ML25155B867
| ML25155B867 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 06/04/2025 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML25155B867 (51) | |
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From:
Thomas Hayden Sent:
Wednesday, June 4, 2025 12:59 PM To:
NuScale-SDA-720DocsPEm Resource
Subject:
FW: Final Safety Evaluation for NuScale Extended Passive Cooling and Reactivity Control Methodology Prop and non-Prop Attachments:
Final Safety Evaluation of the NuScale Power LLC Topical Report TR-124587 Revision 1 Extended Passive Cooling and Reactivity Control Methodology - 042325.docx Tommy Hayden Project Manager Email: thomas.hayden@nrc.gov Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission From: Thomas Hayden Sent: Wednesday, April 23, 2025 11:43 AM To: regulatoryaffairs@nuscalepower.com Cc: Griffith, Thomas <tgriffith@nuscalepower.com>; Bode, Amanda <abode@nuscalepower.com>;
Turmero, Sarah <sturmero@nuscalepower.com>; Mahmoud -MJ-Jardaneh
<Mahmoud.Jardaneh@nrc.gov>; Getachew Tesfaye <Getachew.Tesfaye@nrc.gov>
Subject:
Final Safety Evaluation for NuScale Extended Passive Cooling and Reactivity Control Methodology Prop and non-Prop By letter dated March 28, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML25087A229 (Proprietary) and ML25087A228 (non-Proprietary)),
NuScale Power, LLC (NuScale), submitted Topical Report (TR) TR-124587, Revision 1, Extended Passive Cooling and Reactivity Control Methodology to the U.S. Nuclear Regulatory Commission (NRC). The NRC staff has prepared a final safety evaluation for TR-124587, Revision
- 1. The non-proprietary (ML25098A230) and proprietary (ML25098A229) final safety evaluations are enclosed. The NRC staff has found that TR-124587, Revision 1, is acceptable for referencing in licensing applications for the NuScale small modular reactor design to the extent specified and under the conditions and limitations delineated in the enclosed final safety evaluation.
The NRC staff requests that NuScale publish the accepted version of this TR as soon as possible following receipt of this electronic mail. The accepted version shall incorporate this electronic mail and the enclosed final safety evaluation after the title page. It must be well indexed such that information is readily located. Also, it must contain historical review information, including NRC requests for additional information and accepted responses. The accepted version of the TR shall include a -A (designated accepted) following the report identification number.
If the NRCs criteria or regulations change such that the NRC staffs conclusion in this electronic mail (that the TR is acceptable) is invalidated, NuScale and/or the applicant referencing the TR will be expected either to revise and resubmit its respective documentation or to submit justification for continued applicability of the TR without revision of the respective documentation.
If you have any questions or comments concerning this matter, I can be reached at (301) 415-2956 or via e-mail address at Thomas.Hayden@nrc.gov. The attached documents are both password protected. Password to follow in a separate email.
Docket Nos. 99902078, 05200050 Sincerely, Tommy Hayden Project Manager Email: thomas.hayden@nrc.gov Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
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SAFETY EVALUATION BY THE U.S. NUCLEAR REGULATORY COMMISSION TOPICAL REPORT TR-124587, REVISION 1, MARCH 28, 2025 EXTENDED PASSIVE COOLING AND REACTIVITY CONTROL METHODOLOGY NUSCALE POWER, LLC
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............................................................................................... 4 2.3 NUREG-0800 Standard Review Plan........................................................................... 6 3.0 NUSCALE XPC EVALUATION METHODOLOGY
SUMMARY
........................................ 6
4.0 TECHNICAL EVALUATION
............................................................................................. 8 4.1 Introduction and Scope................................................................................................ 8 4.2 Background and Acceptance Criteria........................................................................... 8 4.3 EMDAP Process.......................................................................................................... 9 4.4 Phenomena Identification and Ranking........................................................................ 9 4.5 Evaluation Model Overview and Computational Tools................................................12 4.6 NRELAP5 Computer Code and Assessment Basis.....................................................14 4.7 Extended Passive Cooling Thermal Hydraulic Analysis Methodology Evaluation........21 4.8 Evaluation for Reactivity Control and Boron Distribution.............................................30 5.0 LIMITATIONS AND CONDITIONS.................................................................................42
6.0 CONCLUSION
S.............................................................................................................44
7.0 REFERENCES
...............................................................................................................45
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. On March 28, 2025, (ML25087A228), NuScale submitted Revision 1 to TR-124587-P (hereafter referred to as the XPC TR).
