ML25051A242

Safety Evaluation for NuScale Xpc TR (Public Version)

ML25051A242
Person / Time
Issue date:	02/25/2025
From:	David Drucker NRC/NRR/DNRL/NRLB
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Download: ML25051A242 (48)

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THIS NRC STAFF DRAFT SE HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ACRS. THIS DRAFT SE HAS NOT BEEN SUBJECT TO FULL NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITION SAFETY EVALUATION REPORT BY THE U.S. NUCLEAR REGULATORY COMMISSION TOPICAL REPORT TR-124587, REVISION 0, JANUARY 5, 2023 EXTENDED PASSIVE COOLING AND REACTIVITY CONTROL METHODOLOGY NUSCALE POWER, LLC This document contains proprietary information pursuant to Title 10 of the Code of Federal Regulations (10 CFR) section 2.390, Public inspections, exemptions, requests for withholding.

Proprietary information is identified by text enclosed within bolded double brackets, as shown here: ((example proprietary text.

ii TABLE OF CONTENTS 1.0 INTRODUCTION AND BACKGROUND............................................................................ 1 2.0 REGULATORY BASIS FOR XPC EVALUATION MODEL REVIEW................................. 2 2.1 Regulatory Requirements............................................................................................... 2 2.2 Regulatory Guide 1.203.................................................................................................. 5 2.3 NUREG-0800 Standard Review Plan............................................................................. 7 3.0 NUSCALE XPC EVALUATION METHODOLOGY

SUMMARY

......................................... 7

4.0 TECHNICAL EVALUATION

............................................................................................... 8 4.1 Introduction and Scope................................................................................................... 8 4.2 Background and Acceptance Criteria............................................................................. 9 4.3 EMDAP Process............................................................................................................. 9 4.4 Phenomena Identification and Ranking.......................................................................... 9 4.5 Evaluation Model Overview and Computational Tools................................................. 13 4.6 NRELAP5 Computer Code and Assessment Basis..................................................... 15 4.7 Extended Passive Cooling Thermal Hydraulic Analysis Methodology Evaluation........ 22 4.8 Evaluation for Reactivity Control and Boron Distribution.............................................. 31 5.0 LIMITATIONS AND CONDITIONS.................................................................................. 43

6.0 CONCLUSION

S............................................................................................................... 45 7.0 REFERENCES................................................................................................................. 46

1 1.0 INTRODUCTION AND BACKGROUND By \ letter dated January 5, 2023, NuScale Power, LLC (NuScale, or the applicant) submitted Topical Report (TR) Extended Passive Cooling and Reactivity Control Methodology, TR-124587, Revision 0, (Agency Wide Documents Access and Management System (ADAMS) Accession No. ML23005A308), to the U. S. Nuclear Regulatory Commission (NRC) staff for review. By \ letter dated July 31, 2023 (ML23205A004), the NRC informed NuScale of its acceptance of TR-124587-P, Revision 0, for a detailed technical review. The conclusions in the staffs SER are based on markups to Revision 0 of the TR provided by NuScale in advance of Revision 1 of the TR being docketed. The XPC TR, presents the NuScale methodology used to evaluate (1) the emergency core cooling system (ECCS) and decay heat removal system (DHRS) extended passive cooling (XPC) of the NuScale Power Module (NPM-20) after a successful initial short-term response to a design basis event; (2) reactivity control during XPC of the NPM-20; and (3) margin to the boron solubility limit for precipitation to demonstrate coolable geometry is maintained in the NPM-20. Certain key information from the US460 plant design (Reference 4) has been considered for the safety evaluation since the US460 cooling pool characteristics are important to the response of the NPM-20 module for the extended cooling period after design basis events. NuScale stated that the TR methodology is an extension of the NuScale loss of coolant accident (LOCA) evaluation model (EM) (Reference 1) and NuScale Non-LOCA EM (Reference 2) and the methodology uses a graded approach to the evaluation model development and assessment process (EMDAP) defined in Transient and Accident Analysis Methods, Regulatory Guide (RG) 1.203. Therefore, any future changes to the LOCA or Non-LOCA EMs need to be assessed by the applicant for their potential impact on the XPC EM. Any subsequent changes to the XPC methodology will require NRC approval. This is listed as Limitation and Condition (L/C) 1 in Section 5, Limitations and Conditions, of this safety evaluation report (SER). This SER documents the results of the NRC staffs in-depth technical evaluation of the XPC TR, and the methodology used to evaluate extended passive cooling and reactivity control in the NuScale NPM-20. The NRC staff performed the review to determine the technical adequacy of the thermal hydraulic methods and modeling techniques, shutdown margin calculations and boron distribution characterization as described in the XPC TR for evaluating extended passive cooling and reactivity control. The applicant developed the NuScale Extended Passive Cooling and Reactivity Control Evaluation Methodology to evaluate extended passive cooling (XPC) NuScale Power Module (NPM-20) response during emergency core cooling system (ECCS) operation and decay heat removal system (DHRS) operation. The US460/NPM-20 design is described in the NuScale standard design approval application (SDAA) (Reference 4). Some of the design features for the NPM-20 are described in the EM topical report 3.2 subsections. This safety evaluation is based on an assessment of the full NPM-20 design and response that implements the methodology.

2 The basic functions of the systems used for extended passive cooling and reactivity control are to:

1. Cool down the reactor coolant system (RCS) after any LOCA or non-LOCA transient.

2. Prevent core uncovery or heatup after the transient from the initiating transient to XPC.

3. Prevent precipitation of the boric acid in the reactor coolant system.

4. Maintain sufficient boron in the RCS to prevent re-criticality with the highest worth control rod stuck out of the reactor core.

The systems used to fulfill extended passive cooling and reactivity control functions must function for a period of at least 72 hours until transition to the recovery phase. The XPC topical report covers the 72-hour period after event initiation. The XPC topical report does not cover timeframes after 72 hours or post-event recovery actions. Therefore, Limitation and Condition 5 has been developed to restrict the use of the methodology to design basis event progression up to 72 hours, and post-event recovery actions must be addressed outside the topical report. This SER reviews the acceptability of the XPC methodology required for analysis of the full spectrum of LOCA breaks and non-LOCA events to assure the four XPC functions discussed above can be demonstrated. The NRC staff reviewed the methodology and modeling for the spectrum of LOCA break sizes and locations and non-LOCA transients for the NPM-20 design. 2.0 REGULATORY BASIS FOR XPC EVALUATION MODEL REVIEW NRC staff has reviewed the XPC and reactivity control analysis methodology described in the XPC TR to determine whether this methodology is acceptable for performing NPM-20 long term cooling and reactivity response calculations and meets the applicable regulatory requirements or meets regulatory requirements in part. This section of the SER describes the regulatory basis and supporting guidance documents that the NRC staff uses to determine whether the methodology described in the XPC EM is acceptable. 2.1 Regulatory Requirements The relevant LOCA requirements for this area of review and the associated acceptance criteria, given in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, (SRP) Section 15.6.5 Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, Revision 3, (ADAMS Accession No. ML070550016), include the primary acceptance criteria set in general design criteria (GDC) 35, Emergency Core Cooling, in Appendix A, General Design Criteria for Nuclear Power Plants, of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, (b)(4) and (b)(5). These requirements are also relevant to non-LOCA events that utilize ECCS for long-term analyses. Other relevant requirements for this area of review include: GDC 26, Reactivity Control System Redundancy and Capability, for anticipated operational occurrences (AOOs), and GDC 27 for postulated accidents specify that a stuck control rod is to

3 be assumed for both AOOs and postulated accidents. GDC 34 is also relevant because it specifies residual heat removal requirements for AOOs. 10 CFR 50.46 and 10 CFR Part 50, Appendix K, ECCS Evaluation Models, present the acceptance criteria for ECCS for light-water nuclear power reactors and the required and acceptable features of the EMs employed to demonstrate compliance with these regulations. The NuScale XPC EM is intended to conform to the required and acceptable features of Appendix K to 10 CFR 50. The XPC EM is an extension of the NuScale LOCA EM (Reference 1), the disposition of the 10 CFR 50 Appendix K requirements that apply to the long-term cooling phase are applied in the same manner as for the LOCA EM including identification of exemptions. 2.1.1 10 CFR 50.46 ECCS and Appendix K to 10 CFR Part 50 Requirements The regulations in 10 CFR 50.46 (a)(1)(i) require that each light-water reactor, fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding, must be equipped with an ECCS, the performance of which is evaluated for the most severe postulated accident. 10 CFR 50.46 The regulations in 10 CFR 50.46(a)(1)(i) require that ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and provides for two alternative options, as mentioned above, for acceptable EM analytical techniques and methods: realistic with consideration of uncertainties or conservative in accordance with Appendix K to 10 CFR

50.

As noted above, the NuScale XPC EM is intended to conform to the second category of EMs, which 10 CFR 50.46(a)(1)(ii) indicates are to be developed in conformance with the required and acceptable features of Appendix K, to 10 CFR Part 50. The XPC EM is an extension of the NuScale LOCA EM (Reference 1), the disposition of the 10 CFR 50 Appendix K requirements that apply to the long-term cooling phase are applied in the same manner as for the LOCA EM including identification of exemptions. Furthermore, 10 CFR 50.46(c)(2) defines an EM as the calculational framework for evaluating the behavior of the reactor system during a postulated LOCA. An EM includes one or more computer programs and all other information necessary for applying the calculational framework to a specific LOCA (the mathematical models used, the assumptions included in the programs, the procedure for treating the program input and output information, the parts of the analysis not included in the computer programs, values of parameters, and all other information necessary to specify the calculational procedure). ECCS Performance Criteria The regulations at 10 CFR 50.46(a)(1)(i) require that ECCS calculated cooling performance following postulated LOCAs conform to the criteria set forth in 10 CFR 50.46(b). This regulation defines the criteria for calculated ECCS cooling performance during postulated LOCAs in 10 CFR 50.46(b)(1) through 10 CFR 50.46(b)(5). The applicable portions for the XPC TR are as follows:

4 (b)(4) Coolable Geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling. (b)(5) Long-Term Cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. 2.1.2 10 CFR 50, Appendix A GDC establish minimum requirements for the principal design criteria (PDC) for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The NuScale SDAA (Reference 4) for the US460 (which incorporates the NPM-20 module) includes PDC that were developed based on the GDC. The NRC staff notes that the applicant provided PDC for GDC 34, Residual Heat Removal, and 35. PDC 34 and 35 proposed by the applicant are functionally identical to GDC 34 and 35 for this evaluation model. The NuScale-proposed PDC 35, based on GDC 35, establishes the required safety function of the ECCS, as described in US460/NPM-20 SDAA (Reference 4): A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling would be prevented, and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure. The NuScale PDC 34, based on GDC 34, establishes the required safety function of the DHRS, as described in the US460/NPM-20 SDAA (Reference 4). PDC 34 states: A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure.

5 GDC 26, is relevant as it relates to the control of reactivity changes so that SAFDLs are not exceeded during AOOs. This control is accomplished by provisions for appropriate margin for malfunctions (e.g., stuck rods). GDC 27, Combined Reactivity Control Systems Capability, relates to controlling the rate of reactivity changes to ensure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. 2.2 Regulatory Guide 1.203 RG 1.203, Transient and Accident Analysis Methods, provides guidance for developing and evaluating EMs for accident and transient analyses. Section D, Implementation, states that the guide is approved for use as an acceptable means of complying with the NRC regulations and for evaluating submittals of new or modified EMs proposed by vendors or operating reactor licensees that, in accordance with 10 CFR 50.59 [Changes, tests and experiments], require NRC staff review and approval. The XPC EM has been developed as a deterministic analysis approach intended to meet the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K and relevant GDCs. The XPC TR states that the approach to the development of the model follows that outlined in RG 1.203. Within RG 1.203, the phenomena identification and ranking table (PIRT), is identified as a key requirement for evaluation model development. Section 3 of the XPC EM TR documents the PIRT that NuScale developed for the NPM-20. Section 4.4 of this SER provides the NRC staffs review of this PIRT. 2.2.1 Evaluation Model Concept In accordance with 10 CFR 50.46(c)(2), RG 1.203 states that the EM constitutes the calculational framework for evaluating the behavior of the reactor system during a postulated transient or a design-basis accident. As such, the EM may include one or more computer programs, special models, and all other information needed to apply the calculational framework to a specific event, such as procedures for treating the input and output information, specification of those portions of the analysis not included in the computer programs for which alternative approaches are used, and all other information needed to specify the calculational procedure. It is the entirety of an EM that ultimately determines whether the results are in compliance with applicable regulations, and therefore the development, assessment, and review processes must consider the entire EM. Most EMs used to analyze the events in SRP Chapter 15, Transient and Accident Analysis, rely on a system-level code that describes the transport of fluid mass, momentum, and energy throughout the RCS. The XPC EM uses the NuScale NRELAP5 Version 1.7 systems analysis computer code, which was developed from the Idaho National Laboratory (INL) RELAP5-3D computer code, as well as MATLAB, hand calculations, and CASMO/SIMULATE. 2.2.2 Evaluation Model Development and Assessment Principles RG 1.203 defines the following six basic principles as important to follow in the Evaluation Model Development and Assessment Process (EMDAP).

