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LLC Submittal of Topical Report Rod Ejection Accident Methodology, TR-0716-50350, Revision 3
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Issue date:	10/20/2023
From:	Griffith T; NuScale
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	LO-152611 TR-0716-50350-NP, Rev 3
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LO-152611 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com October 20, 2023 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Topical Report Rod Ejection Accident Methodology, TR-0716-50350, Revision 3

REFERENCES:

1. NuScale letter to NRC, NuScale Power, LLC Submittal of Topical Report Rod Ejection Accident Methodology, TR-07116-50350, Revision 2, dated December 17, 2021 (ML21351A399) 2.

NRC letter to NuScale, Supplement to Audit Plan for the Regulatory Audit of NuScale Power Topical Report Supplement Entitiled Statistical Subchannel Analysis Methodology, TR-108601, Revision 1 Incorporating a Limited Scope Audit for Rod Ejection Accident Methodology, TR-0716-50350, Revision 2, dated April 19, 2023 (ML23107A227)

NuScale Power, LLC (NuScale) hereby submits Revision 3 of the topical report entitled, Rod Ejection Accident Methodology, TR-0716-50350. The purpose of this submittal is to provide an updated revision of the rod ejection accident methodology topical report. The content of the revision is consistent with descriptions provided to the NRC during audit of the previous revision of the topical report (Reference 2). Revision 3 supersedes the previously submitted revision of this topical report (Reference 1). contains the proprietary version of the report entitled, Rod Ejection Accident Methodology, TR-0716-50350, Revision 3. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. Enclosure 2 contains the nonproprietary version of the report.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

LO-152611 Page 2 of 2 10/20/23 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com If you have any questions, please contact Wren Fowler at 541-452-7183 or at sfowler@nuscalepower.com.

Sincerely, Thomas Griffith Manager, Licensing NuScale Power, LLC Distribution:

Mahmoud Jardaneh, NRC Getachew Tesfaye, NRC Stacy Joseph, NRC : Rod Ejection Accident Methodology, TR-0716-50350-P, Revision 3, proprietary version : Rod Ejection Accident Methodology, TR-0716-50350-NP, Revision 3, nonproprietary version : Affidavit of Carrie Fosaaen, AF-152612 Thomas Griffith

LO-152611 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com :

Rod Ejection Accident Methodology, TR-0716-50350-P, Revision 3, Proprietary Version

LO-152611 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Rod Ejection Accident Methodology, TR-0716-50350-NP, Revision 3, Nonproprietary Version

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 Licensing Topical Report

Rod Ejection Accident Methodology October 2023 Revision 3 Docket: 99902078 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 Licensing Topical Report

No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 Licensing Topical Report

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights.

Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

© Copyright 2023 by NuScale Power, LLC iv CONTENTS Abstract....................................................................................................................................... 1 Executive Summary.................................................................................................................... 2 1.0 Introduction..................................................................................................................... 5 1.1 Purpose................................................................................................................. 5 1.2 Scope.................................................................................................................... 5 1.3 Abbreviations and Definitions................................................................................ 5 2.0 Regulatory Considerations............................................................................................ 8 2.1 Regulatory Requirements...................................................................................... 8 2.1.1 Reactor Coolant System Pressure............................................................8 2.1.2 Fuel Cladding Failure.................................................................................8 2.1.3 Core Coolability.........................................................................................9 2.1.4 Fission Product Inventory..........................................................................9 2.2 Regulatory Criteria for NuScale........................................................................... 10 2.2.1 Reactor Coolant System Pressure..........................................................10 2.2.2 Fuel Cladding Failure...............................................................................10 2.2.3 Core Coolability....................................................................................... 11 2.2.4 Fission Product Inventory........................................................................ 11 3.0 Overview and Evaluation Codes.................................................................................. 12 3.1 Overview............................................................................................................. 12 3.1.1 Reactivity Considerations........................................................................12 3.1.2 Reactor Coolant System Pressure Behavior...........................................13 3.2 Analysis Computer Codes and Evaluation Flow.................................................. 13 3.2.1 Core Response........................................................................................15 3.2.2 System Response....................................................................................20 3.2.3 Detailed Thermal-Hydraulic and Fuel Response.....................................20 3.2.4 Accident Radiological Evaluation.............................................................21 4.0 Identification of Important Phenomena for Rod Ejection Accident.......................... 22 4.1 Industry Phenomena Identification and Ranking Table for Rod Ejection Accident............................................................................................................... 22 4.2 Electric Power Research Institute Technical Report............................................ 24 4.3 Standard Review Plan Section 15.4.8 Initial Conditions...................................... 25

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

5.0 Rod Ejection Accident Analysis Methodology........................................................... 27 5.1 Rod Ejection Accident Analysis General Assumptions........................................ 27 5.1.1 Cycle Design............................................................................................27 5.1.2 Cycle Burnup...........................................................................................27 5.1.3 Core Power..............................................................................................27 5.1.4 Single Active Failure................................................................................28 5.1.5 Automatic System Response of Non-Safety Systems.............................28 5.1.6 Loss of Alternating Current Power...........................................................28 5.2 Core Response Methodology.............................................................................. 29 5.2.1 Calculation Procedure.............................................................................29 5.2.2 Analysis Assumptions and Parameter Uncertainties for Core Response................................................................................................31 5.2.3 Results and Downstream Applicability.....................................................33 5.3 System Response............................................................................................... 33 5.3.1 Calculation Procedure.............................................................................33 5.3.2 Analysis Assumptions and Parameter Treatment for System Response................................................................................................35 5.3.3 Results and Downstream Applicability.....................................................36 5.4 Detailed Thermal-Hydraulic and Fuel Response................................................. 36 5.4.1 Subchannel Calculation Procedure.........................................................36 5.4.2 Analysis Assumptions and Parameter Treatment for Subchannel Response................................................................................................38 5.4.3 Fuel Response Calculation Procedure....................................................40 5.4.4 Results and Downstream Applicability.....................................................41 5.4.5 Sensitivity Studies....................................................................................41 5.5 Radiological Assessment.................................................................................... 43 5.6 Method Summary................................................................................................ 43 6.0 Sample Rod Ejection Calculations.............................................................................. 45 6.1 Rod Ejection Accident Sample Analysis System Pressure Response Results................................................................................................................ 47 6.2 NRELAP5 Minimum Critical Heat Flux Ratio Impacts......................................... 49 6.3 VIPRE-01 Sensitivities........................................................................................ 49 6.3.1 Computational Time Steps.......................................................................49 6.3.2 Code Axial Node Lengths........................................................................50

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

© Copyright 2023 by NuScale Power, LLC vi 6.3.3 Two-Phase Flow Correlation Options......................................................51 6.3.4 Numerical Solution Damping Factors......................................................52 6.3.5 Radial Power Distribution and Nodalization.............................................53 6.3.6 Fuel Rod Heat Transfer...........................................................................55 7.0 Summary and Conclusions.......................................................................................... 57 8.0 References..................................................................................................................... 59 8.1 Source Documents.............................................................................................. 59 8.2 Referenced Documents....................................................................................... 59 Appendix A.