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 1) 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 2) and NuScale non-LOCA EM (Reference 3) 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 (SE).
This SE 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, subcriticality 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 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 1). Some of the design features for the NPM-20 are described in the XPC 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.
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.
- 3. Prevent precipitation of the boric acid in the reactor coolant system.
2
- 4. Maintain sufficient boron in the RCS to prevent re-criticality with the highest worth control rod stuck out of the reactor core.
Pursuant to NRC regulations, the systems used to fulfill extended passive cooling and reactivity control functions must function for a period of at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> until transition to the recovery phase. The XPC topical report provides the analysis methodology for the 72-hour period after event initiation. The analyses in the XPC topical report do not include time frames after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and post-event recovery actions must be addressed separately. This SE 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 SE describes the regulatory basis and supporting guidance documents that the NRC staff used to determine whether the methodology described in the XPC EM was acceptable.
2.1 Regulatory Requirements The relevant LOCA requirements for this area of review and the associated acceptance criteria, given in Design-Specific Review Standard for NuScale SMR Design, (DSRS) Section 15.6.5 Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, Revision 0, (ADAMS Accession No. ML15355A309),
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, sections (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, Combined Reactivity Control Systems Capability, for postulated accidents specify that a stuck control rod is to be assumed for both AOOs and postulated accidents. GDC 34, Residual Heat Removal, is also relevant because it specifies residual heat removal requirements for AOOs.
2.1.1 10 CFR 50.46 ECCS and Appendix K to 10 CFR Part 50 Requirements The regulations at 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. NuScale states that the XPC EM is intended to conform to the required and acceptable features of Appendix K to 10 CFR 50.
3 10 CFR 50.46 The regulations at 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 LOCA.
The regulations in 10 CFR 50.46(a)(1)(i) also require that ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and provides for two alternative options 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, NuScale states that the 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 2), 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). The criteria for calculated ECCS cooling performance during postulated LOCAs is defined in 10 CFR 50.46(b)(1) through 10 CFR 50.46(b)(5). The applicable portions for the XPC TR are as follows:
(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 1) for the US460 (which incorporates the NPM-20 module) includes PDC that were developed based on the GDC. In the
4 XPC TR, NuScale provided the PDC for GDC 34, Residual Heat Removal, and GDC 35, Emergency Core Cooling. PDC 34 and 35 proposed by the applicant are functionally identical to GDC 34 and 35 for this evaluation model. NuScale did not provide PDCs for GDC 26 and 27, which are applicable to water-cooled nuclear power reactors pursuant to 10 CFR Appendix A.
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 1):
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 1). 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.
GDC 26, Reactivity Control System Redundancy and Capability, 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.
NuScale states that 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
5 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 SE provides the NRC staffs review of this PIRT.
2.2.1 Evaluation Model Concept Consistent 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):
(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
6 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 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.
(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.
DSRS 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
7 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:
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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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
8 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 SE, 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 TR are discussed below and then summarized in Section 5 of this SE. The conclusions from the review are discussed below and then summarized in Section 6 of this SE.
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 TR 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 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 TR, with respect to the US460 and NPM-20, and on the determination whether the proposed methodology meets the applicable regulations.
NuScale states that their XPC EM is an extension of the NuScale LOCA EM (Reference 2) and NuScale non-LOCA EM (Reference 3) and is developed using a graded approach to the EMDAP defined in RG 1.203.
The XPC TR provides a description of the methodology used by NuScale for XPC and long-term reactivity control analyses, and this methodology is reviewed in this SE for compliance with applicable regulatory criteria. However, the XPC TR does not provide any final licensing analyses, and this review of the XPC TR 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 TR provides a description of how the NuScale XPC EM conforms to the EMDAP of RG 1.203. XPC TR Table 2-1 states that TR Section 8.0 describes NuScale
9 procedures implementing the QA program that governs requirements for documentation and verification of the EM. Section 4.5.3 of this SE discusses the QA requirements and evaluation.