6 (1) Determine requirements for the EM. (2) Develop an assessment base consistent with the determined requirements. (3) Develop the EM. (4) Assess the adequacy of the EM. (5) Follow an appropriate quality assurance (QA) protocol during the EMDAP. (6) Provide comprehensive, accurate, up-to-date documentation. RG 1.203 discusses the NRC staffs regulatory position and provides guidance concerning methods for calculating transient and accident behavior. Part C of RG 1.203 provides Regulatory Positions on aspects of an EMDAP that address the basic principles identified above and offer additional guidance. Regulatory Position 1, Evaluation Model Development and Assessment Process (EMDAP) RG 1.203 identifies four basic elements developed to describe an EMDAP. The elements address directly the first four EMDAP basic principles and provide guidance in 20 individual steps. In addition, Regulatory Position 1 includes requirements for reaching an adequacy decision. The basic elements of Regulatory Position 1 are identified below. Element 1: Establish Requirements for EM Capability Element 2: Develop Assessment Base Element 3: Develop EM Element 4: Assess EM Adequacy Decision Regulatory Position 2, Quality Assurance RG 1.203 discusses QA during development, assessment, and application of an EM and the requirements of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50. Regulatory Position 3, Documentation RG 1.203 provides guidance on the requirements to document the development of LOCA EMs. Regulatory Position 4, General Purpose Computer Programs RG 1.203 provides guidance on development of general purpose transient analysis computer programs designed to analyze a number of different events for a wide variety of plants. Specifically, Regulatory Position 4 states that application of the EMDAP should be considered as a prerequisite before submitting a general purpose transient analysis computer program for review as the basis for EMs that may be used for a variety of plant and accident types. Regulatory Position 5, Graded Approach to Applying the EMDAP Process RG 1.203 provides guidance on the extent to which the full EMDAP should be applied for a specific application based on the following four EM attributes: (1) Novelty of the revised EM compared to currently accepted models. (2) Complexity of the event being analyzed. (3) Degree of conservatism in the EM.

7 (4) Extent of any plant design or operational changes that would require reanalysis. Appendix A of RG 1.203, Additional Considerations in the Use of this Regulatory Guide for ECCS Analysis, describes uncertainty determination and provides guidance for best-estimate LOCA analyses. Appendix A of RG 1.203 refers to NUREG-0800, Standard Review Plan, Sections 15.6.5 and 15.0.2, Review of Transient and Accident Analysis Method. 2.3 NUREG-0800 Standard Review Plan SRP Section 15.0.2 is the companion SRP section for RG 1.203. SRP Section 15.6.5 describes the review scope, acceptance criteria, review procedures, and findings relevant to ECCS analyses. 3.0 NUSCALE XPC EVALUATION METHODOLOGY

SUMMARY

The NPM-20 has several unique features that required NRC staff to perform detailed reviews of the NuScale XPC EM to determine whether this methodology is adequate. The NuScale design is a small modular pressurized water reactor that relies on natural circulation during normal plant operation and that uses a unique high-pressure containment as an integral part of the ECCS to keep the reactor core covered with a collapsed liquid level above the top of the active core through all potential design-basis LOCA events and non-LOCA events where ECCS is eventually actuated. The NPM-20 has an ECCS supplemental boron design feature that provides additional soluble boron for recirculation into the Reactor Pressure Vessel (RPV) during ECCS operation to maintain subcriticality. During a NuScale NPM-20 LOCA or ECCS actuation, the high-pressure water and steam leaving the RPV is contained in the (containment vessel) CNV. The CNV is designed to enable the ECCS system to return cooled RCS liquid to the downcomer to prevent core uncovery during design basis LOCAs or for non-LOCA events where ECCS is actuated. During a LOCA, the ECCS valves, two reactor vent valves (RVV) and two reactor recirculation valves (RRV), receive a signal to open. However, the RRVs are blocked from opening by the inadvertent actuation block (IAB) Valve until the pressure differential between the RPV and CNV drops below the IAB threshold. When the RVVs open on an ECCS signal, steam generated inside the RPV from decay heat and stored energy exits the RPV through the RVVs and condenses on the inside of the CNV wall. Once the RRVs open, inventory is returned from the CNV to the RPV through the RRVs. During a NuScale NPM-20 LOCA or ECCS actuation, the ECCS supplemental boron system adds boron to the CNV liquid inventory by collecting condensate from the steam released from the RPV and using it to dissolve boron pellets in baskets mounted to the containment wall (ML24346A366). In addition, mixing tubes that collect condensate, not used for boron pellet dilution, direct the condensate to the bottom of containment to promote mixing of the inventory in the containment. Because of the unique features of the NuScale NPM-20 CNV design and the NuScale ECCS system, the NRC staff review of the NuScale XPC EM topical focused particular attention on the ability of the NuScale LOCA EM to address the following design issues and phenomena:

8 The capability to predict the collapsed liquid level in the RPV so that the NuScale power module maintains the collapsed liquid level in the RPV above the reactor core during design basis LOCA and non-LOCA events with ECCS actuation. The core remains subcritical for 72 hours assuming the worst rod stuck out, including the impacts of boron precipitation and volatility. In addition, the NRC staff review of the XPC EM topical focused particular attention on the capability of the NuScale NRELAP5 computer code to accurately model the tests performed at the NPM scaled model NuScale Integral System Test (NIST-2) facility and on confirming that the geometric dimensions and operating conditions of NIST-2 adequately represent the NPM-20 design. Because the NuScale design relies on maintaining a collapsed liquid level above the top of the reactor core and remaining subcritical, the NRC staff evaluation of the NuScale XPC Evaluation Methodology is limited to consideration of the conservative assumptions and modelling assumptions to determine that this design objective is adequately modeled. The determination to support the SDAA that the collapsed liquid level remains above the top of the core and is subcritical will be determined as part of that separate review of the application of this methodology. The NRELAP5 computer code, Version 1.6 (ML23011A012), was submitted as the systems analysis computer code for the NuScale XPC Evaluation Methodology. NuScales primary changes to the INL RELAP5-3D version included implementation of a new helical coil SG (HCSG) component, and the addition of new containment condensation models to describe the unique design features of the ECCS operation of the NPM-20. Subsequently during the US460 review, NuScale submitted NRELAP5 Version 1.7 (ML24228A242) as the systems analysis computer code for the NuScale XPC Evaluation Methodology, replacing NRELAP5 Version 1.6.

4.0 TECHNICAL EVALUATION

This section of the SER, summarizes and evaluates the information in the sections of the TR against the regulatory requirements for that section. The L/Cs on the approval of the XPC LTR are discussed below and then summarized in Section 5 of this SER. The conclusions from the review are discussed below and then summarized in Section 6 of this SER. In addition, the NRC staff performed audits of information provided by the applicant in support of the NRC staffs review of the TR that are referred to throughout this evaluation. In all instances where audits are referred to, unless noted otherwise, the audit plan and audit report of those XPC EM audits are available or referenced in the audit report (ML24263A009). 4.1 Introduction and Scope Section 1.1 of the XPC LTR, states that the purpose of the NuScale EM is to evaluate the XPC NuScale NPM response during ECCS operation and DHRS operation. Section 1.1 also states: The topical report describes: ECCS long-term cooling (LTC) and extended DHRS passive cooling analysis scope methodology acceptance criteria

9 methodology for demonstrating that the acceptance criteria are met for the NPM The XPC EM addresses: evaluation of XPC for decay and residual heat removal evaluation of boron transport phenomena evaluation of criticality during extended cooling with DHRS and ECCS evaluation of boron precipitation The NRC review and safety evaluation are focused on the purpose described in the LTR, with respect to the US460 and NPM-20, and to determine whether the proposed methodology meets the applicable regulations. NuScale states that their XPC EM is an extension of the NuScale LOCA EM (Reference 1) and NuScale non-LOCA EM (Reference 2) and is developed using a graded approach to the EMDAP defined in RG 1.203. The XPC LTR provides a description of the methodology used by NuScale for XPC and long-term reactivity control analyses, and this methodology is reviewed in this SER for compliance with applicable regulatory criteria. However, the XPC LTR does not provide any final licensing analyses, and this review of the XPC LTR does not evaluate the acceptability of the NuScale NPM-20 or provide any conclusions on the acceptability of the NuScale NPM-20 design. 4.2 Background and Acceptance Criteria Section 2 of the XPC LTR provides a description of how the NuScale XPC EM conforms to the EMDAP of RG 1.203. XPC LTR Table 2-1 states that LTR Section 8.0 describes NuScale procedures implementing the QA program that governs requirements for documentation and verification of the EM. Section 4.5.3 of this SER discusses the QA requirements and evaluation. 4.3 EMDAP Process Section 2.4 of the XPC LTR provides a brief description of the NPM and a brief summary of NPM operation. Section 2.4 provides the XPC EM roadmap and an overview of the EMDAP process. Section 3.2 of the XPC LTR provides a general description of the NPM-20 design and key design features. The XPC LTR is an extension of the LOCA and non-LOCA LTRs, which also follow the EMDAP process. 4.4 Phenomena Identification and Ranking As discussed in Section 3.1 of the XPC LTR, NuScale developed the NPM-20 PIRT for XPC conditions involving extended DHRS and ECCS operation as an extension of the PIRT for the previously approved US600 design which was described in the design certification application (DCA) LTC TR (Reference 3). Additionally, the PIRT for the XPC EM was modified to account for the design changes for the US460 design, which contains multiple NPM-20 modules. The NRC staff focused the review of the XPC PIRT on the changes and modifications to the DCA LTC PIRT that reflect design changes for the US460 and NPM-20.

10 NuScale documented phenomena and processes of high and medium ranked importance in Table 3-4 in Section 3.4 of the XPC LTR and did not document low ranked phenomena or processes. Typically, the low ranked phenomena and processes are included in EM LTRs. The XPC EM PIRT is based on the previous DCA US600 PIRT and includes phenomena that are familiar to the NRC staff based on the entirety of the DCA US600 review. The new design considerations for the US460 NPM-20 PIRT have been reviewed by the NRC staff and are the focus of the current review. Therefore, having the low ranked PIRT included in the XPC LTR is not necessary in this specific review of the XPC EM. NuScale has defined the acceptance criteria evaluated by the PIRT as subcriticality, coolable geometry, and coolant collapsed liquid level. The PIRT process for the XPC LTR evaluates these acceptance criteria for the extended ECCS and extended DHRS phases after design basis events. NuScale defines design-basis events as having three phases. Phase 1 is addressed separately by NuScale in the applicable LOCA and non-LOCA LTRs, while the two phases considered in the XPC PIRT are as follows: ECCS Phase 2: Long-term recirculation with liquid flow from containment into the RPV through the recirculation valves, and vapor flow from the RPV into containment through the vent valves (ML24346A343) DHRS Phase 3: Extended stable natural circulation. Primary system power and flow rates reflect decay power levels. DHRS is actuated and secondary side flow rates and pressures decrease along with primary side pressure and temperature. During extended DHRS operation, depending on the integral DHRS heat removal and RPV fluid mass, the mixture level can decrease from the pressurizer (PZR) to the riser outlet region, and below the top of the riser. NuScale described the process used to evaluate and update the DCA US600 LTC PIRT to the US460 NPM-20 PIRT. The process described by NuScale to develop the US460 XPC PIRT from the US600 PIRT is summarized below:

1. Identify new components that need to be addressed due to design changes,

2. Consolidate component and phenomena descriptions.

3. Identify the appropriate importance ranking and knowledge level for the consolidated phenomena, as applied to the consolidated components. The general approach is to retain the highest importance level and lowest knowledge level.

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}} The NRC staff reviewed the information provided as a result of implementing the process but did not provide an evaluation of the process itself. Additionally, the NRC staff notes that the outcome of the process and PIRT are dependent on the US460 specific design and design changes from the US600 design as well as the final application of the methodology to the specific design and associated design features. Therefore, when applying the methodology, the entirety of the specific design application must be taken into account. When performing the review of the PIRT, the NRC staff reviewed the entirety of the US460 NPM-20 design and response.