NRC Acceptance of NuScale Validation of SIMULATE-3K.......................... 62

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

© Copyright 2023 by NuScale Power, LLC vii TABLES Table 1-1 Abbreviations......................................................................................................... 5 Table 1-2 Definitions.............................................................................................................. 7 Table 2-1 Method for addressing regulatory criteria............................................................ 10 Table 3-1 High-level discipline and code interface............................................................... 14 Table 4-1 Plant transient analysis phenomena identification and ranking table rankings............................................................................................................... 23 Table 4-2 Fuel response phenomena identification and ranking table rankings.................. 23 Table 5-1 Example uncertainties for rod ejection accident calculations............................... 33 Table 5-2 Default VIPRE-01 convergence parameters and options.................................... 38 Table 5-3 Sensitivity studies for rod ejection subchannel evaluations................................. 42 Table 5-4 Summary of simulation types needed to implement method............................... 43 Table 6-1 Sensitivity studies for rod ejection subchannel evaluations................................. 45 Table 6-2 NRELAP5 MCHFR impacts from sensitivity evaluation....................................... 49 Table 7-1 Summary of NuScale criteria and sample evaluation results............................... 58 Table A-1 Range of comparison for SPERT-III.................................................................... 62 Table A-2 Summary of selected SPERT-III cases................................................................ 63 Table A-3 Tabulated results and comparisons of selected SPERT-III cases....................... 64 Table A-4 NEACRP Benchmark Results Comparison......................................................... 70 FIGURES Figure 3-1 Calculation schematic for analyzing rod ejection accident................................... 14 Figure 5-1 Control rod assembly layout for the NuScale Power Module............................... 32 Figure 5-2 PCMI failure threshold curves for unlined RXA fuel cladding temperatures equal to or above 500 °F, and below 500 °F................................. 40 Figure 6-1 Power response at 55 percent power, end of cycle............................................. 46 Figure 6-2 Power response at 100 percent power, beginning of cycle................................. 47 Figure 6-3 Power response for peak reactor coolant system pressure evaluation................ 48 Figure 6-4 Pressure response for peak reactor coolant system pressure evaluation........... 48 Figure 6-5 Time step effect on power forcing function.......................................................... 50 Figure 6-6 Effect of axial node size (inches) on critical heat flux.......................................... 51 Figure 6-7 Effect of VIPRE-01 two-phase flow model options on critical heat flux................ 52 Figure 6-8 Effect of VIPRE-01 damping factors on critical heat flux..................................... 53 Figure 6-9 Example case-specific core radial power distribution at time of peak power................................................................................................................... 54 Figure 6-10 Example case-specific hot assembly radial power distribution at time of peak power.......................................................................................................... 54 Figure 6-11 Radial nodalization sensitivity MCHFR comparison............................................ 55 Figure 6-12 Effect of heat transfer inputs on critical heat flux................................................. 56 Figure A-1 Test 43 SIMULATE-3K comparison to SPERT-III................................................ 65 Figure A-2 Test 70 SIMULATE-3K comparison to SPERT-III................................................ 66 Figure A-3 Test 60 SIMULATE-3K comparison to SPERT-III................................................ 67 Figure A-4 Test 81 SIMULATE-3K comparison to SPERT-III................................................ 68 Figure A-5 Test 86 SIMULATE-3K comparison to SPERT-III................................................ 69

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

Abstract This report documents the NuScale Power, LLC, (NuScale) methodology for the evaluation of a control rod ejection accident (REA) in the NuScale Power Module (NPM). This methodology is used to demonstrate compliance with the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 13 and GDC 28, and the acceptance criteria and guidance in Regulatory Guide (RG) 1.236, NUREG-0800 Standard Review Plan (SRP) Section 4.2, and SRP Section 15.4.8.

The methodology described herein uses a variety of codes and methods. The three-dimensional neutronic behavior is analyzed using SIMULATE5 and SIMULATE-3K; the reactor system response is analyzed using NRELAP5; and the subchannel thermal-hydraulic behavior and fuel response, including transient fuel enthalpy and temperature increases, is analyzed using VIPRE-01. The software is validated for use to evaluate the REA.

This report includes the identification of important phenomena and input and specifies appropriate uncertainty treatment of the important input for a conservative evaluation. The methodology is discussed and demonstrated by the execution of sample calculations and appropriate sensitivity analyses.

NuScale intends to use this methodology for REA analysis of NPM designs. This report is not intended to provide final design values or results; rather, example values for the various evaluations are provided for illustrative purposes in order to aid the readers understanding of the context of the application of the methodology.

NuScale is requesting Nuclear Regulatory Commission (NRC) review and approval to use the methodology described in this report for design-basis REA analyses of NPM designs.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

Executive Summary The purpose of this report is to describe the methodology that NuScale Power, LLC, intends to use for the analysis of REAs. NuScale is requesting Nuclear Regulatory Commission review and approval to use the methodology described in this report for analyses of design-basis REA events of NPM designs.

NUREG-0800, SRP, Section 15.4.8 (Reference 8.2.4) categorizes the REA as a postulated accident due to frequency of occurrence and types it as a Reactivity and Power Distribution Anomaly. The purpose of this report is to define and justify the methodology for analyzing the REA for the NPM designs for the purpose of demonstrating that fuel failure does not occur. This is accomplished by conservatively applying regulatory acceptance criteria to bounding analyses.

Specific regulatory acceptance criteria that are conservatively treated in this methodology include the following:

hot zero power fuel cladding failure applies the worst-case allowed peak rod differential pressure to the allowed radial average fuel enthalpy limit.

pellet-cladding mechanical interaction (PCMI) failure threshold applies a bounding value of cladding excess hydrogen content to assess fuel enthalpy rise limit.

core coolability limit for fuel melt does not allow any fuel melt to occur.

no fuel cladding failure due to minimum critical heat flux criteria (MCHFR) is allowed.

An REA is an assumed rupture of the control rod drive mechanism (CRDM) or of the CRDM nozzle. Upon this rupture, the pressure in the reactor coolant system (RCS) provides an upward force that rapidly ejects the control rod assembly (CRA) from the core. The ejection of the CRA results in a large positive reactivity addition, leading to a skewed and severely peaked core power distribution. As the power rapidly rises, fission energy accumulates in the fuel rods faster than it can be deposited into the coolant, raising the fuel temperature. The power rise is mitigated by fuel temperature feedback and delayed neutron effects.

The regulatory requirements for the REA are GDC 13 and GDC 28 from 10 CFR 50, Appendix A (Reference 8.2.1). In order to satisfy GDC 13 and GDC 28, this methodology utilizes the guidance provided in RG 1.236 (Reference 8.2.2), and SRP Sections 15.4.8 and 4.2. This guidance addresses: 1) maximum RCS pressure, 2) fuel cladding failure, 3) core coolability, and 4) fission product inventory.

This report describes the software codes used to evaluate the REA along with appropriate validation for its use in NPM applications. The software is controlled under the NuScale quality assurance program (Reference 8.1.3). The codes used for REA analysis are the following:

CASMO5 - transport theory code that generates pin cell or assembly lattice physics parameters.

SIMULATE5 - three-dimensional, steady-state, nodal diffusion theory reactor simulator code that calculates steady-state predictions (critical boron concentration, boron worth, reactivity coefficients, CRA worth, shutdown margin, power distributions, and peaking factors).

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

SIMULATE-3K-three-dimensional nodal reactor kinetics code that couples core neutronics with detailed thermal-hydraulic models to supply power input to NRELAP5 and VIPRE-01.

NRELAP5 - System thermal-hydraulic code produced by NuScale to produce boundary conditions to apply to the fuel sub-channel code.

VIPRE Fuel thermal-hydraulic subchannel code predicts three-dimensional velocity, pressure, thermal energy fields, radial fuel rod temperature and enthalpy profiles in reactor cores.

This report presents the findings documented in NUREG/CR-6742 (Reference 8.2.25),

Phenomena Identification and Ranking Table (PIRT) for Rod Ejection Accidents in Pressurized Water Reactors Containing High Burnup Fuel, identifying important phenomena. Associated with these phenomena, the Electric Power Research Institute (EPRI) topical report (Reference 8.2.13) for three-dimensional REA analysis identified the key parameters as the following:

ejected CRA worth effective delayed neutron fraction moderator reactivity coefficient Doppler coefficient, and core power peaking Appropriate biasing of these terms and other important parameters are addressed in this report.

As the methodology is developed, each of the important parameters identified in the PIRT are evaluated and are biased appropriately for a conservative evaluation in addressing the NuScale REA regulatory criteria.

The REA methodology includes the following components:

nuclear design and core response system response detailed thermal-hydraulic and fuel response With the rapid nature of the power increase in the REA VIPRE-01 calculations, several deviations from the subchannel methodology (described in Reference 8.2.10), were used to increase convergence and reliability of the final results. The deviations from the subchannel methodology are discussed and justified in this report.

This report describes representative sample calculations employing the REA methodology and demonstrates how the REA behaves when modeling NPM designs. However, NuScale is not seeking approval of the results provided in this report. Appropriately biased key inputs are used for the sample calculations. The NRELAP5 sensitivity studies evaluate changes to RCS average temperature, loss of offsite power, and RCS flow. VIPRE-01 sensitivity calculation results are also provided. Results of the sensitivity cases are discussed. Trends of the important parameters are also presented.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

The REA methodology meets the regulatory requirements following the approved regulatory guidelines. The results of the sample calculations using the REA methodology are provided in the report to demonstrate that the methodology meets the regulatory criteria from References 8.2.2, 8.2.3, and 8.2.4 by meeting the NuScale criteria defined in this report.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

1.0 Introduction A rod ejection accident (REA) is applicable to pressurized water reactor (PWR) designs with control rod assembly (CRA) insertions at the top of the reactor pressure vessel. An REA is an assumed rupture of the control rod drive mechanism (CRDM), or of the CRDM nozzle. Upon this rupture, the pressure in the reactor coolant system (RCS) provides an upward force that rapidly ejects the CRA from the core. The ejection of the CRA results in a large positive reactivity addition, leading to a highly skewed and severely peaked core power distribution. As the power rapidly rises, fission energy accumulates in the fuel rods faster than it can be deposited into the coolant, raising the fuel temperature. The power rise is mitigated by fuel temperature feedback and delayed neutron effects.