4.3 EMDAP Process Section 2.4 of the XPC TR 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 TR provides a general description of the NPM-20 design and key design features. The XPC TR is an extension of the LOCA and non-LOCA TRs, which also follow the EMDAP process.
4.4 Phenomena Identification and Ranking As discussed in Section 3.1 of the XPC TR, 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 4). 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.
NuScale documented phenomena and processes of high and medium ranked importance in Table 3-4 in Section 3.4 of the XPC TR and did not document low ranked phenomena or processes in Table 3-4. Typically, the low ranked phenomena and processes are included in EM TRs. 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 TR 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 TR evaluates these acceptance criteria for the extended ECCS and extended DHRS phases after design basis events. NuScale separates design-basis events into phases. Short-term phases are addressed separately by NuScale in the applicable LOCA and non-LOCA TRs, while the long-term 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:
10
- 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.
- 4. ((
- 5. ((
}} 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. NRC staff evaluated the XPC phases selected for the PIRT discussion sand find that they are consistent with RG 1.203. The NRC staff determined that the NuScale XPC phases selected as the basis for their PIRT process are therefore 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 TR. 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 1). 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 to apply the XPC EM in Table 3-3 of the XPC TR. 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 are 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, to form the basis for the NRC staffs approval of the XPC TR, because the PIRT was constructed and updated based on the design of the US460 and NPM-
- 20. Applicability evaluation(s) would be needed to apply the method to a design that is not the NPM-20 (Reference 1). L&C #2 has been developed to ensure that the applicability is only for the NPM-20 unless an applicability assessment is made and submitted to the NRC for review and approval.
11 4.4.2 Figures of Merit In Section 3.3 of the XPC TR, 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 following NuScale XPC EM FOMs are acceptable: (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. The FOMs are acceptable because they show conservatism with respect to applicable long term cooling acceptance criteria or show that the acceptance criteria are met. The NRC notes that the FOMs are acceptable for the NuScale XPC EM and the LOCA (Reference 2) and non-LOCA (Reference 3) TRs provide the methodology for showing their respective short term acceptance criteria are met. 4.4.3 PIRT Rankings Section 3.4 of the XPC TR 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 TR. 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 TR, 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 TR. 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. 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 consistent with Reg Guide 1.203. 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 acceptable 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 TR, which document a summary of the NuScale assessment of the high and medium phenomena from the PIRT identified in
12 Section 3.4 of the XPC TR. 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 TR 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 TR 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 TR. The NRC staff found agreement generally 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 considering other compensating 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 SE, 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 TR 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 TR 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 and consistent with Reg Guide 1.203. 4.5 Evaluation Model Overview and Computational Tools NRC staff reviewed Section 4.1 of the XPC TR 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 TR, NuScale correlates the EM FOMs of collapsed liquid level, subcriticality, and coolable geometry (Section 3.3 of the XPC TR) with the EM acceptance criteria summarized below (Section 2.3.2 of the XPC TR), respectively: Collapsed liquid level remains above the top of active fuel, demonstrating adequate decay heat removal for at least 72 hours after event initiation
13 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. 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 because they are consistent with the Reg Guide 1.203 evaluation model development process. 4.5.2 Computational Tools Section 4.1.2 of the XPC TR 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 TR (Reference 2). The NRC staff notes that further details about NRELAP5 and how it is used for non-LOCA events is described in the non-LOCA TR (Reference 3). NRC staff confirmed that the NRELAP code is adequately described in the LOCA TR 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 TR 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 SEs for the LOCA and non-LOCA TRs (References 2 and 3, respectively). Considering both the code and NPM design changes, the staff finds that NRELAP5 Version 1.7 is generally applicable for use in the XPC EM because it is consistent with the LOCA and non-LOCA TRs and the validation basis for long term cooling is provided, subject to L&C #8 of this SE.