11 NRC staff evaluated the XPC phases selected for the PIRT discussions. The NRC staff determined that the NuScale XPC phases selected as the basis for their PIRT process are acceptable for establishing the ranking phenomena that must be considered in the XPC evaluation model with consideration for how the upstream design basis event response can impact the long-term phase figures of merit. 4.4.1 NPM Design Features and EM Applicability NuScale describes the design features of an NPM in Section 3.2 of the XPC LTR. The NRC staff confirmed that the design description and key design features documented in Section 3.2 and respective subsections are consistent with the US460 NPM-20 design features documented in the SDAA (Reference 4). The NRC staff notes that the design description and features documented in these sections dont align with the design of the DCA US600. NuScale summarizes key features of the plant design or plant design requirements that must be met in order to apply the XPC EM in Table 3-3 of the XPC LTR. The NRC staff confirmed that the design description and design features are the minimum needed attributes to apply the XPC EM methodology. The NRC staff finds that these design features in combination with the application of the methodology is necessary for the methodology to be valid because the NRC staff notes that the documented information in Section 3.2 is not a complete description of the design or description of interfacing design attributes provided in totality and must be taken in combination with the design in which it is applied. The NRC staff notes that the integrated design, US460 and NPM-20, in which the methodology is applied is integral to the NRC staffs approval of the applicability of the methodology. The design features described in Section 3.2 are necessary, along with the integral design response of the US460 and NPM-20, which forms the basis for the NRC staffs approval of the XPC LTR, because the PIRT was constructed and updated based on the design of the US460 and NPM-

20. Applicability evaluation(s) would be needed in order to apply the method to a design that is not the US460 and NPM-20 (Reference 4). L/C 2 has been developed to ensure that the applicability is only for the US460/NPM-20 unless an applicability assessment is made and submitted to the NRC for review and approval.

4.4.2 Figures of Merit In Section 3.3 of the XPC LTR, NuScale discussed the Figures of Merit (FOM) selected for their XPC EM, which are subcriticality, coolable geometry (boron concentration below solubility limit for precipitation) and collapsed liquid level above the top of active fuel. NRC staff finds that the NuScale XPC EM FOMs, namely that (1) the core remains subcritical (2) boron concentration remains below the solubility limit for precipitation and (3) the collapsed liquid level in the RPV remains above the core at all times during all scenarios, show conservatism with respect to applicable acceptance criteria and are acceptable FOMs for the NuScale XPC EM provided that all acceptance criteria for the short term LOCA (Reference 2) and non-LOCA (Reference 3) LTRs acceptance criteria are met.

12 4.4.3 PIRT Rankings Section 3.4 of the XPC LTR discusses the results of the NuScale XPC EM PIRT process and provides a list of High and Medium Ranked Phenomena and the Phenomena Identification and Ranking Summary Table. The NuScale PIRT identified phenomena and processes that could occur during a long-term event phase, ranked the relative importance of each, and assessed the knowledge level for each. The relative importance was ranked as High, Medium, Low, or inactive, as identified in Section 3.1 of the XPC LTR. Knowledge level was divided into fully known, known, partially known, or very limited knowledge. Finally, the component of the NPM for which the phenomenon or process was ranked was identified. In Table 3-4 List of Extended Passive Cooling High and Medium Importance Ranked Phenomena and Components, of the XPC LTR, NuScale provides the listing of the findings of their final PIRT for phenomena ranked of high and medium importance. The table includes the importance ranking and knowledge ranking for each phenomenon, and the applicable components. NuScale only addressed the high and medium ranked phenomena and processes in their XPC LTR. In order to assess low ranked phenomena, the NRC staff reviewed the NuScale DCA US600 PIRT report. The NRC staff review did not identify any low to medium ranked phenomena or processes that should have been ranked medium or high. Therefore, NRC staff finds that the NuScale XPC LTR table for high and medium ranked phenomena is acceptable. The NRC staff finds that addressing the high and medium PIRTs is acceptable because the NRC staff did not identify any PIRTs that were not included based on an independent PIRT evaluation of the US460 and NPM-20 phenomena. The NRC staffs review of the applicants XPC PIRT was informed by an independent and related PIRT performed by the NRC staff. There were some differences in the formulation of the NRC staff and applicants PIRTs, but both PIRTs focused on boron dilution during extended DHRS and ECCS operation. The NRC staff compared the phenomena and phenomenon importance rankings related to extended DHRS and ECCS cooling from the NRC staff and applicants PIRTs. Therefore, the NRC staff finds that the PIRT selections and rankings are adequate as a basis for the NuScale XPC EM in combination with the rest of the method and code assessments provided. NRC staff also reviewed Sections 4.4.3 and 4.4.4 of the XPC LTR, which document a summary of the NuScale assessment of the high and medium phenomena from the PIRT identified in Section 3.4 of the XPC LTR. The PIRT assessment that NuScale presented describes the phenomena, the importance of the phenomena in parts of the EM calculation, and how the phenomena are addressed in the XPC evaluation method. The NRC staff notes that the assessments provided in Sections 4.4.3 and 4.4.4 of the XPC LTR have some similarities to the basis information from the DCA LTC PIRT. NRC staff found that the rationale for the rankings in Table 3-4 of the XPC LTR are not always comprehensive. Therefore, the NRC staff is not making individual findings on each of the assessments made by NuScale in Sections 4.4.3 and 4.4.4. Rather, the NRC staff is considering the insights provided by the assessments from NuScale and considering them in combination with the code and overall methods assessments provided in the other sections of the XPC LTR.

13 The NRC staff found generally good agreement for the applicants XPC PIRT and the independent and related PIRT performed by the NRC staff but identified some areas that were not fully addressed in the PIRT. These potential items are addressed by consideration for other aspects of the applicants methodology, bounding sensitivities performed by the applicant, and confirmatory analysis performed by the NRC staff described in the relevant sections of this SER, with the exception of non-condensable gas effects on condensation heat transfer for boron transport. The NRC staff reviewed the statements about CNV Wall Heat Transfer provided in Table 4-17 in the XPC LTR and submitted audited calculation results (ML24346A360) performed by the applicant with respect to the impact of non-condensable gas effects for boron transport. NuScale states in Table 5-5 of the XPC LTR that non-condensable gas contributes to the overall pressure and, as a result, the RPV pressure, and saturation temperature is higher compared to cases where non-condensable gasses are not present. This impact is also shown in the calculation audited by the NRC staff. However, given the overall increased pressure, less steam is condensed in the CNV and the resulting RRV flow containing boron is reduced. Therefore L/C 3 has been developed to ensure that minimal non-condensable gases are in the overall system by requiring (1) the CNV to be maintained at a vacuum with insignificant initial non-condensable gas and safety related means to recombine or remove the gases released from the RPV after ECCS is initiated or (2) a calculation is presented in the application that demonstrates that the amount of non-condensable gases present do not impact the margin to re-criticality. Based on these findings, and subject to L/C 3, the NRC staff find reasonable assurance that the applicants PIRT is adequate in the context of the applicants broader methodology. 4.5 Evaluation Model Overview and Computational Tools NRC staff reviewed Section 4.1 of the XPC LTR descriptions for EM overview and computational tools to determine whether the descriptions of the overview and computational tools used are suitable for performing XPC safety analyses. 4.5.1 Evaluation Model Overview In Section 4.1.1 of the XPC LTR, NuScale correlates the EM FOMs of collapsed liquid level, subcriticality, and coolable geometry (Section 3.3 of the XPC LTR) with the EM acceptance criteria summarized below (Section 2.3.2 of the XPC LTR), respectively: Collapsed liquid level remains above the top of active fuel, demonstrating adequate decay heat removal for at least 72 hours after event initiation. The core remains subcritical (keff < 1) assuming the highest worth control rod stuck out (WRSO), for at least 72 hours after event initiation. Boron concentration remains below precipitation limits, supporting demonstration that coolable geometry is maintained. The Section 4.1.1 of the XPC TR further summarizes the calculational framework through Figure 4-1 and describes the computational tools used to evaluate the EM against the acceptance criteria.

14 The NRC staff reviewed the correlation between the FOMs and the acceptance criteria and find that they are adequate for the calculational framework in the XPC EM. 4.5.2 Computational Tools Section 4.1.2 of the XPC LTR discusses and lists the computational tools used for XPC EM evaluations. NuScale describes three computational devices: NRELAP5 system thermal-hydraulic code CMS5 code suite comprising the lattice physics code CASMO5, linkage code CMSLINK5 for nuclear data library generation, and core simulator code SIMULATE5 boron transport calculation scripts implemented in MATLAB or other appropriate computational script for efficiency in the calculation process NuScale states that NRELAP5 is NuScales system thermal-hydraulics code used to simulate the NPM system response to design basis events, including AOOs and postulated accidents and which is described in detail in the LOCA LTR (Reference 1). The NRC staff notes that further details about NRELAP5 and how it is used for non-LOCA events is described in the non-LOCA LTR (Reference 2). NRC staff confirmed that the NRELAP code is adequately described in the LOCA LTR for LOCA design basis events and non-LOCA topical report for non-LOCA design basis events. The NRC staffs most recent review of the NuScale NRELAP5 computer code for the NPM-20 focused on NuScale changes and additions after NRELAP5 Version 1.4, which is the NRELAP5 code version used for the NuScale NPM-160 DCA. The general applicability of the NRELAP5 (Version 1.4) code changeswhich is to say, what makes NRELAP different from RELAP5-3Dto a NuScale NPM were reviewed by the NRC staff during the DCA reviews. In the time since the completion of the DCA review, there have been NRELAP5 code version updates (now Version 1.7), changes to NPM design, and the scope of NuScales LOCA LTR changed. The changes in code version and NPM model version changes due to the changes in the NPM design were evaluated by the NRC staff in the SERs for the LOCA and Non-LOCA LTRs (References 1 and 2). Considering both the code and NPM design changes, NRELAP5 Version 1.7 is generally applicable for use in the XPC EM, subject to L/C 8 of this SER. NuScale states that the CMS5 code suite is used for nuclear analysis of the NPM reactor core and is described in the Nuclear Analysis Codes and Methods Qualification LTR, TR-0616-48793-P-A, Revision 1 (ML18348B036). NRC staff confirmed that the CMS5 code suite is adequately described in TR-0616-48793-P-A. NuScale states that boron transport calculation scripts implement the boron transport analysis methodology in MATLAB or another computational script to perform the calculational process. The calculational process that uses a script is described in Sections 6.2 and 6.2.5 of the XPC LTR. The NRC staff finds that a computational script that implements the calculations for the XPC boron transport methodology described in Section 6.2.5 of the XPC LTR is adequate for use in performing XPC EM calculations.

15 4.5.3 Quality Assurance Requirements The XPC EM, computational tools and calculational process are covered by the NuScale QA requirements. The QA requirements are addressed in NuScale Topical Report: NuScale Power, LLC Quality Assurance Program Description," MN-122626-A, Revision 1. The NRC staff reviewed the QA requirements and documented its approval in its SER of the topical report (ML24033A318). Further, the NRC staff inspected NuScales design control process and code development procedures. These inspections are documented in inspection report dated April 12, 2024 (ML24099A129). Subsequent to the inspection, based on the NRC staffs review of the categorization of some of the engineering documents and calculations underlying portions of the XPC EM TR, the staff requested NuScale to confirm that information presented or conclusions stated within the XPC EM TR are drawn from engineering documents subject to design verification in accordance with the NuScale QAPD, Section 2.3.1, Design Verification. As part of the NRC staffs audit, NuScale made available to staff a list of all engineering documents supporting such information or conclusions in the XPC TR where NuScales classification of these documents did not require design verification in accordance with QAPD, Section 2.3.1. Based, in part, on that list, the NRC staff identified a set of documents for NuScale to confirm the level of design verification that had been performed. NuScale responded and identified that it had confirmed that all but one of the documents had met applicable verification requirements of Appendix B to 10 CFR Part 50 and ASME NQA-1. The NuScale response stated that the confirmation was made through a review of applicable procedural instructions and authentication for each of the records, including the assigned roles of the signatories. The one remaining document was confirmed to have been subject to software integrity level 3 (safety-related) software controls that require appropriate verification. Based on the statements made by NuScale in response to this question, the staff finds with reasonable assurance that QA controls consistent with RG 1.203 have been implemented for the non-LOCA EM. 4.6 NRELAP5 Computer Code and Assessment Basis The Section 4.2 of the XPC LTR states that as part of LOCA EM development (Reference 1) and non-LOCA EM development (Reference 2), NRELAP5 was validated against a range of legacy test data, NuScale-specific tests performed to validate prediction of separate effects phenomena, and NuScale-specific integral effects testing performed to assess code prediction of integral effects during the short-term LOCA transient and short-term non-LOCA transients. Additionally, Section 4.2 of the XPC LTR states that tests that were performed to support the LOCA LTR were performed at the NIST-2 test facility and, though they were designed to support the short term LOCA response, the tests were executed for 24 hours, which contains data during the ECCS LTC phase. Therefore, the NuScale validation of NRELAP5 for XPC is mainly supported by the integral effect test runs performed at NIST-2. As mentioned in Section 4.5.2 of this SER, the NRELAP5 computer code is described in the LOCA and non-LOCA LTRs. The constitutive models used in the XPC LTR are the same as those described in the LOCA and non-LOCA LTRs. The assessment bases for those models are approved as part of those LTR reviews for the specified FOMs. Therefore, the NRC staff review of the NuScale NRELAP5 computer code assessment for the XPC EM focused on specific modeling and phenomena needed for assessment and validation for the XPC phase, following LOCA and non-LOCA events, and for boron transport.