The CRDM design in the NuScale Power Module (NPM) is consistent with existing PWR designs (top entry); therefore, REA is the appropriate reactivity insertion accident to analyze for NPM designs.

1.1 Purpose The purpose of this report is to describe the methodology that NuScale intends to use for REA analysis of NPM designs. This methodology is used in the analysis that supports results reported in Section 15.4.8 of a Final Safety Analysis Report.

1.2 Scope This report describes the assumptions, codes, and methodologies used to perform REA analysis. This report is intended to provide the methodology for performing this analysis; the input values and analysis results presented in the report are for demonstration of the analytical methodology and are not meant to represent final analysis results or design values. Analysis results and comparisons to applicable specified regulatory criteria from regulatory guidance are provided for illustration to aid the understanding of the context of the application of these methodologies.

The intention of the methodology herein is to demonstrate that no fuel failure occurs, therefore there is no dose consequence associated with the REA.

1.3 Abbreviations and Definitions Table 1-1 Abbreviations Term Definition BOC beginning of cycle CHF critical heat flux CRA control rod assembly CRDM control rod drive mechanism DTC Doppler temperature coefficient

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

Term Definition EOC end of cycle EPRI Electric Power Research Institute FGR fission gas release FTC fuel temperature coefficient GDC general design criterion HFP hot full power HZP hot zero power IR importance ratio KR knowledge ratio LOCA loss-of-coolant accident MCHFR minimum critical heat flux ratio MOC middle of cycle MTC moderator temperature coefficient NPM NuScale Power Module NRC Nuclear Regulatory Commission NRF nuclear reliability factor PCMI pellet-cladding mechanical interaction PDIL power dependent insertion limit PIRT phenomena identification and ranking table PWR pressurized water reactor RCS reactor coolant system REA rod ejection accident RG regulatory guide RPV reactor pressure vessel SAF single active failure SRP Standard Review Plan TH thermal-hydraulics WRSO worst rod stuck out

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

Table 1-2 Definitions Term Definition eff effective delayed neutron fraction Courant number A stability criterion for numerical analysis that is calculated by: u x t/x, where u is the axial velocity, t is the time step size, and x is the axial node size. It is a dimensionless number used as a necessary condition for convergence of numerical solutions of certain sets of partial differential equations.

FH enthalpy rise hot channel factor IR importance ratio: phenomena score on a scale between 0 and 100 with an increasing score representing increasing importance to the methodology KR knowledge ratio: phenomena score on a scale between 0 and 100 with an increasing score representing increasing knowledge of phenomena MWd/MTU megawatt days per metric ton of uranium

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

2.0 Regulatory Considerations 2.1 Regulatory Requirements The REA is the PWR design basis accident under the scope of reactivity insertion accidents. The regulatory basis for the REA is derived from the General Design Criteria (GDC) of 10 CFR 50 (Reference 8.2.1) Appendix A, specifically GDC 13 and GDC 28.

GDC 13 addresses the use of plant design features and instrumentation that are involved in the termination of an REA. GDC 28 addresses the design of the reactivity control system to limit the degree of power excursion possible during an REA.

This methodology considers the criteria provided in NUREG-0800, the Standard Review Plan (SRP), Sections 4.2 and 15.4.8 (Reference 8.2.3 and Reference 8.2.4) and the guidance described in Regulatory Guide (RG) 1.236 (Reference 8.2.2).

Evaluation criteria specific to REAs, or more generally to reactivity insertion accidents, have been identified in this section to provide a basis for satisfying the above-noted GDCs.

These criteria can be grouped into the following categories: RCS pressure, fuel cladding failure, core coolability, and fission product inventory. Section 2.2 identifies where in this report each of these specific criteria are addressed.

This report presents the NuScale REA methodology and demonstrates that the applicable regulatory acceptance criteria, described in this section, are met.

2.1.1 Reactor Coolant System Pressure The maximum RCS pressure acceptance criterion is defined in References 8.2.2 and 8.2.4 as The maximum reactor pressure during any portion of the assumed excursion should be less than the value that result in stresses that exceed the Service Limit C as defined in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This acceptance criterion can be met by showing the maximum RCS pressure does not exceed 120 percent of the design pressure.

2.1.2 Fuel Cladding Failure The regulatory criteria for evaluating fuel cladding failure are defined in References 8.2.2 and 8.2.3. These criteria are the following:

For zero power conditions, the high temperature cladding failure threshold is expressed in the following relationship based on the internal rod pressure:

Internal rod pressure system pressure: Peak radial average fuel enthalpy =

170 cal/g, and Internal rod pressure > system pressure: Peak radial average fuel enthalpy =

150 cal/g.

For intermediate and full power conditions, fuel cladding failure is presumed if local heat flux exceeds the critical heat flux (CHF) thermal design limit.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

The pellet-cladding mechanical interaction (PCMI) failure threshold is a change in radial average fuel enthalpy greater than the corrosion-dependent limit depicted in Figure B-1 of Reference 8.2.3. This criterion is bounded by the conservative application of the change in enthalpy limit as a function of cladding excess hydrogen given in Reference 8.2.2.

2.1.3 Core Coolability The regulatory criteria for evaluating core coolability are defined in References 8.2.2 and 8.2.3. These criteria are the following:

Peak radial average fuel enthalpy must remain below 230 cal/g.

Peak fuel temperature must remain below incipient fuel melting conditions.

Mechanical energy generated as a result of (1) non-molten fuel-to-coolant interaction and (2) fuel rod burst must be addressed with respect to reactor pressure boundary, reactor internals, and fuel assembly structural integrity.

No loss of coolable geometry due to (1) fuel pellet and cladding fragmentation and dispersal and (2) fuel rod ballooning.

Core coolability conditions due to fuel failure are avoided for the NuScale REA methodology in that CHF is not permitted to occur. Given that CHF does not occur, the fuel rods do not heat up enough to rupture, and core coolability issues due to post-CHF conditions are not possible. Also, PCMI failures are precluded by assuring that the criterion for limiting cladding excess hydrogen content delineated in Section 2.1.2 is met. In addition, the NuScale criteria adopted and delineated in Section 2.2.3 establishes significant margin to the first two criteria. Therefore the last two criteria above are eliminated.

2.1.4 Fission Product Inventory The regulatory criteria for evaluating the fission product inventory are defined in Appendix B of Reference 8.2.2 and in Reference 8.2.3. This criteria is not applicable because fuel failures are not permitted in the methodology described in this topical report.

The revised transient fission gas release (FGR) correlations are listed below. The total fission product inventory is equal to the steady state gap inventory plus the transient FGR derived with the following correlations:

Peak Pellet Burnup < 50 GWd/MTU: Transient FGR (percent) = [(0.26

H) - 13]

Peak Pellet Burnup 50 GWd/MTU: Transient FGR (percent) = [(0.26

H) - 5]

where, FGR = fission gas release, percent (must be > 0)

H = fuel enthalpy increase (cal/g)

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© Copyright 2023 by NuScale Power, LLC 10 2.2 Regulatory Criteria for NuScale Table 2-1 summarizes how the regulatory acceptance criteria from References 8.2.2, 8.2.3, and 8.2.4 are addressed and applied to the NuScale REA methodology within this report.

Table 2-1 Method for addressing regulatory criteria Criteria Criteria Section Method Section Maximum RCS pressure 2.2.1 5.3 Hot zero power (HZP) fuel cladding failure 2.2.2 2.2.2 FGR effect on cladding differential pressure 2.2.2 N/A CHF fuel cladding failure 2.2.2 2.2.3 Cladding excess hydrogen-based PCMI failure 2.2.2 5.4.3 Incipient fuel melting cladding failure 2.2.2 2.2.2 Peak radial average fuel enthalpy for core cooling 2.2.3 2.2.4 Fuel melting for core cooling 2.2.3 2.2.3 Fission product inventory 2.2.4 5.5 2.2.1 Reactor Coolant System Pressure The maximum RCS pressure acceptance criterion of 120 percent of design pressure is used in the methodology. For an NPM design pressure of 2200 psia, for example, the peak pressure during the REA is limited to 2640 psia. RCS conditions are calculated with the NRELAP5 code.