14 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 TR, 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 TR. 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 TR is adequate for use in performing XPC EM calculations because the script adequately applies the boron transport calculational framework. (ML24346A354). 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 2. The NRC staff reviewed the QA requirements and documented its approval in its SER of the topical report (ML25031A372). 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 (ML25056A127) 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 XPC EM. 4.6 NRELAP5 Computer Code and Assessment Basis The Section 4.2 of the XPC TR states that as part of LOCA EM (Reference 2) development and non-LOCA EM (Reference 3) development, 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 TR states that tests that were performed to support the LOCA TR 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
15 during the ECCS LTC phase and integral effects tests were performed at the NIST-2 facility for extended ECCS cooling. 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 SE, the NRELAP5 computer code is described in the LOCA and non-LOCA TRs. The constitutive models used in the XPC TR are the same as those described in the LOCA and non-LOCA TRs. The assessment bases for those models are approved as part of those TR 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. 4.6.1 NRELAP5 Assessments Section 4.2 of the XPC TR 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 SE: 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 TR states that during extended ECCS operation, with the SG and DHRS in operation, NRELAP5 sensitivity analyses show that (( }} 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 TR safety evaluation (Reference 5), the staff noted that the NIST NRELAP5 model was updated ((
16
}}
(( }}. Based on the results of the various assessment sensitivity studies, the staff confirmed the applicants conclusion that the NRELAP5 model responses are consistent with physics-based expected results and that there are very negligible effects on the event FOMs as described in the LOCA TR safety evaluation (Reference 5). 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 TR. 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. Because of this, the NRC staff finds
17 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 TR and notes that the test suite covers a LOCA scenario with ECCS actuation with variations (( }}. Because of this, the NRC staff finds that the suite of tests, in combination with the validation activities from the LOCA TR, 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 TR 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 TR 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 TR. The table compared the sequence timing between the test data and the simulation. The NRC staff finds that the comparisons are adequate because 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 TR, 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 TR 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 TR show (( }} pressures and RPV level as seen in the extended ECCS runs presented in Section 4.2.2 of the XPC TR. The NRC staff reviewed the sensitivity studies performed in Section 4.2.3 of the XPC TR 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 (( }}.
18 The CPV level and temperature comparisons show that NRELAP5 (( }} parameters. NuScale states that the mechanism for energy removal in the CPV is heat transfer (( }} 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 TR 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 entirely 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 because the model provides a reasonable representation of the phenomenological response.
19 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 sufficient to benchmark the NRELAP5 computer code and justify its use for XPC analyses for extended ECCS cooling because the suite of tests provides an adequate range of LOCA scenarios with 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 notes 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.
20 4.6.5 NuScale NIST-2 non-LOCA SG/DHRS Test Assessment In Section 4.2.5 of the XPC TR, NuScale references the non-LOCA TR (Reference 3) for NRELAP5s ability to simulate total DHRS heat removal (ML24305A290). The non-LOCA NIST-2 tests are described in non-LOCA TR 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 5-226 shows that RPV pressure is (( }} for the long term and SG steam drum pressure is (( }}. Non-LOCA Figure 5-227 shows that the SG level and DHRS level are (( }}. Non-LOCA Figure 5-228 shows that steam drum level is (( }} by NRELAP5. Non-LOCA Figure 5-229 shows that the DHRS inlet and outlet header predictions are (( }}. Non-LOCA Figures 5-234, 5-236 and 5-237 show that the RPV level, pool level and pool temperature is (( }}. As noted in Sections 4.6.3 and 4.6.4 of this SE, 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 (( }}. 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 (ML25008A173) 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 TR adequately addresses the boundary conditions used for the pool. The NRC staff also notes that modeling in the cooling pool cannot capture all of the realistic 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.
21 4.7 Extended Passive Cooling Thermal Hydraulic Analysis Methodology Evaluation Section 5.0 of the XPC TR and related subsections describe the NuScale NPM-20 XPC thermal hydraulic methodology. The XPC TR analysis methodology credits the ECCS and DHRS for long term decay heat removal. The XPC TR 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 TR 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 TR. 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 events 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. 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
22 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 lessen 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, the recovery actions from the post 72-hour conditions are not known and are not covered by the TR. 1 The Low-low riser level signal is 460-472 from bottom of the pool (38.3 to 39.3 ft) -upper riser hole elevations (( }}.