16 4.6.1 NRELAP5 Assessments Section 4.2 of the XPC LTR summarizes the NuScale-specific integral effects tests performed at the NIST-2 facility to support NRELAP5 code validation during extended ECCS cooling. NuScale discusses the comparison of the NRELAP5 analysis of these integral effects tests versus experimental data in Section 4.2 (ML24305A290) and related subsections and presents their justification of the adequacy for modeling of the important phenomena during the LTC phase. NRC staff reviewed the integral effects tests and focused on determining the acceptability of the NuScale XPC evaluation methodology for performing design basis XPC analyses. This NRC staff review was limited to the applicability of NuScale methodology and use of the NRELAP5 computer code to perform XPC analysis for the LTC phase for design basis events. The NRC staff reviewed the following tests, and the evaluations are presented in Sections 4.6.2 through 4.6.5 of this SER: NIST-2 LTC extended ECCS tests NIST-2 LOCA ECCS tests NIST-2 Non-LOCA Tests The NRC staff observed that the tests listed above do not include conditions where the steam generator (SG) and DHRS are in operation after ECCS actuation. Section 4.4.3.24 of the XPC LTR states that during extended ECCS operation, with the SG and DHRS in operation, NRELAP5 sensitivity analyses show that varying the SG heat transfer coefficient over a range of 6.25 to 700 percent has a negligible impact on minimum riser collapsed liquid level, CNV pressure and moderator temperature. The NRC staff reviewed and audited details about NuScales NRELAP5 sensitivity analysis and information (ML25008A172 (nonproprietary) and ML25008A173 (proprietary)). In addition to the sensitivity analysis results, the submittal provided a thermal resistance evaluation. The results of the thermal resistance evaluation in combination with the NRELAP5 sensitivity analysis results confirm that after ECCS actuation, the heat transfer response with respect to the SG does not have a significant impact on the integrated system long-term response conditions. The NRC staff finds that the treatment of the SG and DHRS in operation after ECCS actuation for the XPC analysis is adequate given that the heat transfer response with respect to the SG does not have a significant impact on the integrated system long-term response conditions. As discussed in the LOCA LTR safety evaluation (Reference 1), the staff noted that the NIST NRELAP5 model was updated to include riser flow leakage paths of instrumentation taps, that previously was unaccounted for. These leakage paths from the riser to the downcomer, although not scaled to the NPM-20, were used in a riser hole sensitivity assessment provided by the applicant (ML25030A348 (nonproprietary) and ML25030A349 (proprietary)). The purpose of the sensitivity study was to validate the effect of riser hole modeling used in NPM methodology analyses and NIST integral test assessments to confirm that NRELAP5 modeling shows expected phenomenological results from the presence of riser holes. The applicant provided an assessment of selected LOCA and non-LOCA accidents and NIST benchmarks to show the sensitivity in model results in response to variations in riser hole size. Based on the results of the various assessment sensitivity studies, the staff agreed with the applicants conclusion that the NRELAP5 model responses are consistent with physics-based expected results and that

17 there are very negligible effects on the event FOMs as described in the LOCA LTR safety evaluation (Reference 1). 4.6.2 Test Facility The NIST-2 facility is a scaled, non-nuclear reactor that uses electric heater rods to represent the heat produced from fission. The NIST-2 facility is a modification of the NIST-1 facility and is described in the LOCA LTR. The NIST-2 test facility is intended to model a scaled representation of the major NPM components with minimum distortions relative to the actual NPM in order and provide the measurements necessary for validation of the NRELAP5 model used for safety analysis. Even though NuScale attempted to minimize the distortions between the NIST-2 scaled test facility and the NPM, NRC staff notes that distortions cannot be eliminated. Therefore, NRC staff evaluated the NIST-2 facility design and tests for NRELAP5 code evaluation against important XPC phenomena and not as testing to directly evaluate the safety or acceptability of the NPM-20 design. The NRC staff reviewed the NIST-2 facility design and determined that the areas of potential distortions were adequately considered because of the NRELAP5 validation activities documented in the LOCA TR EM, and that the thermal hydraulic phenomena are relatively benign during the LTC phase in comparison to the short-term LOCA phases with respect to their respective FOMs, in particular collapsed liquid level above the core. 4.6.3 NuScale NIST-2 Extended ECCS Tests (LTC-01) Assessment The validation of NRELAP5 for extended ECCS cooling is supported by six integral effect test runs performed at NIST-2 (LTC-01 testing program). The NRELAP5 validation benchmarks were performed for five NIST-2 tests which were initiated with a broken chemical and volume control system (CVCS) discharge line break, and a sixth test run simulated an inadvertent opening of an RVV (Run 1 through Run 6). Blowdown of the NIST-2 RPV inventory into containment resulted, with the event transitioning to ECCS recirculation and then to LTC. The entirety of the NIST-2 facility was used for the tests except for the DHRS. The tests continued for 24 hours. From these tests, the applicant compared the experimental data to the NRELAP5 predictions for multiple parameters. For each of the runs, the applicant compared the NRELAP5 results that included the following key parameters: (1) CNV pressure, (2) RPV pressure, (3) CNV level, and (4) RPV level. Additional parameters were compared and reviewed by the NRC staff to confirm the behavior of the tests relative to the key parameters. The NRC staff reviewed and audited details about NuScales NRELAP5 nodalization model for NIST-2, which is similar to the model used for the NPM-20. The NRELAP5 model is a complete one-dimensional representation of the NIST-2 test facility. The NRC staff finds that the NuScale NIST-2 NRELAP5 model provides an acceptable representation of the NIST-2 test facility in order to evaluate the capability of NRELAP5 to model NIST-2 tests. The NRC staff reviewed NuScales test matrix given in Section 4.2.2 of the XPC LTR and notes that the test suite covers a LOCA scenario with ECCS actuation with variations (( }} The NRC staff finds that the suite of tests, in combination with the validation activities

18 from the LOCA LTR, is sufficient to benchmark the NRELAP5 computer code and justify its use for XPC analyses for extended ECCS cooling. The NRC staff reviewed and audited details of the scaling for the ECCS long term cooling tests (LTC-01) described in Section 4.2.2 of the XPC LTR and notes that a distortion related to the scaling of the decay heat in the tests is present. (( }} Based on the information and description provided in the XPC LTR and additional information supplied by the applicant (ML24346A343), the NRC staff finds that the distortion is acceptable for the LTC phase given (1) (( }} (2) (( }} the discrepancy is reduced in terms of the scaling, and (3) no new phenomena are expected that are safety significant. The NRC staff reviewed the sequence of events for LTC-01 Run 1 provided in Table 4-3 of the XPC LTR. The table compared the sequence timing between the test data and the simulation. The NRC staff finds that the sequence timing for the test data and the simulation match well. In Sections 4.2.2.5 through 4.2.2.10 of the XPC LTR, NuScale compared the experimental data to the NRELAP5 predictions for the six runs in the LTC-01 test suite. The comparison showed reasonable agreement during the LTC period for the parameters of RPV pressure, CNV pressure, RPV downcomer level and CNV level for the base run (Run 1). The comparisons for the CNV level show reasonable agreement for the runs, however Run 2 through Run 6 show that the NRELAP CNV level (( }} but still close to the uncertainty band. However, NRELAP5 (( }}. For the remaining run (Run 2 through Run 6) comparisons, NRELAP (( }} more than in Run 1. NuScale stated that the predicted RPV level is (( }} Section 4.2.3.5 of the XPC LTR discusses the sensitivities performed by NuScale for LOCA Run 1, where the parameters of (( }}. The NRC staff confirmed that the LOCA runs documented in Section 4.2.3 of the XPC LTR show (( }} pressures and RPV level as seen in the extended ECCS runs presented in Section 4.2.2 of the XPC LTR. The NRC staff reviewed the sensitivity studies performed in Section 4.2.3 of the XPC LTR and confirmed that (( }} then the comparison results for pressures and RPV level could be reasonable. The staff reviewed and audited details of sensitivity analyses for conservative treatment of the CPV pool boundary conditions (ML25008A172 (nonproprietary) and ML25008A175 (proprietary)) which show that the conservative treatment of the pool boundary conditions in the sensitivity studies lead to a response that is (( }}. The CPV level and temperature comparisons show that NRELAP5 (( }} parameters. NuScale states that the mechanism for energy removal in the CPV is heat transfer ((

19 }} and would impact the temperature response. The behavior in the rest of the comparisons generally has reasonable trends with respect to the data and reflects some of the inconsistencies observed in the pressure, level and CPV parameter comparisons. Overall, the NRC staff finds that the NRELAP5 predictions of the NIST-2 tests (LTC-01) are acceptable for long term cooling analyses for the collapsed liquid level figure of merit since the CNV and RPV pressure and level show adequate agreement, when considering the sensitivities performed and when accounting for (( }}, and also taking into account the conservative modeling considered in Section 5.4 and 4.4.3.32 of the XPC LTR that adequately addresses the boundary conditions used for the pool. The NRC staff also agrees that modeling in the cooling pool cannot capture all the realistic phenomena and that small differences in the predictions of cooling temperature and level are not crucial to the success of the benchmarks. 4.6.4 NuScale NIST-2 LOCA Test Extended ECCS Cooling Assessment The validation of NRELAP5 for LOCA extended ECCS cooling is supported by seven integral effect test runs performed at NIST-2 to replicate four NIST-1 tests: HP-06, HP-07, HP-09, and HP-49 for transient scenarios (( }}. The NRELAP5 validation benchmarks were performed for NIST-2 tests that were initiated with: 1) 100 percent CVCS discharge line break,

2) 100 percent High Point Vent line break, 3) Inadvertent opening of a single RVV and 4)

Inadvertent opening of a single RRV. Blowdown of the NIST-2 reactor pressure vessel inventory into containment resulted, with the event transitioning to ECCS recirculation and then to long term cooling. The entirety of the NIST-2 facility was used for the tests except for the DHRS. The tests continued for 24 hours. From these tests, the applicant compared the experimental data to the NRELAP5 predictions for multiple parameters. For each of the runs, the applicant compared the NRELAP5 results that included the following key parameters: (1) CNV pressure, (2) RPV pressure, (3) CNV level, (4) RPV level. Additional parameters were compared and reviewed by the NRC staff to confirm the behavior of the tests relative to the key parameters. The NRC staff reviewed and audited (ML24263A009) details about NuScales NRELAP5 nodalization model, which is similar to the model used for the NPM. The NRELAP5 model is a complete one-dimensional representation of the NIST-2 test facility. NRC staff finds that the NuScale NIST-2 NRELAP5 model provides an acceptable representation of the NIST-2 test facility in order to evaluate the capability of NRELAP5 to model the NIST-2 tests. The NRC staff reviewed NuScales test matrix, given in Section 4.2.3 of the XPC TR, and notes that the test suite covers a range of LOCA scenarios with ECCS actuation. The NRC staff finds that the suite of tests, in combination with the validation activities from the LOCA TR, is

20 sufficient to benchmark the NRELAP5 computer code and justify its use for XPC analyses for extended ECCS cooling. The NRC staff reviewed the sequence of events for LOCA extended ECCS cooling Run 1, Run 3 and Run 4 (100 Percent CVCS Discharge Line Break Case, Inadvertent RVV Opening Case and Inadvertent RRV Opening Case, respectively) provided in XPC TR Tables 4-5, 4-6 and 4-7. The tables compared the sequence timing between the test data and the simulation. The NRC staff finds that the sequence timing for the test data and the simulation match well for Table 4-5 and Table 4-7. The TR sequence of events for Table 4-6 Run 3 did show differences in valve actuation timing. However, this difference is likely due to (( }} as discussed below. In XPC topical report Sections 4.2.3.5 through 4.2.3.7, NuScale compared the experimental data to the NRELAP5 predictions for three Runs (Run 1, Run 3 and Run 4) in the LOCA extended ECCS cooling test suite. The comparison showed reasonable agreement during the long-term cooling period for the parameters of RPV pressure, CNV pressure, and CNV level for Run 1, Run 3 and Run 4. The comparisons for the RPV level show that NRELAP5 (( }} while the CNV level has reasonable agreement. NuScale stated that the predicted RPV level is (( }} Section 4.2.3.5 discusses the sensitivities performed by NuScale for LOCA Run 1, where the parameters of (( }}. The NRC staff confirmed that the LOCA runs documented in TR Section 4.2.3 show (( }} pressures and RPV level as seen in the extended ECCS runs presented in TR section 4.2.2. The NRC staff reviewed the sensitivity studies performed in TR section 4.2.3 and confirmed that (( }}, then the comparison results for pressures and RPV level could be shown to be reasonable. NuScale states (( }}. The staff reviewed and audited details of sensitivity analyses for conservative treatment of the CPV pool boundary conditions (ML25008A172 (nonproprietary) and ML25008A175 (proprietary) which show that the conservative treatment of the pool boundary conditions in the sensitivity studies lead to a response that (( }} Overall, the NRC staff finds that the NRELAP5 predictions of the NIST-2 tests (LTC-01) are acceptable for long term cooling analyses for the collapsed liquid level figure of merit since the CNV and RPV pressure and level show adequate agreement when considering the sensitivities performed and when accounting for ((

}}, and also taking into account the conservative modeling considered in TR Section 5.4 and 4.4.3.42 that adequately address the boundary conditions used for the pool. The NRC staff also agrees that modeling in the cooling pool cannot capture all the realistic phenomena and that small differences in the predictions of cooling temperature and level are not crucial to the success of the benchmarks. 

