovery actions, each licensee shall provide reliable means to remotely monitor wide-range water level for each spent fuel pool at its site until 5 years have elapsed since all of the fuel within that spent fuel pool was last used in a reactor vessel for power generation. This provision does not apply to General Electric Mark III upper containment pools. RG 1.227, Rev. 0, provides guidance for satisfying the requirements of 10 CFR 50.155(e). This RG endorses, with exceptions and clarifications, the methods and procedures promulgated by NEI in NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Revision 1 (NEI 12-02) dated August 2012 (Ref. 10) as a process the NRC staff considers acceptable for meeting certain requirements in 10 CFR 50.155. The TR indicates that the level instrumentation relative to 10 CFR 50.155(e) (four instruments) are seismically mounted, environmentally qualified, and designed to meet the guidance of NEI 12-02. The level instruments are also provided (( }}. The staff evaluated the applicants description of the level instrumentation provided in the TR and finds that it meets the design criteria recommended in NEI 12-02. The staff also evaluated the level instrumentation power requirements and determined that (( }} would provide sufficient capacity to maintain the level indication function until offsite resource availability is reasonably assured. NEI 12-02 section 4.1 Training, indicates that: Procedures will be developed using guidelines and vendor instructions to address the maintenance, operation and abnormal response issues associated with the new SFP instrumentation. The TR section 8.0 Training requirements as defined by 10 CFR 50.155(d) states that: (( }} As discussed above in section 3.6, the staff finds it would be acceptable to credit the training program for site staff that covers (( }}, as stated in Section 9.0 of the TR and in accordance with Limitation and Condition no. 5.7. The staff finds that level instrumentation designed in accordance with NEI 12-02 and provided with a dedicated backup battery supply, as described in the TR, as well as the capability to allow use of procured offsite equipment, would allow a future applicant to demonstrate it meets the design criteria discussed in RG 1.227. Therefore, an applicant or licensee using this TR in a NuScale design would be required to satisfy Limitation and Condition No. 5.7 to demonstrate compliance with the requirements of 10 CFR 50.155(e).

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4.0 CONCLUSION

Based upon its review as discussed above, subject to the limitations and conditions as described in section 5.0 of this SE, the NRC staff concludes that an applicant or licensee could use TR-141299-P, Revision 1 to demonstrate the NuScale plant designs capability to mitigate BDBEs as defined by 10 CFR 50.155. If an applicant for an operating license under 10 CFR Part 50, or an applicant for a combined license under 10 CFR Part 52, is not able to demonstrate compliance with an NRC regulation when the plant specific design is complete, the applicant would be required to justify an exemption from the applicable regulatory requirement. The NRC staff will evaluate the regulatory compliance of a plant specific design during future licensing reviews conducted in accordance with 10 CFR Part 50 or 10 CFR Part 52, as applicable. As discussed in the TR, the TR could be applied generically, therefore the final design with which this TR may be utilized is currently unknown. The NRC staff will make a final determination of the acceptability of an applicants compliance with 10 CFR 50.155 during future licensing activities when the detailed design is complete as outlined in an operating license or combined license application that references this TR.

20 5.0 LIMITATIONS AND CONDITIONS The staffs approval is limited to the application of this methodology to the NuScale reactor design, specified in TR Section 1.2, Scope, with the following limitations and conditions: 5.1 An applicant or licensee referencing the TR must meet the Conditions of Use listed in section 1.3. To satisfy Condition of Use #1 an applicant or licensee must provide a plant specific design that includes (( }} with the described system functions listed in TR Table 4-1 and the system design features and equipment classifications as described in TR section 4.0 - 4.15 for each system. 5.2 An applicant or licensee referencing the TR must address (e.g., in plant procedures) how plant operators will ensure, during the initial coping phase, that the following actions can be achieved to provide inventory makeup to the UHS at the start of the final long-term coping phase: (1) a source of water can be identified and will be available in sufficient quantity (2) the necessary motive equipment such as pumps and generators, and the required electrical power/fuel, can be obtained, staged, and implemented and (3) any required debris removal will be accomplished to support placement of equipment and access to site connections. Alternatively, an applicant or licensee can justify why the plant-specific application requires no preplanning before a BDBEE to address these mitigating actions. 5.3 An applicant or licensee referencing the TR must address (e.g., in plant procedures) how control room operators will be sustained for 72 hours in the control room and subsequently exit the control room at 72 hours post-event once breathable air is depleted if debris blockage prevents egress from the control room. Alternatively, an applicant or licensee can justify why the plant-specific application requires no preplanning before a BDBEE to address these mitigating actions. 5.4 An applicant or licensee referencing the TR must address (e.g., in plant procedures) how site support personnel will ascertain plant conditions in order to determine necessary coping requirements during the initial phase, once on-site power systems are depleted, and at the start of the final long-term coping phase. Alternatively, an applicant or licensee can justify why the plant-specific application requires no preplanning before a BDBEE to address these mitigating actions. 5.5 An applicant or licensee referencing this TR must address (e.g., in plant procedures) how plant operators will ensure that any required debris removal will be accomplished to allow plant access to support replacement of SFP level monitoring instrumentation power supply equipment, as described in section 9.0 of the TR, and that a preplanned source is established to provide the power supply equipment when required. Alternatively, an applicant or licensee can justify why the plant specific application requires no preplanning before a BDBEE to address these mitigating actions. 5.6 An applicant or licensee referencing the TR must provide a Fire Protection Program in accordance with 10 CFR 50.48. 5.7 An applicant or licensee referencing this TR must ensure that the training program for site staff that covers installed plant equipment includes required activities related to replacement of the SFP level monitoring instrumentation power supply equipment as described in section 9.0 of the TR.

6.0 REFERENCES

1. NuScale Power, LLC, TR-141299, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155, Revision 0, September 11, 2023, (ML23254A360 (public) and ML23254A361 (non-public)).

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2. Memorandum from Tesfaye, G., NRC, to Jardaneh, M., NRC Audit Plan for the staff review of the NuScale generic licensing topical reports, December 20, 2023 (ML23349A078).
3. NuScale Power, LLC, TR-141299, NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events Defined by 10 CFR 50.155, Revision 1, June 26, 2024, (ML24178A398 (public) and ML24178A399 (non-public)).
4. Regulatory Guide (RG) 1.226, Flexible Mitigation Strategies for Beyond-Design-Basis Events, Rev. 0 (ML19058A012)
5. NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. 4 (ML16354B421)
6. NRC, SECY-19-0066, Staff Review of NuScale Powers Mitigation Strategy for Beyond-Design-Basis External Events, June 26, 2019 (ML19148A443)
7. NuScale Power, LLC, Response to NuScale Topical Report Audit Question, A-MBDBE.LTR-8 (ML24178A401 (public) and ML24178A403 (non-public)).
8. RG 1.227, Wide-Range Spent Fuel Pool Level Instrumentation, issued June 2019 (ML19058A013)
9. Federal Register Volume 84, No. 154, Mitigation of Beyond-Design-Basis Events, August 9, 2019, pages 39684 - 39709
10. Nuclear Energy Institute (NEI) in document NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Revision 1, dated August 2012}}
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