23 4.7.2 XPC NRELAP Model As discussed in Section 5.2 of the XPC TR and related subsections, NuScale developed the NRELAP5 XPC input model from the detailed NRELAP5 basemodel developed for short term LOCA evaluation model calculations (Reference 2). 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 TR. Additionally, some of the key XPC model differences are ((
}}
(( }}. The reactor pool model (( }}. As indicated in Section 4.6 of this SE, the validation testing at the NIST-2 facility for the NPM-20 indicates that reasonable boundary conditions for the pool are needed to get an adequate comparison between the data and NRELAP5 model. NuScale described (( }} is acceptable because such modeling 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 identified as a required design value to be verified as part of the ASME QME-1 qualification program to be consistent with the analyses. The NRC staff confirmed that the (( }}.
24 Since validation of the NRELAP5 code for XPC depends heavily on the LOCA TR 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 TR). 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 TR. 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 TR 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 TR 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 TR and evaluated in Section 4.8 of this SE. 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 which would indicate that there is overall more mixing in the system ex. (enhance flow from the CNV into the RPV). NuScale states (( }}. ((
}}
25 (( }}. (( }}. 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 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
26 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 increases void fraction which maximizes boron content in the core for precipitation and reduces potential for return to power for dilution. Higher riser void produces 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 when 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., as it pertains to riser hole flow calculation, overestimating decay heat tends to increase the predicted mixture level), when taken as a whole, the NRC staff noted that the set 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.
27 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 notes 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 TR 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 long term, the NRC staff finds that the statepoint calculations are adequate to represent the long term results because the initial calculations are run for 12 hours where most relevant phenomena is captured and run to 24 hours for minimum collapsed liquid level and the results from the statepoint method align reasonably well with the initial calculations. The XPC TR 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.
28 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, 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. 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.
29 4.7.4 Initial Conditions and Biases Section 5.4 of the XPC TR 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 TR 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, (ML24305A290) 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 TR. Table 5-5 of the XPC TR states that non-condensable gas is biased low for boron transport analyses. As discussed in Section 4.4.3 of this SE, 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 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 TR 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 TR, when implemented, will provide reasonable results appropriate for determining whether FOMs are met.
30 4.8 Evaluation for Reactivity Control and Boron Distribution The NuScale XPC thermal hydraulic analysis methodology is described in Section 5 of the XPC TR. 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 TR. This section of the SE discusses the evaluations associated with reactivity control, boron transport and boron distribution. Section 6 of the XPC TR 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 TR 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 TR describes the general approach and acceptance criteria used to show that the RCS boron concentration remains below 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 precipitation does not occur. The NRC staff finds that the acceptance criteria are 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-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 TR 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 TR and is similar to the boron transport
31 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 TR, 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 TR, 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 (( }}. A break in the discharge line creates a boron transport pathway between (( }}. (( }}. (( }}. ((
32 }}. The XPC TR 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 (( }}. 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 (( }}. 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 (( }} 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 (( }} in mixing between the lower CNV and upper CNV. The NRC staff confirmed that a (( }} 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 either (1) a 25 ppm margin between the core boron concentration and critical boron concentration, and analysis which demonstrates that limiting events remain subcritical when assuming a (( }} in mixing between the (( }}, or (2) L&C analysis that explicitly considers the density difference demonstrates that the methodology is conservative with respect to delayed mixing in the CNV (ML24305A290).. 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 perform an initial test for the first module only for dissolution testing during the initial test program or reference testing performed by another licensee. The test must be consistent, to the extent practical, 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 (ML24305A290). 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
33 to ensure the system as a whole meets the fundamental design requirements of the safety analysis. The test acceptance criteria must confirm, either implicitly or explicitly, 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 6) 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 adequately considered because 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 TR methodology. 4.8.2.2 Boron Loss Mechanisms NuScale described the boron loss mechanisms in Section 6.2.4 of the XPC TR. For boron precipitation analyses, (( }} as described in Section 7.2.4 of the XPC TR. NuScale describes the following mechanisms that remove boron from the analyzed system: (1) (( }}. (2) ((
}}
(( }}.