21 4.6.5 NuScale NIST-2 non-LOCA SG/DHRS Test Assessment In Section 4.2.5 of the XPC LTR, NuScale references the non-LOCA LTR (Reference 2) for NRELAP5s ability to simulate total DHRS heat removal. (ML24305A290). The non-LOCA NIST-2 tests are described in non-LOCA LTR Section 5.3.7. The NIST-2 test results were reviewed by the NRC staff. The NRC staff noticed that some of the results in the longer term showed some differences in the test results vs NRELAP5, although the trends are generally in agreement. The following differences in the comparisons are some illustrative examples of those noted by the NRC staff. Non-LOCA Figure 5-220 shows that RPV pressure is (( }} for the long term and SG steam drum pressure is (( }}. Non-LOCA Figure 5-221 shows that the SG level and DHRS level are (( }}. Non-LOCA Figure 5-222 shows that steam drum level is (( }} by NRELAP5. Non-LOCA Figure 5-223 shows that the DHRS inlet and outlet header predictions are (( }}. Non-LOCA Figures 5-228, 5-230 and 5-231 show that the RPV level, pool level and pool temperature is (( }}. As noted in Sections 4.6.3 and 4.6.4 of this SER, these would be expected to be closer if the boundary conditions for the cooling pool are appropriately accounted for to achieve a conservative result. The staff reviewed and audited details of sensitivity analyses for conservative treatment of the CPV pool boundary conditions (ML25008A172 (nonproprietary) and ML25008A175 (proprietary)) which show that the conservative treatment of the pool boundary conditions in the sensitivity studies lead to a response that is conservative relative to the physical response. During the audit (ML24263A009) the NRC staff noted that in the NIST-2 non-LOCA RUN2 test, the test data shows oscillations in the condensate flow. Additionally, the NRC staff observed that NRELAP5 runs (( }}. The NRC staff finds that the nodalization and boundary conditions are adequate given the information reviewed and that (1) when the SG is uncovered the response is insensitive to the treatment of the SG-DHRS as shown in information provided by the applicant and (2) using the nodalization and boundary conditions for the pool as described in information provided by the applicant (ML25008A172 (nonproprietary) and ML25008A175 (proprietary)) would lead to a conservative result in the NPM-20 XPC long term analyses. Overall, the NRC staff finds that the NRELAP5 predictions of the NIST-2 tests are acceptable for LTC analyses for the collapsed liquid level FOM when considering the sensitivities performed for the LOCA tests, and that the conservative modeling considered in Sections 5.4 and 4.4.3.42 of the XPC LTR adequately addresses the boundary conditions used for the pool. The NRC staff also agrees that modeling in the cooling pool cannot capture all of the realistic

22 phenomena and that small differences in the predictions of cooling temperature and level are not crucial to the success of the benchmarks as long as the modeling is treated conservatively. 4.7 Extended Passive Cooling Thermal Hydraulic Analysis Methodology Evaluation Section 5.0 of the XPC LTR and related subsections describe the NuScale NPM-20 XPC thermal hydraulic methodology. The XPC LTR analysis methodology credits the ECCS and DHRS for long term decay heat removal. The XPC LTR states that the analysis demonstrates that the top of active fuel remains covered and maximum temperature cases remain within pressure and temperature limits for the RPV and CNV, and the XPC LTR provides minimum temperature cases to provide boundary conditions for reactivity control, boron transport and boron precipitation analyses. 4.7.1 Description of Extended Passive Cooling Scenarios The long-term core cooling phase starts after DHRS and/or ECCS is actuated and the NPM reaches a quasi-steady state condition such that steam from the PZR region of the RPV is released to the CNV through the RVVs, the steam is condensed on the CNV walls, and the condensed liquid flows from the CNV through the RRVs back into the downcomer core inlet. This recirculation flow loop continues, and the NPM-20 is gradually cooled. This LTC configuration is reached through both LOCA and non-LOCA initiating events. The non-LOCA initiating events generally involve DHRS cooldown for an extended period and can subsequently transition to ECCS actuation and ECCS cooling. Consequently, meeting the XPC EM acceptance criteria for the XPC FOMs must be demonstrated for both LOCA and non-LOCA events. NuScale describes the characteristics of XPC in Section 5.1 of the XPC LTR. Depending on the initiating event and the event progression, there are three long term heat removal scenarios: (1) Early ECCS actuation following a design basis event such as a LOCA or an inadvertent opening of an ECCS valve (2) DHRS cooling after a non-LOCA or small beak LOCA event followed by eventual ECCS actuation (3) DHRS cooling without ECCS actuation for an extended duration Scenario 1 In the first scenario, the ECCS valves open early in the event progression at relatively high RCS temperature and pressure conditions and provide decay and residual heat removal by transferring heat to the containment pool via the CNV walls. The RVVs open first followed by the RRVs once the IAB differential pressure threshold is reached. The steam released from the RCS through the ECCS valves condenses on the CNV wall and collects at the bottom of the CNV. Some of the condensed steam is collected from the CNV wall and directed into baskets that contain boron which is dissolved and transported with the condensate to the bottom portion of the CNV. Once the recirculation loop is established, the level in the RPV riser is relatively low but remains above the core and above the lower riser hole elevation.

23 The design of the lower riser holes is intended to allow recirculation of liquid from the riser into the downcomer to maintain a mixed boron concentration in the RPV. Downcomer dilution can occur due to condensation forming on the SG tubes with the RPV water level below the top of the riser and upper riser holes. Scenario 2 In the second scenario, DHRS is actuated as a result of the event and provides decay and residual heat removal after the reactor scram; without initially actuating ECCS. As DHRS continues to remove heat, the RCS pressure and temperature conditions trend lower as decay heat is removed. During continued decay heat removal, the level in the RPV reduces below the top of the riser. Once the level drops below the top of the riser via DHRS cooling (or potential leakage), the upper riser holes are intended to be designed to allow sufficient mass flow rate between the downcomer and riser to support continued decay heat removal through RPV primary flow without a maldistribution of boron occurring between the downcomer and riser prior to ECCS actuation. Continued decreases in the RPV level through decay heat removal (or leakage) would lead to an ECCS actuation on the low-low level signal (some other ECCS actuation may occur also, i.e. ECCS actuation timer)1 before the upper riser holes are uncovered. After ECCS actuation, Scenario 2 proceeds similar to Scenario 1 except at lower pressure and temperature conditions. Additionally, the upper riser holes are intended to reduce downcomer dilution of the boron content due to condensation forming on the SG tubes during extended DHRS operation with RPV water level below the top of the riser, such that an unacceptable positive reactivity insertion upon opening of the ECCS valves is precluded. The upper riser holes are intended to be designed to allow sufficient mass flow rate of borated liquid between the downcomer and riser to maintain downcomer boron concentrations above the critical boron concentration prior to ECCS operation. Scenario 3 In the third scenario, DHRS is actuated as a result of the event and provides decay and residual heat removal after reactor scram and does not actuate ECCS for 72 hours. This scenario is similar to Scenario 2 except ECCS is never actuated on low-low level or for any other reason (e.g. bypass ECCS timer). Similar to Scenario 2, as DHRS continues to remove heat, the RCS pressure and temperature conditions trend lower with decay heat. During continued decay heat removal, the level in the RPV reduces, potentially to below the top of the riser. Once the level drops below the top of the riser via DHRS cooling, the upper riser holes are intended to be designed to allow sufficient mass flow rate between the downcomer and riser to support continued decay heat removal through RPV primary flow and prevent maldistribution of boron occurring between the downcomer and riser prior to ECCS actuation. At the end of this scenario, at 72 hours, the RPV pressure and temperature conditions are relatively low, with continued DHRS cooling, and the critical boron concentration is met. After the 72-hour period, 1 The Low-low riser level signal is 460-472 from bottom of the pool (38.3 to 39.3) -upper riser hole elevations (( }}.

24 the recovery actions from the post 72-hour conditions are not known and are not covered by the TR. 4.7.2 XPC NRELAP Model As discussed in Section 5.2 of the XPC LTR and related subsections, NuScale developed the NRELAP5 XPC input model from the detailed NRELAP5 basemodel developed for short term LOCA evaluation model calculations (Reference 1). The XPC NRELAP5 model is a coarser version of the LOCA EM NRELAP5 model with some simplifications and other changes made to run longer transient cases. The coarser XPC nodalization modeling is shown in Figure 5-11 of the XPC LTR. Additionally, some of the key XPC model differences are (( }}. The reactor pool model (( }}. As indicated in Section 4.6 of this SER, the validation testing at the NIST-2 facility for the NPM-20 indicates that reasonable boundary conditions for the pool is needed to get an adequate comparison between the data and NRELAP5 model. NuScale described (( }} can adequately account for the uncertainty indicated by the integral testing performed for NRELAP5 performance. NuScale states that NRELAP5 is (( }}. The NRC staff reviewed the equations described by NuScale used in the analysis. The NRC staff noted that the xc value (also referred to as xt) used in the equations is experimentally derived for a given valve configuration and is part of the NuScale valve design specification. Since the value is experimentally derived for a given valve configuration and is part of the NuScale valve design specification, L/C 6 has been developed to ensure that value is verified as part of the ASME QME-1 qualification program to be consistent with the analyses. The NRC staff confirmed that the ((

25 }}. Since validation of the NRELAP5 code for XPC depends heavily on the LOCA LTR base model and assessments, NuScale benchmarked the XPC NRELAP5 model to the LOCA EM input model to show consistency of results. NuScale states that the LOCA comparison case is primarily interested in demonstrating that the ECCS LTC model produces results comparable to those produced by the LOCA model for pressure, temperature, and liquid level response following some transient event. (( }} (Figures 5-12 through 5-15 of the XPC LTR). These results show that the XPC model (( }} Therefore, NuScale states that the coarse model is used for identifying limiting cases for collapsed liquid level and the detailed model is used for the analysis as stated in section 5.2 of the XPC LTR. The staff finds that the coarse model is adequate for use in the boron transport methodology given the conservative treatment of the CNV pool boundary conditions and can be used for identifying limiting cases for collapsed liquid level for evaluation with the detailed model. 4.7.2.1 Lower Riser Hole Flow Evaluation Section 5.2.3 of the XPC LTR provides the methodology used to assess lower riser hole flow during ECCS cooling (low riser level) for use in boron transport analysis. The lower riser hole flow rate assessment is used to determine the flow rate used in the boron transport analysis or for justification for use of a conservative riser hole flow rate. The lower riser hole flow evaluation calculations described in the XPC LTR are a requirement of the methodology and the NRC staff safety evaluation finding. The evaluation calculation uses and results are equivalent to the other topical report methodology analysis results where 10 CFR 50 Appendix B verification is required and applied to successfully implement the methodology for the NRC staff to make a finding for an application. The boron transport analysis is described in Section 7.0 of the XPC LTR and evaluated in Section 4.8 of this SER. Flow through the lower riser holes is used to increase the mixing between the core/riser and downcomer regions, and higher riser hole flow rates increase mixing between the core/riser and downcomer regions and means that there is overall more mixing in the system (more flow from the CNV into the RPV). NuScale states (( }}

26 (( }} (( }} The NRC staffs review and audit (ML24263A009) of NuScales treatment of conditions focused on the calculation of the two-phase mixture level in the riser region (ML24346A356). The NRC staff performed sensitivity calculations and assessment of a conservative two-phase mixture level in the riser. The analytical prediction of mixture level behavior, like many two-phase flow phenomena more generally, remains subject to empirically based limitations and uncertainties. Further, as discussed above, the mixture level prediction in the riser region plays a key role in driving the mixing flow between the riser and downcomer. Therefore, the NRC staff paid particular attention to assuring that the applicants calculation of riser mixture level contains adequate conservatism. As described below, the NRC staffs review focused on the following general areas: Applicability of the (( }} approach proposed by the applicant, including selected correlations and input parameters (e.g., bubble diameter), to the conditions present for the NuScale reactor design. Validation of the (( }} and calculational methods at conditions representative of the NuScale reactor design. Sensitivity calculations performed by the NRC staff, and independent evaluations of different (( }}. Assessment of the adequacy of the conservative margins associated with the applicants calculated results. The NRC staff considers it essential to adequately assure the applicability (( }} to the NuScale reactor design in light of