2.2.2 Fuel Cladding Failure The criteria for evaluating fuel cladding failure are listed below.

For zero-power conditions, the high-temperature cladding-failure threshold is expressed in cladding differential pressure. The peak radial average fuel enthalpy is below the 100 cal/g associated with the maximum peak rod differential pressure of P 4.5 MPa. Thus, the predicted cladding differential pressure does not need to be calculated and the impact of transient FGR on internal gas pressure need not be included for the REA.

Fuel cladding failure is presumed if local heat flux exceeds the CHF thermal design limit. Detailed thermal-hydraulic (TH) conditions are calculated using the VIPRE-01 code.

The PCMI failure threshold is a change in radial average fuel enthalpy greater than the cladding excess hydrogen dependent limit depicted in Figure 5-2. A conservative treatment of Figure 5-2 is applied through a single acceptance criteria of 33 cal/g.

If fuel temperature anywhere in the pellet exceeds incipient fuel melting conditions, then fuel cladding failure is presumed. Fuel temperature predictions must be based upon design-specific information accounting for manufacturing tolerances and

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3

© Copyright 2023 by NuScale Power, LLC 11 modeling uncertainties using NRC approved methods including burnup-enhanced effects on pellet radial power distribution, fuel thermal conductivity, and fuel melting temperature. Incipient fuel melt is determined using Equation 12-3 from Reference 8.2.11 while applying a conservative pellet burnup value. Equation 12-3 is applicable for peak rod average burnup to 62 GWd/MTU as identified in Reference 8.2.11.

2.2.3 Core Coolability The regulatory criteria for evaluating core coolability are defined in Reference 8.2.2 and 8.2.3. The following criteria are adopted for the NuScale REA methodology in a bounding fashion:

Peak radial average fuel enthalpy will remain below 230 cal/g.

No fuel melt will occur.

Core coolability concerns due to fuel failure are avoided for the NuScale REA methodology in that CHF is not permitted to occur. Given that CHF does not occur, the fuel rods do not heat up enough to rupture, and coolability issues due to post-CHF conditions are not possible. PCMI failures are precluded by assuring that the criterion for limiting cladding excess hydrogen content delineated in Section 2.2.2 above is met. In addition, the core coolability NuScale criteria delineated above establishes significant margin to the first two criteria from Section 2.1.3. Therefore the last two criteria from Section 2.1.3 are eliminated.

2.2.4 Fission Product Inventory The regulatory transient FGR criteria do not apply to the NuScale REA methodology for the following two reasons:

This methodology requires that no fuel failure occurs, whether due to fuel melt, or transient enthalpy increase, or cladding failure due to minimum critical heat flux ratio (MCHFR), and therefore, the cladding fission product barrier will not be breached.

The regulatory fuel cladding failure criteria in Section 2.2.2, based on cladding differential pressure, incorporates the most limiting criteria for P 4.5 MPa, therefore any increase in pressure that could occur during the transient due to FGR will not change allowed peak radial average fuel enthalpy.

Based on the above two items, the acceptance criterion in Reference 8.2.4 to perform a dose analysis is not required for the NuScale REA methodology.

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3.1 Overview The cause and progression of the REA is described in References 8.2.2 and 8.2.4. For NPM designs, the REA is an assumed rupture of the CRDM or of the CRDM nozzle. An REA will lead to a rapid positive reactivity addition resulting in a power excursion and a skewed and peaked core power distribution. As power rises rapidly, the fission energy accumulates in the fuel rods faster than it can migrate to the coolant, resulting in raised fuel temperatures. The power rise is mitigated by fuel temperature feedback and delayed neutron effects. A reactor trip on high power rate is generated within a few hundredths of a second of the rod ejection and there is a delay before the CRAs are inserted. Some cases with low ejected CRA worth or large negative values of reactivity feedback may not hit the high power rate trip setpoint and will instead settle at a new steady state condition.

The reactor core is protected against severe fuel failure by the reactor protection system and by restrictions of the power dependent insertion limit (PDIL) and axial offset window, which determine the depth of CRA insertion and initial power distribution allowed in the core.

3.1.1 Reactivity Considerations The REA can behave differently based on the static worth of the ejected CRA. For example, REA can behave as follows:

Reactivity insertion close to or greater than effective delayed neutron fraction; this scenario results in a prompt critical scenario.

Reactivity insertion less than the delayed neutron fraction; this scenario is considered sub-prompt critical.

In general, CRAs that are inserted deeper into the core will have a higher static worth.

PDIL insertion depth increases as power decreases. Therefore, higher power cases produce lower ejected CRA worth, and will tend towards the sub-prompt critical scenario.

A higher ejected CRA worth at reduced power can result in prompt critical power excursions. Similarly, a core with a greater positive axial offset will produce a higher static worth.

3.1.1.1 Prompt Critical In a prompt critical scenario, the energy deposition can be defined by the following equation:

= 2

Equation 3-1

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where, Ed = energy deposition,

= static ejected CRA worth,

= delayed neutron fraction, Cp = fuel heat capacity, and D = Doppler temperature coefficient (DTC).

This equation (Equation 5-90 of Reference 8.2.12) implies that the key parameters affecting the energy deposition during a prompt critical REA are the ejected CRA worth, delayed neutron fraction, fuel heat capacity, and the DTC.

3.1.1.2 Sub-Prompt Critical In a sub-prompt critical scenario, the delayed neutrons limit the power excursion, and instead a jump in power occurs. This prompt jump in power can be approximated by the following equation:

=

Equation 3-2

where, Pj = prompt jump power, and Po = initial power.

This equation (Equation 3-35 of Reference 8.2.12) implies that, for a given CRA worth, a higher initial power will result in a larger prompt jump power, and for these cases, the relationship between and has the most significant impact.

3.1.2 Reactor Coolant System Pressure Behavior The trend of CHF with RCS pressure is described in Section 5.3. Differences between the bounding CHF and RCS overpressure calculations are described in Section 5.3.1.

3.2 Analysis Computer Codes and Evaluation Flow The safety analyses of NuScale Final Safety Analysis Report Chapter 15 non-loss of coolant accident (non-LOCA) transients and accidents are performed using the CASMO5/SIMULATE5 code package for reactor core physics parameters, NRELAP5 for the transient system response, and VIPRE-01 for the subchannel analysis and fuel response. The REA methodology follows a similar approach for use of code packages.

The nuclear analysis portion of the REA transient response is performed using the three-dimensional space-time kinetics code SIMULATE-3K. NRELAP5 is used to simulate the RCS response to the core power excursion, and the VIPRE-01 code is used to model

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© Copyright 2023 by NuScale Power, LLC 14 the core thermal response and to calculate the MCHFR, peak fuel temperature, and enthalpy. The software is controlled under the NuScale quality assurance program (Reference 8.1.3). Figure 3-1 depicts the computer codes used and the flow of information between codes and evaluations to address the regulatory acceptance criteria.

Figure 3-1 Calculation schematic for analyzing rod ejection accident Section 5.2 through Section 5.5 further describe how the power as a function of time and elements of the power distributions calculated by SIMULATE-3K are used as input to NRELAP5 and VIPRE-01. The NRELAP5 calculation then provides the core power (same as the power provided by SIMULATE-3K), core inlet flow, core inlet temperature, and core exit pressure forcing functions to VIPRE-01. A simplified definition of the discipline and code interfaces is presented in Table 3-1, below, arranged such that the discipline in the row receives input from the discipline defined in the column.

Table 3-1 High-level discipline and code interface Discipline Steady-State Nuclear (SIMULATE5)

Transient Nuclear (SIMULATE-3K)

Transients (NRELAP5)

Transient Nuclear (SIMULATE-3K)

Steady-state boundary conditions N/A N/A Transients (NRELAP5)

Reactivity coefficients, kinetics parameters Power vs. time N/A Subchannel (VIPRE-

01)

N/A Radial power distribution (includes FH), axial power distribution Event thermal-hydraulic response (power, flow, temperature, pressure)

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© Copyright 2023 by NuScale Power, LLC 15 3.2.1 Core Response Reference 8.2.6 provides the validation of CASMO5/SIMULATE5 to perform steady state neutronics calculations for NPM designs. Validation of SIMULATE-3K for NPM designs is described in this section.