34 (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 adequately considered because 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 TR methodology. 4.8.2.3 Boron Addition (ESB) Section 6.2.5 of the XPC TR describes the boron addition term used to for the addition of boron from the ESB. Section 7.2.5 of the XPC TR describes the simple method used to evaluate rapid dissolution rates from the ESB to add boron to the system. 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 7) by pouring equilateral cylindrical pellets into the larger cylinder without additional vibration or tapping. Figure 6-2 of the XPC TR 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 because the porosity is based on experimental data and dissolution of the pellets is confirmed through the initial test program testing.
35 (( }}. 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 TR. The NRC staffs evaluation of the testing is described in Section 4.8.3 of this SE. The NRC staff finds 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 is used to confirm the performance of the dissolution rate for the NPM-20, 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 TR describes the boron dissolution testing performed by NuScale. NuScale states that the testing was done to perform timed dissolution tests of pelletized boron oxide. The test facility and the test matrix description are provided in Section 4.3.1 of the XPC TR. 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 TR 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 TR. NuScale provided an assessment of the results from the dissolution tests and presented them in Section 4.3.3 of the XPC TR. 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 TR 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.
36 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 TR describes the methodology for determining the critical boron concentration. The critical boron concentration is defined as the minimum 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) over a range of moderator temperatures down to the lowest possible temperature during cooldown and is used to determine the most reactive core conditions. ((
}}
(( }}, NRC staff finds the treatment of axial offset acceptable. Section 6.3 of the XPC topical report states that an NRF is applied to the critical boron concentration to account for SIMULATE5 code uncertainty. It states that an (( }}. The NRC staff issued RAI-10298 R1, XPC.TR-21 (MLML25083A094) to review the applicability of the NRF. The RAI response shows that the NRF is based on (( }}, as described in TR-0616-48793-P-A.
37 Additionally, TR-0616-48793-P-A describes a process to verify and update, as necessary, NRF values based on NPM startup test data. Per NuScales responses related to TR-0616-48793-P-A (i.e. NRC Question No. 29752 of eRAI No. 8807), this process provides assurance that unexpected uncertainties associated with reactor physics parameters will not result in operation of the NuScale Power Module outside the bounds of the safety analysis. Although these surveillance tests are not performed at the range of conditions representative of XPC transients, future upward adjustments to NRFs may provide indication that additional scrutiny of the XPC NRF is warranted, particularly if the adjustment is made to CBC, DBW, ITC/MTC, or CRA bank worth NRFs which relate to the parameter of interest and would impact the XPC-specific NRF. In review of the NRF, NRC staff also considered other aspects of the XPC TR that provide some measure of conservatism, including (( }} and consideration of off-nominal operating histories prior to an XPC transient. NRC staff finds that use of the XPC specific NRF, in combination with other aspects of the methodology described above which do contribute to conservatism and with the TR-0616-48793-P-A NRF update methodology, to be acceptable because the NRF was (( }}. As discussed in XPC TR Section 4.4.3.1, for analyses considering the operating history of reduced-power operation before an event, ORIGEN calculations for decay heat are determined considering the power level during reduced-power operation, including time at 0% power, the duration of reduced-power operation, the duration of power ascent after reduced power-operation and the final power level prior to reactor shutdown. The applicants calculated decay heat curves were reduced (( }} in ORIGEN decay heat analyses. XPC TR Table 5-6 states that these calculated decay heat curves are used as inputs to the XPC subcriticality analyses. Additionally, Section 6.3 of the XPC TR states (( }}. The response to RAI 10298 (ML25051A167) and the calculations reviewed in the audit provide boron transport subcriticality analyses that consider pre-transient conditions for (1) plant shutdown and reduced-power operation, (2) durations of reduced power operation (3) durations of power ascent after reduced-power operation and (4) power levels prior to the event. L&C #10 has been developed to ensure that operational histories are accounted for, including reduced-power operations and rapid power ascensions and the impacts on short term xenon changes and the potential for low decay