27 the unique reactor design and geometry, as well as the use of the correlation over a reduced pressure range. Ensuring the applicability of empirical correlations is generally best accomplished through validation of the correlations, as implemented using the applicants calculational procedures, against representative test data. However, the NRC staff's audit review found no evidence that the applicant had performed a validation of (( }} at representative conditions using its intended calculational procedure. In light of these limitations associated with the validation of the (( }} proposed by NuScale, the NRC staff's audit (ML24263A009) and review focused upon performing independent assessments and sensitivity calculations to assess the adequacy of the applicants modeling of the two-phase mixture level. The assessments performed by the NRC staff included comparisons of industry drift flux correlations to the applicants calculations with (( }}, as well sensitivity calculations which considered the impact of the uncertainties in the (( }}. NuScale applies the (( }} to the core. This correlation tends to underpredict drift velocities and over predict void at the core exit. Predictions are adequate and better at void fractions below 15% but underpredict drift velocity above this void fraction. The higher void at the core exit also produces higher void in the riser section along with increased flow rates through the riser holes. This produces higher boron content in the core and lowers downcomer boron for potential dilution events. More mixing is promoted that lowers core pressure which increases void in the core, resulting in conservative conditions for both precipitation and dilution events. Lower drift-velocities increase void fraction which maximizes boron content in the core for precipitation and reduces potential for return to power for dilution. Higher riser void produces an increase in riser flow and enhances mixing in the core and downcomer regions. As such, low drift velocity is conservative for key phenomena governing both precipitation and dilution events. Literature reviews of drift flux models and correlations used in the industry under-predict drift velocity and over-predict void, as does Ishiis drift flux correlations for bubbly and churn turbulent flow regimes. The Dix model is routinely mentioned as a better correlation over the full range of void fractions, (it more accurately predicts void and level swell in test data with heated rod bundles). Since the drift flux correlations over predict level swell (and void) because of the low drift velocities (typically predict velocities below 1 ft/sec where at low pressure near 14.7 psia, the drift velocity is 3.0 ft/sec and higher for even lower pressures). So, use of the drift velocity correlations for precipitation and dilution is conservative because it maximizes void and boric acid concentration (less liquid in the mixing volume). The NRC staffs sensitivity calculations compared with results from using (( }} for lower riser hole flow rate showed that using reasonable riser hole flow rates would provide reasonable results in comparison with the uncertainty in the drift flux models. Finally, the NRC staff's review weighed the conservative margins associated with the applicants calculation of two-phase mixture level against uncertainties identified during the audit (ML24263A009) and review. The NRC staff reviewed the conservatisms identified by NuScale ML24346A356. While the NRC staff recognizes that some of the applicants conservative modeling practices may have countervailing impacts (e.g., overestimating decay heat tends to increase the predicted mixture level), when taken as a whole, the NRC staff agreed that the set

28 of conservatisms identified by the applicant provides additional margin, albeit unquantified, in the calculated mixture level. In the NRC staff's judgment, incorporation of adequate conservative margin in the two-phase mixture level calculation is essential to offset inherent calculational uncertainties. For instance, as discussed above, the applicants calculations did not include validation of its proposed (( }} and their implementation. In light of the conservatisms in the applicants calculations, the NRC staff considered these uncertainties to have been adequately addressed. The NRC staffs review found that there is reasonable assurance that the mixture level in the riser is conservatively calculated by the applicant. A lower flow rate though the lower riser holes means that there is lower void at the core exit and a lower void in the riser section along with the decreased flow rates. This produces lower boron content in the core (less boiling) and increases the downcomer boron for potential dilution events. Less mixing is promoted that increases core pressure which reduces the void in the core, resulting in less conservative conditions. Additionally, more voiding in the core exit and riser increases the steam flow to the CNV which drives condensation and additional boron from the CNV into the RPV from the RVVs. The NRC staff's conclusion is based primarily upon the following points that have been elaborated further in the discussion above: The applicant ensured that its calculated void fractions in the riser are conservative because of the conservative assumptions used in the calculations. The NRC staff performed sensitivity calculations for lower riser flow rates and independent evaluations using additional correlations with diverse validation bases. The results of these assessments and calculations support a finding that the mixture level calculated by the applicant is conservative. As discussed above, the applicant included sufficient conservatism in its calculation of the two-phase mixture level to address calculational uncertainties expected at the low-void conditions. (( }} to the nominal loss form. The NRC staff agrees that a conservative loss coefficient is applied, which is a reasonable uncertainty as compared to the nominal hole form loss. 4.7.3 Events Evaluated Section 5.3 and associated subsections of the XPC LTR describe the thermal hydraulic events evaluated using the XPC EM methodology. The NRELAP5 calculations are performed starting from event initiation and are run until quasi-steady conditions are reached. Then statepoint analyses are used to evaluate the later time periods using the conditions beginning from the end of initial calculation, with set boundary conditions for decay heat and pool level and temperature. The NRC staff audited and reviewed the statepoint calculation results (ML24346A362) in comparison to the initial calculations without using the statepoint method. Although direct comparison to initial calculations that dont use the statepoint method were not presented for the

29 long term, the NRC staff finds that the statepoint calculations are adequate to represent the long term results given that the initial calculations are run for 12 hours where most relevant phenomena is captured and run to 24 hours for minimum pool level and the results from the statepoint method align reasonably well with the initial calculations. The XPC LTR thermal hydraulic analysis methodology addresses the calculation of the minimum collapsed level, residual and decay heat removal capability and thermal hydraulic conditions considered in the boron transport analysis methodology. The minimum collapsed liquid level calculation is impacted by the RCS inventory (including isolation timing), and timing for ECCS. Minimum level is calculated via maximum cooldown with minimum RCS inventory and maximum losses to the CNV to confirm that the collapsed liquid level is maintained above the active fuel. The events analyzed include (1) the full LOCA spectrum for pipe breaks inside and outside containment, (2) SG tube failure, and (3) inadvertent operation of the ECCS. Calculations for RCS heat removal capability and boron transport thermal hydraulic conditions include (1) Maximum temperature via minimum cooldown during conditions biased to minimize decay heat removal and maximize module temperatures to justify sufficient ECCS capacity to maintain level above the top of the reactor core, and RPV and CNV integrity, (2) Minimum temperature via maximum cooldown to confirm that the collapsed liquid level is above the active fuel and that the minimum RCS temperature precludes boron precipitation during the LTC evaluation period with conditions biased to evaluate effective decay heat removal and the coolest moderator conditions for boron precipitation and subcriticality. Thermal-hydraulic results from the NRELAP5 calculations are used in downstream boron transport analyses to assess the margin to subcriticality and to boron precipitation. The events evaluated are: (1) large liquid and vapor space breaks inside containment with ECCS actuation early in the transient, (2) large and small breaks outside containment and (3) CVCS injection line liquid space break inside containment (( }}, and (4) non-LOCA events that add dilute water to containment prior to ECCS actuation. The NRC staff finds that for the events evaluated during which ECCS is actuated, for the FOMs related to collapsed liquid level, residual and decay heat capability and boron transport, the thermal hydraulic analyses are adequate because they represent the most limiting events and are biased to the limiting conditions for the associated FOMs. For events that dont actuate ECCS, the DHRS is used to remove residual and decay heat. NuScale states that during DHRS operation, RCS inventory is retained inside the RPV such that maintaining the collapsed liquid level over the top of the core is not challenged. Additionally, NuScale states that the RPV liquid reaches cooler temperatures, with a higher boron concentration, during ECCS cooling compared to extended DHRS cooling; therefore, DHRS cooling conditions are non-limiting for boron precipitation analysis. The NRC staff confirmed that the collapsed liquid level is not a concern during DHRS cooling because a low water level would actuate ECCS and DHRS would no longer be the primary method of heat removal. Additionally, the NRC staff confirmed that temperatures are lower and the boron concentration is higher during the ECCS cooling phase.

30 NuScale states that the XPC EM extended DHRS cooling calculations are performed to demonstrate the system decay heat capacity for up to 72 hours under conditions challenging DHRS heat removal and provide input to boron dilution calculations with conditions biased to maximize the potential for boron redistribution. The events considered by NuScale are events for increases in secondary side inventory and events that disable one train of DHRS. Additionally, for boron transport evaluations, the small pipe break outside of containment and leak inside containment below the LOCA break range are evaluated. The NRC staff finds that for the events evaluated, for extended DHRS operation, the residual and decay heat capability and boron transport thermal hydraulic analyses are adequate because they represent the most limiting events and are biased to the limiting conditions for the associated FOM. 4.7.4 Initial Conditions and Biases Section 5.4 of the XPC LTR describes the key initial condition and boundary condition biases for extended ECCS cooling and extended DHRS cooling event NRELAP5 thermal hydraulic analysis. The initial conditions and boundary conditions used in the XPC LTR methodology analysis are selected to provide conservative event responses and RCS conditions with respect to minimum collapsed liquid level, decay heat removal capacity and boron transport. Therefore, NuScale used six general limiting sets of conditions: (1) Minimum Collapsed Liquid Level, (2) Heat Transfer Capacity - Maximum Temperature, (3) Boron Transport Analyses - Minimum Temperature, (4) Boron Transport - Sensitivities, (5) DHRS Cooling Cases Demonstrating Decay Heat Removal and (6) Extended DHRS Cooling Cases Providing Boron Transport Analyses. NuScale considered a broad range of assumptions and initial conditions in the methodology to determine the limiting responses and conditions for the associated FOM, including power availability, decay heat, single failures, reactor pool conditions, RCS conditions, riser hole loss, non-condensable gas effects, ECCS valve capacity and PZR conditions. The NRC staff reviewed the initial conditions and boundary conditions described in Section 5.4 of the XPC LTR. Table 5-5 of the XPC LTR states that non-condensable gas is biased low for boron transport analyses. As discussed in Section 4.4.3 of this SER, biased low non-condensable gas may be nonconservative because higher amounts of gas cause increased pressure which would result in a decrease in RRV flow and less boron into the RPV. Therefore L/C 3 has been developed to ensure that minimal non-condensable gases are in the overall system by requiring (1) the CNV to be maintained at a vacuum with insignificant initial non-condensable gas and safety related means to recombine or remove the gases released from the RPV after ECCS is initiated or (2) a calculation is presented in the application that demonstrates that the amount of non-condensable gases present do no impact the margin to re-criticality. The NRC staff finds that the initial conditions and boundary conditions, subject to L/C 3, are adequate because they are set to provide the most limiting biases and conditions for the associated FOMs. The NRC staff also notes that the event-specific electrical power assumptions (AC/DC), single failures, and the need for operator actions necessary to mitigate

31 XPC events are reviewed by the NRC staff through a design review for application of the method to a specific design. As such, these assumptions are subject to L/C 9. 4.7.5 Representative Results Section 5.5 of the XPC LTR provides representative thermal hydraulic results using the XPC methodology. The representative calculations provide the response to the following events: (1) minimum level with discharge line break outside containment, (2) maximum temperature with inadvertent opening of an RVV with loss of AC and DC, (3) minimum temperature with inadvertent opening of an RVV with loss of AC and DC, and (4) minimum temperature with small liquid break outside containment coincident with loss of AC. The NRC staff reviewed the provided event descriptions and event response. Based on its review of NuScales representative calculations, the NRC staff concluded that the representative analyses provided adequate and expected results, appropriately illustrating that implementation of the methodology as specified in the XPC LTR, when implemented, will provide reasonable results appropriate for determining whether FOMs are met. 4.8 Evaluation for Reactivity Control and Boron Distribution The NuScale XPC thermal hydraulic analysis methodology is described in Section 5 of the XPC LTR. This NRELAP5 methodology is used to calculate the XPC cooldown for LOCA and non-LOCA transients. These NRELAP5 analyses do not include the impact of boron on coolant properties or track boron distribution. However, these NRELAP5 analyses are used to calculate inputs into the NuScale boron transport methodology as described in Sections 6 and 7 of the XPC LTR. This section of the SER discusses the evaluations associated with reactivity control, boron transport and boron distribution. Section 6 of the XPC LTR provides the NuScale methodology used to assess the NuScale acceptance criteria for maintaining subcriticality. NuScale relies on maintaining a collapsed liquid level above the top of active fuel to maintain a coolable geometry. 4.8.1 General Approach and Acceptance Criteria Section 6.1 of the XPC LTR describes the NuScale general approach and acceptance criteria for maintaining the boron concentration in the core required for subcriticality, and Section 7.1 of the XPC LTR describes the general approach and acceptance criteria used to show that the RCS boron remains below the temperature for the boron precipitation limit. The acceptance criterion used to assess subcriticality is to ensure that the calculated boron concentration in the core region remains above the calculated critical boron concentration for 72 hours. The acceptance criteria used to assess maintaining coolable geometry is to ensure that the boron remains below the solubility/precipitation limit to ensure that boron is available to be well mixed in the core region. The acceptance criteria are found to be appropriate because they are conservative with respect to determining criticality and boron solubility or precipitation. NuScale describes the general steps for performing subcriticality analyses and boron precipitation analyses. The general steps are mainly: (1) perform NRELAP5 thermal-hydraulic calculations biased for dilution or precipitation, (2) evaluate boron transport using thermal-