3.2.1.1 CASMO5 CASMO5 (Reference 8.2.15) is a multi-group two-dimensional transport theory code used to generate pin cell or assembly lattice physics parameters, including cross-sections, nuclide concentrations, pin power distributions, and other nuclear data used for core performance analysis for light water reactors. The code is used to generate a neutron data library for use in the three-dimensional steady-state nodal diffusion code SIMULATE5, and the three-dimensional transient nodal code SIMULATE-3K.

CASMO5 solves the two-dimensional neutron transport equation by the Method of Characteristics. The code produces a two-dimensional transport solution based upon heterogeneous model geometry. The CASMO5 geometrical configuration consists of a square pitch array containing cylindrical fuel rods of varying composition. The code input may include burnable absorber rods, cluster control rods, in-core instrument channels, and water gaps, depending on the details of the assembly lattice design.

The CASMO5 nuclear data library consists of 586 energy groups covering a range from 0 to 20 mega electron volts (MeVs). Macroscopic cross sections are directly calculated from the geometries and material properties provided from the code input. Resonance integrals are used to calculate effective absorption and fission cross sections for each fuel rod in the assembly, and Dancoff factors are calculated to account for the shadowing effect in an assembly between different rods.

CASMO5 runs a series of depletions and branch cases to off-nominal conditions in order to generate a neutron data library for SIMULATE5 or SIMULATE-3K. These calculations form a case matrix, which functionalize boron concentration, moderator temperature, fuel temperature, shutdown cooling (isotopic decay between cycles or over long outage times),

and CRA positioning with respect to exposure. The same neutron data library produced by the automated case matrix structure in CASMO5 and used for steady-state neutronic analysis in SIMULATE5 can be used for transient neutronic analysis in SIMULATE-3K.

For the REA analysis, CASMO5 is used to produce a neutron data library for steady-state neutronic calculations performed with SIMULATE5, and for transient neutronic calculations performed with SIMULATE-3K. The use of CASMO5 in this report is consistent with the methodology presented in Reference 8.2.6.

3.2.1.2 SIMULATE5 SIMULATE5 (Reference 8.2.16) is a three-dimensional, steady-state, nodal diffusion theory, reactor simulator code. It solves the multi-group nodal diffusion equation, employing a hybrid microscopic-macroscopic cross-section model that accounts for depletion history effects. SIMULATE5 output includes steady state nuclear analysis predictions, such as critical boron concentration, boron worth, reactivity coefficients, CRA

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In general, the SIMULATE5 core model is based on input of the core geometry and material compositions, core operating conditions, and core configuration. Core geometry is fully represented with radial nodes corresponding to a quarter of an assembly at numerous axial levels. Material properties and cross-sections are assigned to each node.

Section 4.1 identifies that ejected rod worth is the highest ranked phenomena for this event, as well as describing the two sub-components of worth: control rod and flux redistribution. Static worth is the difference between two static criticality calculations.

Reference 8.2.6 provides a robust justification for the ability of SIMULATE5 to accurately predict critical conditions, power distributions, and depletion for of a broad range of reactor designs and operating conditions. Therefore, the SIMULATE5 code provides an excellent tool for predicting ejected rod worths. Additionally, code uncertainty, in the form of nuclear reliability factors (NRFs), conservatively bound differences in code prediction to measurement results and corresponding measurement uncertainty.

Control rod worth is primarily a function of cross-sections and number densities, resulting in a straightforward validation assessment.

The other component of ejected rod worth, flux redistribution, is more complex, with dependence on current power distribution (and the corresponding depletion history that is the integration of power distribution throughout irradiation history). As a result, the intra-assembly power gradient is important for the determination of ejected control rod worth, and thus for the dynamic consequences of the event. Like existing PWRs, an intra-assembly power gradient exists in all assemblies in the NPM core, especially for assemblies on the periphery. This gradient is due to geometric buckling (i.e., radial leakage). Such gradients occur regardless of core design, though lower leakage cores will tend to exhibit lower intra-assembly power gradients than higher leakage cores. Therefore, bounding evaluations of all intra-assembly power gradients, which inherently includes the corresponding depletion history, are evaluated in each application of this method.

Because SIMULATE5 has been shown to reliability predict critical conditions, power distributions, and depletion, it also reliably predicts the ejected rod worth, including consideration of phenomena such as intra-assembly power and depletion gradients.

For the REA analysis, SIMULATE5 is used to initialize the cycle-specific model and reactor conditions for the REA simulation in SIMULATE-3K. SIMULATE5 writes an initial condition restart file containing the core model geometry, including CRA positioning, reactor operating conditions, and detailed depletion history, to establish the initial core conditions before the start of the REA transient. The restart file contains the explicit neutron library data produced in CASMO5 necessary for SIMULATE-3K calculations, and automatically accounts for differences between the SIMULATE5 calculation model and the data necessary for the SIMULATE-3K calculation model to properly execute.

The use of SIMULATE5 in this report is consistent with the methodology presented in Reference 8.2.6.

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© Copyright 2023 by NuScale Power, LLC 17 3.2.1.3 SIMULATE-3K SIMULATE-3K (References 8.2.17, 8.2.18, and 8.2.19) is a three-dimensional nodal reactor kinetics code that couples core neutronics with detailed TH models. The neutronic model solves the transient three-dimensional, two-group neutron diffusion equations using the quadratic polynomial analytic nodal solution technique, or the semi-analytic nodal method. The code incorporates the effects of delayed neutrons, spontaneous fission in the fuel, alpha-neutron interactions from actinide decay, and gamma-neutron interactions from long term fission product decay.

The TH module consists of a conduction model and a hydraulics model. The conduction model calculates the fuel pin surface heat flux and within-pin fuel temperature distribution.

Heat conduction in the fuel pin is governed by the one-dimensional radial heat conduction equation. The heat source is comprised of prompt fission and decay heat. Material properties are temperature and burnup dependent, and gap conductance is dependent on exposure and fuel temperature. The three-dimensional hydraulic model is nodalized with one characteristic TH channel per fuel bundle (no cross flow) and a variable axial mesh.

The hydraulics model calculates the flow, density, and void distributions for the channel.

The TH module is coupled to the neutronics module through the fuel pin heat generation rate, which is based on reactor power. The TH module provides the neutronics module with data to determine cross-section feedback associated with the local thermal conditions. Cross-section feedback is based on coolant density, fuel temperature, CRA type, fuel exposure, void history, control rod history, and fission product inventory. The heat transferred from the fuel to the coolant provides the hydraulic feedback.

The SIMULATE-3K core model is established from SIMULATE5 restart files, which provide core model geometry and loading pattern, fuel assembly data, nodal information containing radial and axial mesh, and detailed depletion history. SIMULATE-3K uses the same cross-section library created from CASMO5 data that was used in SIMULATE5.

The SIMULATE-3K input file is modified for differences between the codes, ((2(a),(c) In summary, the core model and initial conditions for the SIMULATE-3K analysis are set by reading the appropriate SIMULATE5 restart file, making required adjustments to account for differences between the codes, biasing reactivity coefficients (Section 5.2.1.1), and providing transient-specific inputs (Section 5.2.1.2). SIMULATE-3K is used for transient neutronic analysis of the REA at various times in core life, power levels, CRA positions, and initial core conditions. The transient REA analysis determines total core power, reactivity insertion, three-dimensional power distributions, and power peaking.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 18 A combination of CASMO5, SIMULATE5, and SIMULATE-3K are used to calculate the core response and reactivity-related inputs for the downstream evaluations discussed in the following sections. The power response for the accident is determined by SIMULATE-3K for both NRELAP5 and VIPRE-01. 3.2.1.4 Validation of SIMULATE-3K The validation of SIMULATE-3K to determine the transient neutronic response during an REA includes comparisons to steady state neutronics calculations from SIMULATE5, multiple transient benchmark studies performed by the code vendor, Studsvik Scandpower Inc. (Studsvik), and benchmarks performed by NuScale described in this section. Steady-state neutronics calculation comparisons between SIMULATE-3K and SIMULATE5 demonstrate the ability of the SIMULATE-3K neutronics calculation methodology to accurately predict the effects of core physics parameters important to the REA event. These parameters include reactivity coefficients, including moderator temperature coefficient (MTC) and DTC, CRA and ejected worth, delayed neutron fraction, radial and axial power distributions, and power peaking factors. For all parameters except MTC, SIMULATE-3K results were in very good agreement with SIMULATE5 results. SIMULATE-3K MTC results were close to SIMULATE5 results, with SIMULATE-3K values generally more positive than the SIMULATE5 values. This is conservative for the REA analysis, because a more positive MTC limits the negative reactivity insertion from moderator feedback during the event. Section 3.2.1.2 provides further discussion on the robust demonstration of the SIMULATE5 predictions of static ejected rod worths and power distributions, including the consideration of important phenomena such as intra-assembly power and depletion gradients. SIMULATE5 to SIMULATE-3K benchmarks over a broad range of operating conditions show excellent agreement for rod worths and power distributions. Therefore, SIMULATE-3K appropriately models these phenomena. Furthermore, code uncertainty is applied in the analysis to conservatively ensure an under-estimation of rod worth does not occur. SIMULATE-3K REA analysis for NuScale includes SIMULATE5 uncertainty factors on key core physics parameters important to reactivity. These parameters include delayed neutron fraction, ejected CRA worth, inserted CRA worth, MTC, and DTC. Uncertainties calculated with SIMULATE5 are applied to these parameters to either increase the positive reactivity insertion associated with an ejected CRA, or decrease the negative reactivity insertion associated with moderator and fuel temperature feedbacks and associated with the worth of the CRAs after a reactor trip. The agreement between steady-state SIMULATE-3K and SIMULATE5 calculations of the effects of these core physics parameters allow for the adoption of the NRFs determined for SIMULATE5 (Reference 8.2.6) to be used by SIMULATE-3K for NuScale REA analysis. Section 7 of Reference 8.2.6 provides additional detail on the determination and application of NRFs to account for code uncertainty in the transient analyses described in Section 5.2.1. In addition to steady-state comparisons, Studsvik has performed numerous benchmarks demonstrating the ability of SIMULATE-3K to model and accurately predict core physics