heat. L&C #10 requires technical specification controls to establish operational limits for reactor coolant system boron concentration and the corresponding power ascension rate to account for operational histories that can result in low decay heat and short-term changes in xenon concentration. The operational lower limit for the reactor coolant system boron concentration as a function of power history, (e.g. integral reactor down power in units of cumulative megawatt hours below full power), must provide the initial core boron concentration for evaluation of the boron transport subcriticality events. Evaluation of the boron transport subcriticality events must include as input the full matrix of decay heat calculation cases identified in Table 1, Table 2 and Table 3 (ML25051A167). The power ascension rate limit must be consistent with the limiting ascension rate assumed in the operational minimum boron concentration limit analysis. 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
38 methodology, and because the method applies an adequate nuclear reliability factor based on XPC conditions, and has adequate conservatisms such as (( }}, and the most limiting point in the cycle and most limiting conditions. 4.8.5 Criticality Margin Assessment Section 6.4 of the XPC TR 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 TR, 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
39 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 TR can be used with the biases described in Section 5.4.4 of the XPC TR. 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 TR. In Section 7.2, (( }}. 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: (1) Using Appendix K power assumptions,
40 (2) Using Appendix K decay heat assumptions, (3) Assuming the worst single failure and assumptions to maximize void fraction in the core, such as axial power shape, (4) Justifying the drift flux model used in the core and upper plenum riser regions as applicable to the thermal hydraulic conditions. The staff notes that there is noteworthy uncertainty in the NRELAP5 drift flux model below atmospheric pressure and its ability to accurately capture the thermal hydraulic conditions and phenomena down to very low pressures in the NPM-20 design. (5) Justifying the mixing volumes considered. The staff notes that the (( }}. (6) Using conditions and models for predicting the minimum RCS pressures which could be below atmospheric pressure should be employed, (7) Demonstrating that sufficient recirculating flows and downward flow in the bypass can be achieved, (8) Considering plate out, (9) Accounting for possible dilutions sources, and (10) Assuming: No credit for increase in boric acid solubility due to other solutes, no credit for elevated boiling point due to boric acid concentration, solubility of boric acid in steam is zero, boric acid dilution sources neglected, and core inlet flow subcooling should be maximized. In consideration of the above assessment elements, the staff recognizes that noteworthy uncertainties related to some of these elements, as well as other elements in the XPC evaluation model, exists. However, other factors exist in the evaluation model that can compensate in part for some of these uncertainties. The boron precipitation analysis methodology assumes that the boron concentration in the (( }} is fully mixed and uniformly distributed during extended passive cooling. The staff notes that including the (( }} adds uncertainty and typically should be treated separately. For current plants with boron distribution models based on test data, the boron concentration in the core has a gradient, with the top half of the core having a higher concentration than the lower half. Several conservatisms exist in the NuScale model which could address or mitigate the gradient and mixing assumptions, which are: (1) NuScale uses the CNV saturation temperature for comparison to the solubility temperature limit, which adds conservatism relative to using the moderator temperature in the RPV for comparison to the solubility limit; (2) The pool (( }} which causes lower CNV and moderator temperatures; (3) A review of the NRELAP5 response and staff confirmatory analyses, through sensitivity analyses and audit, indicates a fair amount of mixing.; (4) Calculations reviewed via audit show that the maximum allowable Mode 1 boron concentration operational limit (including zero power, no xenon conditions) of 1900 ppm is used to set the initial boron concentration in the RCS. Conservatively assuming pre-accident boron concentration at this limit (including the assumptions for solubility temperature limit comparison, pool temperature modeling and RCS mixing sensitivities) provides some potential that the uncertainties are minimized and compensated for through the NuScale model assumption (( }} boron concentration. Therefore, the staff developed L&C #11 to ensure that the uncertainties are minimized and compensated for in the evaluation model for the mixing and transport assumptions for boron precipitation analyses by assuming pre-accident boron concentration at the maximum allowable Mode 1 operational limit as an initial condition in the RCS.
41 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 notes 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 notes that boric acid plate out on the SG tubes as well as in the RRVs and lower riser holes is also small so 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 TR 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 TR for boron precipitation is similar to Section 6.2 of the XPC TR with respect to the description of boron transport. However, the description explains the parts of the method that maximize boron for potential precipitation. (( }}. NuScale provided a solubility limit from References 10.2.15 and 10.2.16 of the XPC TR and presented the solubility data in Figure 7-2 of the XPC TR. NuScale stated that the solubility limit in Figure 7-2 is used in the boron precipitation methodology and isa lower limit due to the low system pressure and chosen temperature. The NRC staff reviewed the referenced information and Figure 7-2. The NRC staff finds that the boron solubility limit used in the methodology is adequate because 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 because of the conservative transport terms used. A discussion of the boric acid behavior for two events is given below in the next section.
42 4.8.8 Representative Results Sections 6.6 and 7.5 of the XPC TR 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 TR 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 TR 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 1900 ppm (1.09%, 1090 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 NRC staff finds 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 1. 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 the -A of TR-0516-49422-P, Loss-of-Coolant Accident Evaluation Model, Revision 5 (Reference 2) or the -A of TR-0516-49416-P, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 5 (Reference 3), must be assessed by the applicant for their potential impact on the XPC EM. Any subsequent changes to the XPC methodology require NRC approval.
- 2. Use of the XPC EM, Rev. 1, is limited to evaluations of the NPM-20 design. An applicant or licensee seeking approval to use the XPC EM for a design other than the NPM-20, such as the 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 NPM-20 requires NRC staff review and approval of the applicants or
43 licensees determination of applicability. Changes made to an NPM-20 through established change processes are addressed in L&C #8.
- 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 initial liquid volumes in the CNV, both borated and unborated, 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 analysis demonstrates that the limiting cases remain subcritical assuming ((
}}, 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 pursuant to applicable regulatory requirements, outside the scope of analyses performed in accordance with this EM. An applicant must identify important considerations for 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 that is used in the application of the XPC EM be consistent with the RVV design described in the application, and identified as a required design value to be verified as part of the ASME QME-1 qualification program.
- 7. An applicant or licensee seeking to apply this methodology to a design must include an initial test (first module only) for ESB dissolution testing during the initial test program, unless such testing has already been performed by a licensee. The test must be consistent, to the extent practical, with the analysis methods and the 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, either implicitly or explicitly, that the as-built functionality of the ESB performance is within the bounds established for the test, which includes: 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 control process specifically approved by the NRC staff for changes to NRELAP5 and the NPM model, use of NRELAP5 is limited to Version 1.7 (v1.7) in conjunction with NPM-20 basemodel Revision 5 (or later NRELAP5 versions and/or new NPM-20 basemodel revisions if the changes 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
44 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 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 evaluation model to a design for analysis of boron transport subcriticality shall provide technical specification controls for conditions where operation is below the established reactor coolant system boron concentration threshold, which is calculated to demonstrate subcriticality margin exists for all conditions despite any short-term xenon changes or low decay heat resulting from operation at reduced power levels, and for rapid power ascensions.
The technical specification controls shall establish a reactor coolant system boron concentration limit, as a function of power history, below which operation is not allowed. To bound the allowable operating conditions, the operational limit for the minimum reactor coolant system boron concentration must be used in the evaluation of the boron transport subcriticality events. The range evaluated must include as input the full matrix of decay heat calculation cases identified in Table 1, Table 2 and Table 3 in RAI response 10298 (ML25051A167), XPC.LTR-6. The power ascension rate limit must be consistent with the limiting ascension rate assumed in the operational minimum boron concentration limit analysis.
- 11. An applicant or licensee seeking to apply the XPC evaluation model to a design for analysis of boron transport precipitation must use the maximum allowable Mode 1 (including zero power no xenon conditions) boron concentration operational limit to set the initial boron concentration in the RCS.
6.0 CONCLUSION
S This SE documents the results of the technical evaluation of Topical Report (TR) TR-124587-P, Extended Passive Cooling and Reactivity Control Methodology, Revision 1, which is an extension of the NuScale LOCA evaluation model (EM) (Reference 2) and NuScale non-LOCA EM (Reference 3) to perform post-LOCA and non-LOCA LTC and reactivity control analyses of the US460 plant with the NPM-20 module design. 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 SE, 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
45 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. 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.
7.0 REFERENCES
- 1. NuScale Power, LLC, Submittal of Revision 1 to Standard Design Approval Application, ADAMS Package Accession No. ML23306A033.
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