32 hydraulic analysis input and transport methodology, (3) determine, for subcriticality analyses, critical boron concentration for a range of conditions (4) compare, for subcriticality analyses, the boron transport analysis results to the appropriate critical boron concentration for the calculated conditions. For subcriticality analyses, the critical boron concentration is the amount of boron needed to prevent the core from going critical. For boron precipitation analyses, the saturation temperature of the CNV vapor space is compared to the solubility limit, or precipitation temperature for boric acid as a function of concentration in the core/riser mixing volume. 4.8.2 Boron Transport Subcriticality Methodology The boron transport methodology is presented in Section 6.2 of the XPC LTR and is intended to conservatively represent the boron concentration in the core region against the boron concentration required to remain subcritical. The boron transport methodology for boron precipitation is presented in Section 7.2 of the XPC LTR and is similar to the boron transport methodology used for subcriticality but conservatively maximizes the boron concentrations in the RCS mixing volumes to demonstrate the concentrations remain below the solubility limit. The NuScale methodology for boron dilution (( }}. Boron transport between the various volumes is evaluated by NuScale for several different transport mechanisms based on thermal hydraulic analyses. The NRC staff reviewed the overall method, which uses a combination of the LTC NRELAP5 model, providing only thermal-hydraulic conditions (e.g., average moderator temperature for a constant core power, reactor flow versus power, etc.), and SIMULATE5, capturing the reactivity components, such as moderator and Doppler feedback and rod worths. The NRELAP5 thermal hydraulic analyses provide the conditions vs time for limiting events for the CNV and RCS. Boron transport and concentration in each of the distinct volumes are then calculated based on the thermal hydraulic conditions and boron transport mechanisms. 4.8.2.1 Boron Transport Mechanisms NuScale described the boron transport mechanisms in Section 6.2.3 of the XPC LTR, which states that the transport terms minimize boron entering the core region and maximize boron leaving the core for the subcriticality calculations. For boron precipitation analyses, described in Section 7.2.3 of the XPC LTR, NuScale states that the transport terms maximize boron entering the core region and minimize boron leaving the core. (( }} A break in the injection line creates a boron transport pathway (( }}. The injection line transport pathway is (( }}.

33 A break in the discharge line creates a boron transport pathway between (( }}. (( }} The XPC LTR states (( }}. However, the amount of mixing and time in which it would occur was not provided nor was validation of the assumption presented. NuScale provided CFD analysis results in response to RAI-10298 XPC-2 (ML25030A348 (nonproprietary) and ML25030A349 (proprietary)) which show that the CNV is well mixed in 30 minutes. However, the CFD analysis does not consider the density difference between the borated and unborated liquid with respect to mixing. Therefore, the CFD analysis is nonconservative with respect to a best estimate model which considers the density difference. An independent confirmatory CFD analysis performed by the NRC staff confirmed that the density difference should be included and would impact the timing of the mixing in the CNV on the order of hours. Not accounting for the density difference leads to a delayed mixing time and nonconservative results at the beginning of events in which instantaneous mixing is assumed in the lower CNV. The density difference was not directly addressed by NuScale. However, NuScale stated in Section 9.0 of the XPC topical report that the boron dilution transport methodology results must demonstrate at least 25 ppm margin between the core region boron concentration and the critical boron concentration, or for cases with less than 25 ppm minimum margin, analysis demonstrates the methodology is conservative with respect to delayed onset of mixing between the upper and lower CNV volumes due to liquid density difference in the volumes. The response to RAI-10298 XPC-2 (ML25030A348 (nonproprietary) and ML25030A349 (proprietary)) states that the 25 ppm margin would account for an approximately 6-hour delay in mixing between the lower CNV and upper CNV. The NRC staff confirmed that a 6-hour delay would be adequate to account for the density difference between the borated and unborated liquid based on independent confirmatory analyses with respect to mixing. Therefore L/C 4 has been developed to ensure that the CNV mixing assumptions adequately consider the density difference between the borated and unborated liquid by requiring a 25 ppm margin

34 equivalent to a delay of at least 6 hours in mixing between the lower and upper CNV or an analysis that explicitly considers the density difference demonstrates that the methodology is conservative with respect to delayed mixing in the CNV. The CNV related transport terms are dependent on the as-built plant design response with respect to condensate flow rates, mixing tube flow rates, boron dilution and transport. Therefore L/C 7 was developed to ensure that these parameters are adequately reflected in the as built module, as well as confirm the mixing and transport assumptions utilized in the XPC EM are valid and adequate. This condition requires an applicant or licensee seeking to apply this methodology to a design to include an initial test for the first module only for dissolution testing during the initial test program. The test must be consistent with the analysis methods and code of record based on the predictions of the as-tested conditions. In other words, the test conditions and results would need to be compared to the results predicted by the code of record, when calculated using the XPC EM. The test must be designed to demonstrate acceptable performance of the as-built ECCS Supplemental Boron (ESB) based on predicted system response under expected test conditions to ensure the system as a whole meets the fundamental design requirements of the safety analysis. The test acceptance criteria must confirm the as-built functionality of the ESB, including boron basket dissolution rates, condensate rail collection capability, mixing tube flow, lower containment boron concentration and mixing. With respect to boric acid volatility, the NRC staff reviewed the applicants method to calculate the boron potentially lost due to plate out on the surfaces of the NPM during the long-term cooling period. The NRC staff notes that this mechanism is not a major contributor to boron loss early in the period but becomes more important as time progresses. The NRC staff evaluated (( }} (Reference 5) to determine the volatility of the boric acid in the NuScale design and audited (ML24263A009) calculations supporting the submitted information. The NRC staff further notes that the conclusion section of the audited boron volatility supporting calculations document states (( }} is adequate for the XPC EM. The NRC staff reviewed the mechanisms that transport boron in the system and the treatment of the mechanisms as transport terms. The NRC staff finds that the transport terms have been appropriately implemented into the boron transport methodology. The NRC staff finds that these boron transport mechanisms have been adequality considered given the transport terms are developed to provide conservative results with respect to boron concentration in combination with the conservative treatment of mixing subject to L/C 4 and the required initial test program testing specified in L/C 7 to ensure boron transport is consistent with the analysis and XPC LTR methodology.

35 4.8.2.2 Boron Loss Mechanisms NuScale described the boron loss mechanisms in Section 6.2.4 of the XPC LTR. For boron precipitation analyses, (( }} as described in Section 7.2.4 of the XPC LTR. NuScale describes the following mechanisms that remove boron from the analyzed system: (1) (( }} (2) (( }} (3) (( }} (4) (( }} (5) (( }} The NRC staff reviewed the mechanisms that remove boron from the system and the treatment of the mechanisms as transport terms. The NRC staff finds that the transport terms have been appropriately implemented into the boron transport methodology. The NRC staff finds that these boron transport mechanisms have been adequality considered, given that the transport terms are developed to provide conservative results with respect to boron concentration in combination with the conservative treatment of mixing subject to L/C 4, and the required initial test program testing specified in L/C 7 to ensure boron transport is consistent with the analysis and XPC LTR methodology. 4.8.2.3 Boron Addition (ESB) Section 6.2.5 of the XPC LTR describes the boron addition term used to for the addition of boron from the ESB. Section 7.2.5 of the XPC LTR describes the simple method used to evaluate rapid dissolution rates from the ESB to add boron to the system.

36 NuScale states that (( }} NuScale states that (( }}. Therefore, NuScale stated that the porosity as a function of pellet diameter from the wall is based on experimental data measured by Zhang et al (Reference 6) by pouring equilateral cylindrical pellets into the larger cylinder without additional vibration or tapping. Figure 6-2 of the XPC LTR shows the porosity as a function of pellet diameters from the wall. The NRC staff reviewed the basis for the porosity as a function of the pellet diameter and find that the basis is reasonable when combined with dissolution testing and initial test program testing as specified in L/C 7. (( }} The NRC staff reviewed the approach used to calculate the dissolution rates with respect to solubility and finds the approach is reasonable because it is based on applicable literature with conservative assumptions, and dissolution is biased conservatively (fast or slow) in the analyses which can account for uncertainties. NuScale describes their dissolution rate calculation considering the boron oxide form. The dissolution rates are compared to the dissolution test described in Section 4.3 of the XPC LTR. The NRC staffs evaluation of the testing is described in Section 4.8.3 of this SER. The NRC staff determined that the ESB addition term was adequately determined given the conservative assumptions, comparison to dissolution testing, conservative biasing, and the initial first of a kind test. The initial test program t is used to confirm the performance of the dissolution rate for the NPM20, as specified in L/C 7. The NRC staff finds the biasing for the boron transport methodology to be acceptable because of the conservatism shown in the test assessment and the initial test program for dissolution. 4.8.3 Boron Dissolution Testing Assessment Section 4.3 of the XPC LTR describes the boron dissolution testing performed by NuScale. NuScale states that the testing was done to perform timed dissolution tests of pelletized boron oxide.

37 The test facility and the test matrix description are provided in Section 4.3.1 of the XPC LTR. The test facility description indicates that the testing is not prototypical of the NPM-20 design. However, there is a range of boron basket sizes used in the testing. Effectively, the test allows heated feedwater to be delivered to a dissolver basket at the specified rate. Table 4-8 of the XPC LTR provides the dissolution test matrix. The tests vary the target initial mass of boron oxide, target fluid temperature and target flow rate. The results of the tests are presented Table 4-9 of the XPC LTR. NuScale provided an assessment of the results from the dissolution tests and presented them in Section 4.3.3 of the XPC LTR. NuScale states the dissolution test data are used to assess the boron dissolution computational methods that are part of the boron transport methodology. The transport methodology uses either a fast dissolution bias or slow dissolution bias depending on the most limiting condition for the FOM of concern. The dissolution times from the test data are compared to the biased method used in the boron transport methodology. The biased calculation for dissolution in the boron transport methodology uses the test conditions and boron mass. Figure 4-61 of the XPC LTR shows a plot of the ratios from the calculated biased dissolution rates to the test data. The results of the assessment show that the computational biased method has faster dissolution than the test data for the fast bias and slower dissolution than the test data for the slow bias. The NRC staff reviewed the test facility description and test matrix. Additionally, the NRC staff reviewed the assessment of the results of the test and comparison to the boron transport methodology calculations for dissolution. The NRC staff finds that the tests provide some assurance that the boron transport methodology is performed adequately. However, the test facility is non-prototypical, and the tests arent correlated with NPM-20 conditions. Therefore, the initial test program startup tests required by L/C 7 are used to confirm the performance of the dissolution rate for the NPM-20. The NRC staff finds that the biasing for the boron transport methodology is acceptable because of the conservatism shown in the test assessment and the initial test program for dissolution testing per L/C 7. 4.8.4 Critical Boron Concentration Evaluation Section 6.3 of the XPC LTR describes the methodology for determining the critical boron concentration. The critical boron concentration is defined as the boron concentration needed for the core to remain subcritical at specified conditions (ex. moderator temperature coefficient (MTC), xenon, thermal hydraulic conditions etc.). NuScale states that end of cycle (EOC) conditions are the most limiting due to the large negative MTC. The critical boron concentration is calculated using CMS5 with the associated reactivity uncertainty (nuclear reliability factor) [OPEN ITEM, RAI XPC.LTR-21] over a range of moderator temperatures down to the lowest possible temperature during cooldown and is used to determine the most reactive core conditions. ((

38 }} NRC staff finds the treatment of axial offset acceptable. As discussed in XPC LTR Section 4.4.3.1, for analyses considering the operating history of reduced-power operation before an event, ((

}}. The response to RAI 10298 (XPC.LTR-6) and the calculations reviewed in the audit provide (( 

}}. [OPEN ITEM, RAI XPC.LTR-6] The lowest temperatures during cooldown are calculated conservatively given the conservative pool temperature calculations in the thermal hydraulic analyses. The NRC staff finds that the critical boron concentration methodology is adequate because the critical boron concentration uses an approved code (CMS5 code suite) and corresponding methodology, and because adequate conservatisms are used such as (( }}, nuclear reliability factor [OPEN ITEM, RAI XPC.LTR-21]and the most liming point in the cycle and most limiting conditions.