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 19 parameters during reactor transients. Two of these benchmarks for REA analysis include experiments performed at the SPERT III E-core research reactor (Reference 8.2.20), and the NEACRP control rod ejection study computational benchmark (Reference 8.2.22). The Studsvik SPERT III benchmark provides measured REA transient data for comparison to SIMULATE-3K. SPERT III was a pressurized water nuclear research reactor that analyzed reactor kinetic behavior under conditions similar to commercial reactors. The SPERT III core resembled a commercial reactor, but of a reduced size more closely resembling the core size of NPMs. The fuel type (uranium dioxide), moderator, system pressure, and certain initial operating conditions considered for SPERT III are also representative of NPMs. This benchmark demonstrates the ability of SIMULATE-3K to model fast reactivity transients in a PWR core (Reference 8.2.21). Similarities between the NPM designs and the SPERT III core, and notably the small core size, demonstrate applicability and suitability for SIMULATE-3K REA transient analysis of the NPMs. In addition to the Studsvik benchmarks aforementioned, NuScale has performed a benchmark of the dynamic reactor response simulated by SIMULATE-3K of the SPERT III experiment. The original experiment included on the order of one hundred unique tests at five different sets of thermal-hydraulic conditions, with varying initial static worths at each statepoint. One test from each condition set that generally corresponds to the highest static worth for the statepoint has been benchmarked. A comparison of key parameters demonstrates that SIMULATE-3K compares to SPERT III with generally excellent agreement; differences are within the experimental uncertainty (with few exceptions), and the major and minor phenomena are correctly predicted. In general, the NPM designs are much closer in size and transient conditions to SPERT III than existing PWRs (i.e., there is less distortion). The modeling node size of the experiment and NPM applications is equivalent at one node per 20 to 25 fuel rods (i.e., 4x4 to 5x5 rod matrix); corresponding to a node size of a quarter of an assembly for NPM applications. The NEACRP control rod ejection study is a computational benchmark that includes a reference solution provided by the PANTHER code, and SIMULATE-3K REA transient results are compared against the reference solution. In this benchmark, a rod ejection accident in a typical commercial PWR at HZP conditions is analyzed. The fuel type (uranium dioxide), moderator, system pressure, and certain initial operating conditions considered for NEACRP are also representative of NPMs. The capability of SIMULATE-3K to model reactivity insertions in the NEACRP benchmark analysis (Reference 8.2.23 and 8.2.24) demonstrates suitability of the code for reactivity transient applications, and specifically REA analysis applications. The SPERT III and NEACRP benchmarks demonstrate the combined transient neutronic, TH, and fuel pin modeling capabilities of SIMULATE-3K. SIMULATE-3K results for maximum power pulse, time to peak power, inserted reactivity, energy release, and fuel centerline temperature were in excellent agreement with the results from the two benchmark studies. Because peak power is proportional to the square of ejected rod worth above prompt critical, the excellent agreement of peak power for the same ejected rod worth in the benchmarks also demonstrates excellent agreement of ejected rod worth. The SIMULATE-3K results for each of these benchmark studies establish the ability of the code to accurately model an REA transient event and predict key reactivity and power-related

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 20 parameters. See Appendix A for further details on the NRC acceptance of the validation of SIMULATE-3K. 3.2.2 System Response The NRELAP5 code was developed based on the Idaho National Laboratory RELAP5-3D© computer code. RELAP5-3D©, version 4.1.3 was procured by NuScale and used as the baseline development platform for the NRELAP5 code. Subsequently, features were added to address unique aspects of the NuScale design and licensing methodology. The NRELAP5 code includes models for characterization of hydrodynamics, heat transfer between structures and fluids, modeling of fuel, reactor kinetics models, and control systems. NRELAP5 uses a two-fluid, non-equilibrium, non-homogenous fluid model to simulate system TH responses. The validation and applicability of NRELAP5 to the NPM designs is described in References 8.2.8 and 8.2.9. 3.2.3 Detailed Thermal-Hydraulic and Fuel Response The analysis software VIPRE-01 was developed primarily based on the COBRA family of codes by Battelle Pacific Northwest Laboratories for the Electric Power Research Institute. The intention was to evaluate nuclear reactor parameters including minimum departure from nucleate boiling ratio, critical power ratio, fuel and cladding temperatures, and reactor coolant state in normal and off-normal conditions. The three-dimensional velocity, pressure, and thermal energy fields and radial fuel rod temperature profiles for single-and two-phase flow in reactor cores are predicted by VIPRE-01. These predictions are made by solving the field equations for mass, energy and momentum using finite differences method for an interconnected array of channels assuming incompressible thermally expandable flow. The equations are solved with no channel size restrictions for stability and with consideration of lateral scaling for key parameters in lumped channels. Although the formulation is based on the fluid being homogeneous, non-mechanistic empirical models are included for subcooled boiling non-equilibrium and vapor/liquid phase slip in two-phase flow. Like other core TH codes, the VIPRE-01 modeling structure is based on subchannel analysis. The core or section of symmetry is defined as an array of parallel flow channels with lateral connections between adjacent channels. These channels characterize the dominant, longitudinal flow (vertical) by nodalization with various models and correlations predicting TH phenomena that contribute to inter-channel exchange of mass, enthalpy, and momentum. These channels can represent all or fractions of the coolant channel bordered by adjacent fuel rods (hence "subchannel") in rod bundles. The axial variation in channel geometry may also be modeled with VIPRE-01. Channels may represent closed tubes as well as larger flow areas consisting of several combined (lumped) subchannels or rod bundles. These channels communicate laterally by diversion crossflow and turbulent mixing.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 21 The original VIPRE-01 version (MOD-01) was submitted to the NRC in 1985 for use in PWR and boiling water reactor licensing applications. A safety evaluation report by the NRC was issued the following year (Reference 8.2.26). The NRC accepted MOD-01 with several specific restrictions and qualifications, limiting its use to PWR licensing applications for heat transfer regimes up to the point of CHF. This approval was contingent on: (a) the CHF correlation and its limit used in the application is approved by the NRC and (b) each organization using VIPRE for licensing calculations are to submit separate documentation justifying their input selection and modeling assumptions. In 1990, the MOD-02 version of VIPRE-01 was submitted to the NRC to review an improved and updated version, including changes and corrections from the MOD-01 version. This version was approved with an issued SER in 1993 (Reference 8.2.14) with the same requirements and qualifications as in the MOD-01 SER. Unless otherwise stated, in the remainder of this report a reference to VIPRE-01 is referring to the MOD-02 version. The fuel rod model utilized in VIPRE-01 is important to the fuel failure modes of critical heat flux, fuel temperature, and fuel enthalpy as described in Section 2.1. These parameters are addressed in the fuel rod conduction model, where a fuel design-specific calibration to COPERNIC is performed as described in Reference 8.2.11. This calibration calculation develops a conservative radial profile, theoretical density, and gap conductance that captures the effects of heat transfer from the fuel pellet to the clad, and ultimately to the coolant. In the application of the method, sensitivity studies on bounding fuel heat transfer inputs must be performed to determine the limiting condition. This calibration is applicable to rod ejection because extreme rod ejection example cases are utilized in the calibration. Additionally, performing time step sensitivities in application calculations demonstrates the simulation adequately addresses the unique heat generation and conduction characteristics of this event, which impacts heat flux and timing. These sensitivity studies confirm the appropriate resolution of the numerical solution. The validation and applicability of VIPRE-01 to the NPM designs is described in Reference 8.2.10. 3.2.4 Accident Radiological Evaluation This methodology requires that no fuel failure occurs, whether due to fuel melt, transient enthalpy increase, or cladding failure due to MCHFR, and therefore, the pellet/cladding gap shall not be breached. In addition, because the fuel enthalpy increase limit already incorporates the worst cladding differential pressure because of FGR, cladding failure as a result of cladding differential pressure will not occur. Therefore no accident radiological consequences will occur for the REA.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 22 4.0 Identification of Important Phenomena for Rod Ejection Accident Reference 8.2.25 presents the phenomena identification and ranking tables (PIRT) for REA. The PIRT addresses the parameters for consideration in modeling the REA to address the relevant regulatory guidance. Note that this PIRT is an industry PIRT based on large-scale reactors and is not an internally developed NuScale PIRT. This PIRT is applicable to the NuScale design because the PIRT is focused on PCMI-related cladding failures, and the fuel design used for NuScale is consistent with that used in larger PWRs (see Reference 8.2.7). Phenomena important to the REA are also identified in Section 15.4.8 of the SRP (Reference 8.2.4) and the EPRI technical report for three-dimensional analysis of REA (Reference 8.2.13). The overall goal of the evaluation of an REA is to: evaluate the integrity of the fuel pin during the power transient. confirm no fuel failures due to exceeding the CHF design limit. evaluate the integrity of the RCS during the pressure increase. 4.1 Industry Phenomena Identification and Ranking Table for Rod Ejection Accident Use of the PIRT information allows the development of conservative assumptions in the REA methodology. These assumptions are addressed in more detail in Section 5.0. The PIRTs are split into four categories, two of which are applicable to the NuScale REA methodology: plant transient analysis and fuel rod transient analysis. The other categories relate to testing, which is not within the scope of this methodology. Each phenomenon in the PIRT is assigned two scores, the importance ratio (IR) and knowledge ratio (KR). These are on scales of 0-100, with 100 IR being extremely important and 100 KR being very well-known and understood. IR scores above 75 signify highly important criteria. Therefore, this section will address those items with an IR of 75 or greater for evaluating REA against the regulatory acceptance criteria. The rod ejection accident PIRT (Reference 8.2.25) provides the REA analysis parameters in Tables 3-1 and 3-3. Table 4-1 and Table 4-2 list the important phenomena for the two applicable categories that apply to the NuScale REA methodology: Table 4-1 for the plant transient analysis and Table 4-2 for the fuel response. Note that for Table 4-2, only the initial conditions and fuel and cladding temperature change items are considered.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 23 Table 4-1 Plant transient analysis phenomena identification and ranking table rankings Phenomenon IR Score KR Score Calculation of Power History During Pulse (Includes Pulse Width) Ejected CRA worth 100 100 Fuel temperature feedback 100 96 Delayed neutron fraction 95 96 Fuel cycle design 92 100 Calculation of Pin Fuel Enthalpy Increase During Pulse (Includes Cladding Temperature) Heat capacities of fuel and cladding 94 90 Pin peaking factors 97 100 Table 4-2 Fuel response phenomena identification and ranking table rankings Phenomenon IR Score KR Score Initial Conditions Gap size 96 82 Gas distribution 79 50 Pellet and cladding dimensions 91 96 Hydrogen distribution 100 50 Power distribution 100 89 Fuel-clad gap friction coefficient 75 30 Condition of oxidation (spalling) 100 46 Coolant conditions 93 96 Bubble size and bubble distribution 83 20 Transient power specification 100 94 Fuel and Cladding Temperature Changes Heat resistances in fuel, gap, and cladding 75 77 Heat capacities of fuel and cladding 88 93 Coolant conditions 85 88 It should be noted that additional parameters for the CHF and pressurization calculations not listed above were considered in the NuScale REA methodology. Discussion of other parameters considered for the methodology is identified in Section 5.3.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 24 Ejected CRA worth is calculated by SIMULATE-3K. A larger worth is conservative, as it will maximize the power pulse. In order to maximize the worth, uncertainty factors are applied to the insertion depth of the CRAs and to the static CRA worth. The positive reactivity insertion of ejected CRA worth is traditionally separated into two components and reflected accordingly in codes such as SIMULATE-3K, that of control rod and flux shape. Control rod reactivity is designated as the specific absorption of neutrons in the control rod absorbers, resulting in reduced neutron multiplication and thus less reactivity. Flux shape reactivity is the change due to changes in neutron production, absorption and leakage as a result of the power distribution. For example, shifting power from a location of low enrichment to high enrichment will result in higher production and less absorption, resulting in an increase in reactivity. The opposite is also true, shifting power from a location of high enrichment to low enrichment will result in a decrease in reactivity. Fuel temperature feedback, in the form of DTC, is calculated by SIMULATE-3K. A less negative DTC is conservative, as DTC is the primary component that arrests the power pulse. In order to make DTC less negative, an uncertainty factor is applied. Delayed neutron fraction, eff, is calculated by SIMULATE-3K. A smaller value of eff is conservative, as is shown in Equation 3-1 and Equation 3-2. In order to minimize eff, an uncertainty factor is applied. Fuel cycle design is performed using CASMO5 and SIMULATE5. The sample calculations provided in this report were developed using an equilibrium cycle. In order to capture effects of the fuel cycle design, the REA is analyzed at beginning of cycle (BOC), middle of cycle (MOC), and end of cycle (EOC), as well as at various reactor power values ranging from HZP to hot full power (HFP). Heat capacity of the fuel is used to calculate the enthalpy and temperature increases in the fuel pellets during the event. Pin peaking factors are calculated by SIMULATE-3K. The largest pin peaking during the event is used to model the limiting node. An uncertainty factor is applied that captures manufacturing tolerances and modeling uncertainties. 4.2 Electric Power Research Institute Technical Report The EPRI technical report (Reference 8.2.13) has identified several key parameters for the three-dimensional analysis methodology. These key parameters are the following: ejected CRA worth delayed neutron fraction MTC fuel temperature (Doppler) coefficient core peaking factor time-in-cycle

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 25 The EPRI topical report states that uncertainty is applied to the ejected CRA worth, and the MTC and DTC. The MTC and time-in-cycle are the only parameters not already addressed as part of the PIRT. The MTC value is calculated by SIMULATE-3K. A less negative MTC is limiting, as the moderator heating during the event will reduce the power excursion. In order to make this value conservative, an uncertainty factor is applied. The REA is evaluated at BOC, MOC, and EOC to determine the worst time-in-cycle. Uncertainty application for each of the key parameters except time-in-life is discussed in Section 5.0. 4.3 Standard Review Plan Section 15.4.8 Initial Conditions In addition to the PIRT and the EPRI topical report, the SRP Section 15.4.8 (Reference 8.2.4) provides considerations for the initial conditions of the event. The items identified are as follows: A. A spectrum of initial conditions, which must include zero, intermediate, and full-power, is considered at the beginning and end of a reactor fuel cycle for examination of upper bounds on possible fuel damage. At-power conditions should include the uncertainties in the calorimetric measurement. This spectrum is evaluated. The two percent power uncertainty is applied at HFP conditions. B. From the initial conditions, considering all possible control rod patterns allowed by technical specification/core operating limit report power-dependent insertion limits, the limiting rod worths are determined. The limiting rod worths will occur when the rods are at the PDIL. All calculations will begin from this point. At insertion depths less than the PDIL (i.e., with the rod less inserted), the ejected rod has a smaller static worth. For both prompt critical and sub-prompt critical responses described in Sections 3.1.1.1 and 3.1.1.2, respectively, a smaller static worth ejection for the same rod at the same conditions (i.e., same kinetics parameters and reactivity coefficients) is less limiting. C. Reactivity coefficient values of the limiting initial conditions must be used at the beginning of the transient. The Doppler and moderator coefficients are the two of most interest. If there is no three-dimensional space-time calculation, the reactivity feedback must be weighted conservatively to account for the variation in the missing dimension(s). The application of the reactivity coefficients is discussed in Section 5.0. D. [] control rod insertion assumptions, which include trip parameters, trip delay time, rod velocity curve, and differential rod worth. Reactor trip is conservatively applied in the methodology. However, for the REA evaluation, the reactor trip has a negligible effect on the limiting cases, because the limiting cases are those that experience prompt, or near prompt, criticality due to the reactivity insertion. These cases will turn around based on reactivity

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 26 feedback, primarily due to DTC. Application of a reactor trip delay, reducing the reactor trip worth, or slowing the speed of CRA insertion capture effects occur well after the power peak, and consequently well after MCHFR. E. [] feedback mechanisms, number of delayed neutron groups, two-dimensional representation of fuel element distribution, primary flow treatment, and scram input. Feedback mechanisms are discussed in Sections 3.1.1 and 3.1.2. The number of delayed neutron groups and two-dimensional representation of the fuel element are addressed in the code discussion in Section 3.2.1. For a given set of initial conditions, primary core flow is conservatively treated to minimize any flow increase, as increased flow would cause an increase in MCHFR. Reactor trip input, though not explicitly important per Reference 8.2.25, will still be modeled in a conservative manner as noted in the above item D.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 27 5.0 Rod Ejection Accident Analysis Methodology As discussed in Section 3.2, the software used and the flow of information between specific codes in the REA analysis is depicted in Figure 3-1. This section describes the method for the use of these computer codes in the modeling of the REA in the unlikely event it should occur in an NPM. Major assumptions for each phase of the REA analysis are discussed within the text for that phase, while the general assumptions are presented at the beginning of this section. 5.1 Rod Ejection Accident Analysis General Assumptions 5.1.1 Cycle Design The REA analysis will be performed for each core reload. Each reload may result in a different power response, both in magnitude as well as radial and axial distributions. As the underlying assumption for the NuScale REA methodology is that no fuel failures will occur, this assumption will need to be confirmed for any design changes that affect the input to the REA analysis. The sample calculation results provided in this report are from evaluations performed using an equilibrium cycle. 5.1.2 Cycle Burnup The REA is analyzed at BOC, MOC, and EOC burnups to bound core reactivity conditions. For prompt critical CRA ejections, it is expected that the limiting MCHFR case will occur at EOC because the delayed neutron fraction is minimized at this time, and a smaller delayed neutron fraction typically maximizes the peak power of the event for a given initial power level. For sub-prompt critical jumps, the limiting MCHFR may not be associated with the maximum peak power. When analyzing MOC, the time in cycle of maximum peaking will be considered if it does not occur at BOC. This time in cycle may not necessarily correspond to a burnup halfway between BOC and EOC. In the event that MOC is more limiting than BOC or EOC, additional analyses at other MOC points should be performed to ensure the limiting case is identified. 5.1.3 Core Power The REA is analyzed at power levels ranging from HZP to HFP. The power levels analyzed will bound the PDIL, axial offset limits, and moderator temperature over the NPM power range; these parameters feed into the reactivity insertion from a REA. The xenon distribution is adjusted to provide a top peaked axial power shape at the axial offset window boundary, which maximizes the worth of the ejected rod. For low powers, top peaked axial power shapes are produced which bound possible axial power shapes while operating the core with rods inserted. Therefore, the rod ejection occurs through a bounding top peaked shape to maximize the rod worth. These top peaking conditions provide a simple method to bound operations. The conditions are realistic axial shapes

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 28 because allowed operations could in principle arrive at such an axial offset through a unique operation history. While changing axial power does affect the radial power distribution, this is both expected and necessary given that the two are inherently coupled. The moderator temperature is a function of core power and is set by the operating strategy for the plant. The NRELAP5 analysis accounts for the flow and temperature response based on power, calculated to satisfy mass and energy conservation. The VIPRE-01 analysis uses the calculated core flow and core inlet temperature directly from NRELAP5 as an input forcing function. As a result, this treatment of moderator temperature ensures conservatism of the analysis conditions. 5.1.4 Single Active Failure The conservative single active failure for radially asymmetric scenarios such as REA is a failure of the flux detector in the high flux region. This is implemented by requiring all four detectors to exceed the high power rate in order to cause a reactor trip. This single active failure does not necessarily increase the severity of the accident. However, there are no known single active failures that would increase the severity. No safety-related systems besides analytical reactor trip limits in the module protection system such as those based on power or pressure are credited. The module protection system provides reactor trip limits that are sufficiently redundant and therefore, a CRA insertion delay is assumed. 5.1.5 Automatic System Response of Non-Safety Systems In an REA scenario, the automatic control systems could respond to limit the power, pressure, and level excursions. The following balance-of-plant and control system automatic responses are therefore not credited: Pressure control is disabled to ensure maximum pressure. Inventory control is disabled to maximize pressurizer level, and thus RPV pressure. Feedwater flow is assumed constant, keeping flow from increasing due to the increase in moderator average temperature. Steam pressure is not permitted to decrease as the power increases. CRA motion, besides the ejection and insertion of the CRAs, are not modeled. The above conservatisms are appropriate for both the MCHFR and maximum pressure cases. 5.1.6 Loss of Alternating Current Power The REA analysis, for the purpose of calculating MCHFR, typically assumes that loss of alternating current (AC) power occurs at the time of reactor trip. However, the timing of the loss of AC power has no effect on the rod ejection accident MCHFR results, as shown in Table 6-2.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 29 For the purpose of determining the limiting RCS pressure, the REA is evaluated with loss of AC power at both the time of event initiation and at the time of reactor trip. The timing of the loss of AC power is an integral part of the biasing considerations listed in Section 5.3.1.2. 5.2 Core Response Methodology 5.2.1 Calculation Procedure The core response REA methodology has two distinct stages. The first stage involves static calculations that use SIMULATE5. This stage establishes the initial conditions for the event. The second stage is the transient simulations with SIMULATE-3K. This stage establishes boundary conditions for the downstream plant response and subchannel calculations. The core response calculations are performed at various bounding combinations of power and burnup as described in Section 5.1.3 and Section 5.1.2, respectively, to determine the conditions where it is necessary to examine the plant response and perform subchannel analyses. The power levels that should be considered in the SIMULATE-3K analyses must cover the entire operating domain, and must take into consideration power levels where changes in behavior of safety systems or plant conditions occur (such as changes in allowed CRA positions). 5.2.1.1 Static Calculations SIMULATE5 is used to run the static portion of the REA calculations for the core response analysis. This static assessment involves two calculations: assessment of the worst rod stuck out (WRSO) and development of the restart file to feed the initial conditions to SIMULATE-3K. ((

}}2(a),(c)

The initial conditions of reactor power, inlet temperature, coolant mass flux, fission product material, identification of the CRA groups, positions of the CRAs, and information about the spacer grids are passed as input to SIMULATE-3K for use in the REA simulation. Treatment of system flow in the NRELAP5 and VIPRE-01 models is described in Section 5.3 and Section 5.4, respectively. Coolant mass flux is a unit conversion from system flow, core bypass fraction, and core flow cross-sectional area. The coolant mass flux passed to SIMULATE-3K for a given initial power is then held constant through a modeling option. Core depletion calculations for entire cycles are performed in SIMULATE5 as described in Section 3.2.1.2 at nominal, high, and low bounds of the system flow analytical limit. The depletion calculations are used to bound variations in cross-section feedback from flow and temperature variation and distributions, both instantaneous and long-term.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 30 5.2.1.2 Transient Calculations with SIMULATE-3K The transient core response to the REA event is analyzed with SIMULATE-3K. The transient simulation involves two calculations: conservatively addressing parameter uncertainties, and final simulation of the transient. Conservatism is applied to key nuclear parameters in SIMULATE-3K to produce a conservative transient response from the code. Conservative factors are applied to the delayed neutron fraction, fuel temperature coefficient (FTC), MTC, and the worth for the ejected CRA and the inserted CRAs after reactor trip. These parameters are adjusted to account for the uncertainty determined for their calculation in SIMULATE-3K. This uncertainty is characterized by the NRFs previously determined for SIMULATE5 (Reference 8.2.6) and demonstrated to be applicable to SIMULATE-3K. Section 7 of Reference 8.2.6 provides additional detail on the determination and application of the NRFs used to account for code uncertainty. The conservative factors are numerical multipliers which are used to adjust the nuclear parameters by a desired conservative factor, where the conservative value is a reference value determined from SIMULATE-3K for a particular parameter, plus or minus the applicable NRF. