ML24263A004

Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report TR 141299 P, Revision 1 NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis. NON-PROP

ML24263A004
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Site:	99902078
Issue date:	09/26/2024
From:	Hayden T NRC/NRR/DNRL/NRLB
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EPID L-2023-TOP-0022
Download: ML24263A004 (19)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT TR-141299-P, REVISION 1 NUSCALE POWER PLANT DESIGN CAPABILITY TO MITIGATE BEYOND-DESIGN-BASIS EVENTS DEFINED BY 10 CFR 50.155 NUSCALE POWER, LLC DOCKET NO. 99902078, EPID NO. L-2023-TOP-0022 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 POSITION.

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 Reference (Ref.) 1. The NRC staff conducted an audit for TR-141299-P, Revision 0, starting on December 20, 2023 (Ref. 2) and it concluded 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 and Approval 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 states in TR section 1.3 Conditions of Use, that an adopter of the TR must provide (a) plant specific design information that includes the ((

)) plant equipment 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.

2 The NRC staffs review of the TR was limited to the 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. The TR is not site-specific, as such, 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 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.

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. 5, 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. 4, 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 (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 cooling using ((

)).

3 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 spent fuel pool (SFP) cooling for ((

)).

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 maintain core cooling, containment, and spent fuel pool cooling using ((

)) 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 ((

)). This statement is consistent with the docketed response for audit issue, MBDBE.LTR-8 (Ref. 10), which clarified that ((

)) in the design bases for plants adopting this TR.

While NuScales statement is not unacceptable, 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. As written in Limitation and Condition no. 5.1, the staff determined that the same conditions, are applicable. Furthermore, in Limitation and Condition no. 5.1 the staff clarifies what is required in order to meet Condition of Use #1 from the TR. Condition of Use #2 states that an adopter of the TR must provide a plant specific thermal analysis demonstrating its capability to ((

)).

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 described in the TR as based on a NuScale Power Modules (NPM) ability to maintain core cooling and containment for ((

)). The SFP is part of the UHS.

4 The TR is to be generically applicable to NuScale reactor designs with the capabilities and features discussed in the TR. A supporting analysis to demonstrate these capabilities is to be provided by an applicant or a licensee who adopts the TR.

The potential for various design features as well as the need for analysis to be provided by the future COL applicant 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 section 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. 4),

and RG 1.226, Revision 0 (Ref. 5), 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.

Therefore, the staff finds that the initial assumptions described in TR following a beyond-design-basis external event (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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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, ((

)).

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 the MCR. The TR does not contain any supporting analyses on EDAS, and the EDAS design has not been shown to be acceptable at this time, therefore, 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. Condition of Use #1 requires an adopter of the TR to provide a plant

5 specific design that ((

)) described within the TR. Limitation and Condition 5.1 expands on this requirements by specifying that the adopter must also ensure the plant design includes the described system functions and system design features and equipment qualification as described in the TR. Condition of Use #2 states that an adopter of the TR must provide a plant specific thermal analysis demonstrating ((

)) identified in Table 4-1 of the TR, and the initial conditions described in Section 3.1 of the TR. 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). By satisfying this limitation and condition the augmented quality, capacity, and capability of the EDAS to provide power for a BDBE will be demonstrated.

In section 3.2.4 of the TR, NuScale states ((

)). The staff notes that UHS and SFP level instruments can be powered by EDAS, ((

)), are not discussed and therefore, an applicant must satisfy limitation and condition no. 5.2. By satisfying this condition, an applicant would be able to address the capacity and capability for the portable generators. 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. Therefore, an applicant or licensee following this TR and satisfying Limitation & 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The staffs review of the NuScale electrical power system design to supply power 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, finds that an applicant or licensee following this TR would have to demonstrate that its use of the NuScale design meets the requirements in 10 CFR 50.155(c) for the capacity and capability of the power supplies and 10 CFR 50.155(e) power supplies for spent fuel monitoring, subject to limitations and conditions no. Error!

Reference source not found.1.

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:

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. ((

6

))

((

))

Core coolingduring a loss of all AC power event, containment isolation within the ((

)) of the event preserves the reactor coolant inventory. The decay heat removal system (DHRS) passively removes decay heat for up to the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a loss of all AC power event. By 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the ECCS valves automatically open and the ECCS ((

)).

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 to be maintained ((

)).

SFP CoolingAs stated in TR section 3.2, the SFP, as part of the UHS, is interconnected 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to provide additional assurance that systems respond as designed.

Based on its review of the TR, the NRC staff finds that, subject to satisfaction of limitation and condition no. 5.1, following a loss of all AC power event, an applicant or licensee following this TR would have to demonstrate that its use of the NuScale plant design capabilities meet 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.

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.

7 (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 are assumed to survive the BDBEE, including wind and flood events. Table 4-1 in the TR provides a ((

)).

The staff reviewed the systems and responses to the BDBEE. As described in the TR, NuScale states that the plant systems are protected from BDBEE and should continue to perform to support their safety functions. As stated in limitation and condition no. 5.1, an applicant will be expected to provide plant specific design and analysis to support NuScales assertion and demonstrate that the design supports the expected coping period.

3.3 Safety Functions during a Loss of All Alternating Current Power 3.3.1 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 pf being implemented site-wide and must included 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.

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

8 insertion. ((

))

((

)). Consistent with SECY-19-0066, and limitations and condition 5.1, any applicant or licensee referencing this TR is required to provide plant specific design and analysis 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 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 water can be added to the UHS which also adds water to the SFP.

The staff reviewed the safety functions described in TR section 5.0. TR Table 5-3, Baseline Coping Capability Summary, lists a summary of the core cooling, containment, and SFP cooling functions. The systems listed and their safety functions provide confidence that if a future applicant provides an acceptable supporting analysis, the NPM would be able to perform extended coping. As stated in limitation and condition 5.1 an applicant or licensee using a NuScale design would be required to provide plant specific design and analysis to support the expected coping period. This includes all modes of operation.

3.3.2 Indefinite maintenance of core cooling, containment, and spent fuel pool cooling capabilities 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. 6) state, in part (at 39709):

9 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 12-06, Revision 4, states, in part:

Since site access is considered to be restored to near-normal within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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

10 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the event initiation and that off-site resources should be able to be mobilized by 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such that FLEX strategies and/or resources are not required to be explicitly planned for the period beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. 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 off-site 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 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 TR describes the specified timeframe as ((

)). Therefore, the TR submittal acknowledges that operator actions will be required during the initial coping phase to support replenishment of the UHS at the start of the final, long-term coping phase. ((

)). 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

11 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 limitations and conditions no. Error! Reference source not found.5.2 and Error! Reference source not found. to resolve this issue. Limitation/condition 5.2 requires that an applicant to provide preplanned mitigating actions to ensure that long term coping requirements regarding UHS make up are satisfied, or to 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. 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 5.4 to resolve this issue.

Limitation/condition 5.4 requires that preplanned mitigating actions are established to ensure the sustainability of {request that the preceding text required to perform the mitigating strategy tasks assigned to the remain and not be deleted - please see comment} control room operators post -BDBEE, as described in the LTR, in compliance with 10 CFR 50.155(b)(1)(i) and 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:

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. 5.5 to resolve this issue.

Limitation/condition 5.5 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:

((

12

))

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. Error! Reference source not found. to resolve this issue. Limitation 5.3 requires that preplanned mitigating actions are established to ensure that ((

)) 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 a large plant area due to explosions or fire (LOLA), as required by 10 CFR 50.155(b)(2) 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

13 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 an applicant or licensee that references this TR must satisfy limitation and condition no. 5.6, to demonstrate that the fire protection procedures and equipment design features as stated above and in 10 CFR 50.48 and GDC 3 would provide adequate LOLA coping capability without the need for additional mitigation strategies.

Compliance with limitation and condition 5.6 would provide reasonable assurance of adequate LOLA coping capability.

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 shielding and heat removal or are protected by a siphon break which prevents inadvertent lowering of the pool level below safety limits.

Because the SFP is below grade and designed such that the SFP cannot be drained below safety limits, the NRC staff has determined, subject to satisfaction of limitation and condition no.

5.1 by an applicant or licensee that references this TR, including a demonstration that enhanced mitigation capability is not required for the SFP, that the enhanced SFP mitigation capability such as a diverse or portable makeup capability for mitigating a LOLA event is not required for the SFP.

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 (CNTS) 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

14 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 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 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 reviewed NuScales system description as stated above and determined that the passive plant design would provide reasonable assurance, subject to satisfaction of limitation and condition no. 5.1 by an applicant or licensee that references this TR, that the key safety functions to mitigate potential fuel damage and radiological release are implemented.

3.4.4 Conclusion The NRC staff concludes that an applicant or licensee using this TR for a NuScale design would be required to provide adequate LOLA coping capability ((

)) per NEI 06-12 guidance, upon implementing the conditions and limitations detailed above 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.

15 (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:

((

)).

The proposed strategy for NPM designs could provide for extended coping ((

)). 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 ((

)) [and] (2) provide a plant specific thermal analysis demonstrating

((

)). Therefore, the NRC staff concludes, subject to satisfaction of limitation and condition no. 5.1,that an applicant or licensee following this TR for a NuScale design would 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:

((

16

))

As stated in Conditions of Use no. 2 of TR section 1.3, which is endorsed by 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 installed plant equipment includes required activities related to replacement of the SFP level monitoring instrumentation power supply as described in Section 9.0 of the TR. Therefore, subject to limitations and conditions nos. 5.7, the staff finds that an applicant or licensee referencing this TR for a NuScale design would satisfy the requirements of 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:

(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 5years 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. 8) 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 ((

))

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:

((

))

17 As discussed above in section 3.6, the staff finds it acceptable to credit the training program for site staff that covers ((

))

((

)), as stated in 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 provided capabilities to allow use of procured offsite equipment, would meet 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 & Condition No. 5.7 to demonstrate compliance with the requirements of 10 CFR 50.155(e).

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 a plants 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.

18 5.0 LIMITATIONS AND CONDITIONS The staffs approval is limited to the application of this methodology to the NuScale reactor design 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 the ((

)) 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 and (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 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 batteries/install portable generators, as described in Section 9.0 of the TR, and that a preplanned source is established to provide the replacement batteries or portable generators 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.

5.4 An applicant or licensee referencing the TR must address (e.g., in plant procedures) how control room operators will be sustained for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the control room and subsequently exit the control room at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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.

5.5 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.

5.6 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 as described in Section 9.0 of the TR.

19

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)).

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. NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. 4 (ML16354B421)

5. Regulatory Guide (RG) 1.226, Flexible Mitigation Strategies for Beyond-Design-Basis Events, Rev. 0 (ML19058A012)

6. Federal Register Volume 84, No. 154, Mitigation of Beyond-Design-Basis Events, August 9, 2019, pages 39684; 39709

7. RG 1.227, Wide-Range Spent Fuel Pool Level Instrumentation, issued June 2019 (ML19058A013)

8. 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

9. NRC, SECY-19-0066, Staff Review of NuScale Powers Mitigation Strategy for Beyond-Design-Basis External Events, June 26, 2019 (ML19148A443)

10. NuScale Power, LLC, Response to NuScale Topical Report Audit Question, A-MBDBE.LTR-8, (ML24178A401 (public) and ML24178A403 (non-public)).