39 4.8.5 Criticality Margin Assessment Section 6.4 of the XPC LTR provides a high-level description for the determination of margin to criticality. NuScale states that boron concentration results inside the module that are determined from the boron transport calculations are compared to the appropriate SIMULATE5 calculated critical boron concentration results to demonstrate that the core boron concentration remains above the critical boron concentration for at least 72 hours after event initiation. The NRC staff finds the criticality margin assessment methodology to be acceptable because it utilizes the acceptable boron concentration from the boron transport analysis to compare to the critical boron concentration determined from the acceptable calculated critical boron concentration results. 4.8.6 Simplified Reactivity Control Method Evaluation for Extended DHRS In Section 6.5 of the XPC LTR, NuScale presented a simplified method to demonstrate that subcriticality is maintained during extended DHRS operation. The methodology describes (( }} (( }} (( }} (( }} The NRC staff reviewed the simplified methodology and finds that the method is adequate to evaluate subcriticality in the downcomer during extended DHRS cooling because the method is

40 conservative. Conservatisms considered are that the (( }}. Additionally, NuScale states that if the simplified approach is excessively conservative then the more detailed boron transport approach described in Sections 6.1 through 6.4 of the XPC LTR can be used with the biases described in Section 5.4.4 of the XPC LTR. The detailed boron transport approach includes analyses that transition from DHRS to ECCS cooling and consider BOC, MOC and EOC conditions. The DHRS portion of the long-term detailed analyses are biased to provide conservative results and therefore, the NRC staff finds it to be acceptable for BOC, MOC and EOC conditions in lieu of the simplified method given that all DHRS cooldown events and conditions are evaluated and confirmed to be bounded by the detailed long-term analyses. 4.8.7 Boron Precipitation Methodology Assessment Post-LOCA, and design basis events that actuate ECCS, long term cooling has the objective of maintaining the core at safe temperature levels during the long-term. To assure the core is maintained at acceptably low temperatures, precipitation of boric acid during the event must be avoided. The means of preventing boric acid precipitation for the NPM-20 design is to establish and maintain sufficient natural circulation flow into the RPV by removing heat through the ECCS, which provides cooling via the CNV. Heat removal through the SGs and DHRS provides LTC and heat removal that supports natural circulation with sufficient flow through the core, upper plenum, and lower riser holes to control the boric acid accumulating in the core, upper plenum, and riser so that it mixes with the fluid in the downcomer, keeping the boric acid concentration in the RPV well below the precipitation limit during the long term. Thus, the NPM-20 design demonstrates conformance to Acceptance Criterion 4 and 5 for Light Water-Cooled Reactors as presented in 10 CFR 50.46. The XPC EM methods applicable to precipitation evaluations for the NPM-20 design are discussed in Sections 7.0 and 7.2 of the XPC LTR. In Section 7.2, (( }} It is noted that the NRC staff independently identified means for developing successful LTC methods that can show precipitation is prevented in light water reactors which are applicable to the NPM-20 design. The NRC staff used elements from the independent evaluation in the review. Key precipitation analysis modeling features and initial conditions included in the independent evaluation of precipitation concerns include:

41 (1) (( }}. In consideration of the above assessment elements, the staff recognizes that ((

42 }}. The NRELAP5 code is used to compute the thermal and hydraulic behavior in the RPV and CNV to support the boric acid concentration behavior (( }} the NRC staff believes here for LTC predictions that the under predictive capability of the drift velocities at low pressure is considered conservative since this under prediction of drift velocity will cause the boric acid concentration to be higher at these lower drift velocities. While the drift velocity model can be improved through the use of better predictive drift velocity correlations (such as the Dix model which behaves quite well over the full range of void fractions and at low pressures), use of the NRELAP5 drift velocity models is considered acceptable for computing boric acid concentration (( }}. Lastly, (( }}. Also, the NRC staff further believes that boric acid plate out on the SG tubes as well as in the RRVs and lower riser holes is also small such that blockage of the RRVs and riser holes will not impact the recirculation flow through these valves and holes, which maintains good mixing of the boric acid throughout the RPV during the long term. As mentioned above, Section 7.0 of the XPC LTR provides the boron precipitation evaluation methodology and results. The approach for boron precipitation analyses is similar to the boron transport analysis to show subcriticality with assumptions (( }}. Section 7.2 of the XPC LTR for boron precipitation is similar to Section 6.2 of the XPC LTR with respect to the description of boron transport. However, the description explains the parts of the method that maximize boron for potential precipitation. (( }}. [OPEN ITEM, RAI XPC.LTR-6] NuScale provided a solubility limit from References 10.2.15 and 10.2.16 of the XPC LTR and presented the solubility data in Figure 7-2 of the XPC LTR. NuScale stated that the solubility limit in Figure 7-2 is used in the boron precipitation methodology and is a lower limit due to the low system pressure and chosen temperature. The NRC staff reviewed the referenced information and Figure 7-2. Pending closure of the above noted open item, the NRC staff finds that the boron solubility limit used in the

43 methodology is adequate given that the solubility curve reasonably reflects the data and conservative nature of the treatment of the boron transport terms. The NRC staff finds that the margin to the solubility curve is adequately captured given the conservative transport terms used. A discussion of the boric acid behavior for two events is given below in the next section. 4.8.8 Representative Results Sections 6.6 and 7.5 of the XPC LTR provide representative boron transport and concentration results using the XPC methodology. The representative calculations provide the following: (1) critical boron concentration, (2) transient boron concentrations, and (3) transient boron mass. Based on its review of NuScales representative calculations, the NRC staff concludes that the representative analyses appropriately illustrate that implementation of the methodology as specified in the XPC LTR provides adequate and expected results, demonstrating that the methodology, when implemented, will provide reasonable results appropriate for determining whether FOMs are met, and precipitation following all LOCAs is prevented. It is specifically noted that Fig. 7-3 of the XPC LTR presents a boric acid transient event that shows that the maximum core concentration was found to be 5,000 ppm, compared to the precipitation limit of about 10,500 ppm. Furthermore, this result is also noted in calculations the NRC staff audited for the NPM-20 boron transport analysis, which showed significant margin relative to the limit. Also shown in the audited calculations were the results of another event that show a core concentration of about 10,000 ppm compared to the precipitation limit of about 16,000 ppm. It should be mentioned that the NPM-20 design has a maximum boron concentration that is fixed by the maximum initial concentration of 1,900 ppm (1.09%, 1,090 lbs), plus the CNV boric acid of about 186 lbs. It is noted that the boron mass for this event, after about 62 hrs, (( }} it has little meaning since this amount of boron will have no effect on the maximum concentration in the core. That is, there is a large margin for this event to assure precipitation will not occur for this limiting concentration event. Given these results, and the fact that NRELAP5 predicts sufficient natural circulation flows within the RPV, the core maximum boric acid concentration will remain well below the precipitation limit for the representative analysis. 5.0 LIMITATIONS AND CONDITIONS This section provides a summary of the L/Cs based on the technical evaluation of the NuScale Topical Report (TR) TR-124587-P, Extended Passive Cooling and Reactivity Control Methodology, Revision X. As a result of its in-depth technical evaluation, NRC staff determined that the NuScale XPC EM, can be used for the NuScale NPM-20 design, subject to the limitations and specific restrictions on the use of this model as listed below.

1. Any future changes or revisions to TR-0516-494-P-A, Loss-of-Coolant Accident Evaluation Model, (Reference 1) or TR-0516-49416-P-A, Non-Loss-of-Coolant Accident Evaluation Model, Reference 2), must be assessed by the applicant for their potential impact on the XPC EM. Any subsequent changes to the XPC methodology require NRC approval.

44

2. Use of the XPC EM is limited to evaluations of the US460/NPM-20 design. An applicant or licensee seeking approval to use the XPC EM for a design other than the US460/NPM-20, such as the US600/NPM-160, or another future NPM design, is required to demonstrate the applicability of the XPC EM to the specific NPM design. The use of this methodology for a specific NPM design other than the US460/NPM-20 requires NRC staff review and approval of the applicants or licensees determination of applicability.

3. Use of the XPC EM for a design requires an applicant or licensee seeking to apply this methodology to ensure that either (1) the CNV will be maintained at a vacuum with insignificant initial non-condensable gas and safety related means to recombine or remove the gases released from the RPV after ECCS is initiated or that (2) a calculation is presented in the FSAR that demonstrates that the amount of non-condensable gases present do not impact the margin to re-criticality.

4. Use of the XPC EM for a design requires an applicant or licensee seeking to apply this methodology to adequately consider the density difference between the borated and unborated liquid in the CNV with respect to mixing by ensuring that either (1) the boron dilution transport methodology results show at least 25 ppm margin between the core region boron concentration and the critical boron concentration and would be equivalent to a delay of at least 6 hours in mixing between the lower and upper CNV, or that (2) an analysis is performed that demonstrates that the methodology is conservative with respect to delayed onset of mixing in the CNV volumes due to explicit consideration of density differences between the borated and unborated liquid in the volumes.

5. Use of the XPC EM is approved for use for LOCA and non-LOCA design basis events through the event progression up to 72 hours. An applicant must address post-accident recovery actions in the FSAR.

6. Use of the XPC EM for a design requires the xc (or xt) value for RVV compressible flow to be verified as part of the ASME QME-1 qualification program which is consistent with the values applied to the XPC analyses on a transient basis.

7. An applicant or licensee seeking to apply this methodology to a design must include an initial test (first module only) for dissolution testing during the initial test program. The test must be consistent with the analysis methods and code of record based on the predictions of the as-tested conditions. The test must be designed to demonstrate acceptable performance of the as-built ESB based on predicted system response under expected test conditions to ensure the system as a whole meets the fundamental design requirements of the safety analysis. The test acceptance criteria must confirm the as-built functionality of the ESB, including boron basket dissolution rates, condensate rail collection capability, mixing tube flow, lower containment boron concentration and mixing.

8. Unless changes are made pursuant to a change process specifically approved by the NRC staff for changes to NRELAP5 and the NPM model, use of NRELAP5 is limited to Version 1.7 in conjunction with NPM-20 basemodel Revision 5 (or later NRELAP5 versions and/or new NPM-20 basemodel revisions if the revisions are demonstrated to produce either essentially the same or conservative results and are consistent with the approved methodology, or if the revision is to the basemodel and due to a change made to SSC physical/process input parameters only made via established change control processes (such as 10 CFR 50.59)).

9. An applicant or licensee seeking to apply this methodology to a design must describe in its submittal the following analytical assumptions considered for the evaluation of design basis events described in this TR and receive a separate approval for those assumptions: 1) single failures, 2) electrical power assumptions (AC/DC), or 3) operator actions relied on in

45 the analysis (and therefore necessary to mitigate design basis events) within the 72 hours following event initiation to improve the results relative to the applicable figures of merit for a particular set of initial conditions, including actions taken to prevent accidents and transients from progressing to more severe events.

10. An applicant or licensee seeking to apply the XPC EM to a design for analysis of ((

}}.

11. An applicant or licensee seeking to apply the XPC evaluation model to a design for analysis of ((

}}.

6.0 CONCLUSION

S This SER documents the results of the technical evaluation of Topical Report (TR) TR-124587-P, Extended Passive Cooling and Reactivity Control Methodology, Revision 0, which is an extension of the NuScale LOCA evaluation model (EM) (Reference 1) and NuScale non-LOCA EM (Reference 2) to perform post-LOCA and non-LOCA LTC and reactivity control analyses of the US460 plant with the NPM-20 module design. Subject to the closure of the open items noted in the above evaluation, the NRC staff finds that the proposed methodology is acceptable for meeting the requirements of 10 CFR 50.46(b)(4) and (b)(5) and Appendix K, GDC 26, GDC 27, GDC 34, and GDC 35, evaluated in this SER, for evaluation of LOCA and non-LOCA events for (1) the emergency core cooling system (ECCS) and decay heat removal system (DHRS) XPC of the NuScale Power Module (NPM-20) after a successful initial short-term response to a design basis event; (2) reactivity control during XPC of the NPM-20; (3) margin to boron solubility limit to demonstrate coolable geometry is maintained in the NPM-20; and (4) boron precipitation concerns; subject to the limitations, conditions, and restrictions identified in Section 5.0 above. Pending closure of the open items noted in this SER, the NRC staff finds the NuScale XPC EM appropriate for determining that (1) the core remains subcritical (2) boron concentration remains below the solubility limit and (3) the collapsed liquid level in the RPV remains above the core; that uses this version of the NuScale XPC EM.

46 7.0 REFERENCES

1. NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422-P-A, Revision 3, July 31, 2023, ADAMS Accession No. ML23008A001.

2. NuScale Power, LLC, Non-Loss-of-Coolant Accident Evaluation Model, TR-0516-49416-P-A, Revision 4, July 31, 2023, ADAMS Accession No. ML23005A304.

3. NuScale Power, LLC, Long-Term Cooling Methodology, TR-0916-51299-P, Revision 3, May 2020, ADAMS Accession No. ML20141L817 (Proprietary Version)

4. NuScale Power, LLC, Submittal of Revision 1 to Standard Design Approval Application, ADAMS Package Accession No. ML23306A033.

5. N.E. Kukuljan, J., J. Alvarez, and R. Fernandez-Prini, "Distribution of B(OH)3 between water and steam at high temperatures, Journal of Chemical Thermodynamics, (1999):

31: 1511-1521.

6. Zhang, W., et al., "Relationship between packing structure and porosity in fixed beds of equilateral cylindrical particles," Chemical Engineering Science, (2006): 61:8060-8074.}}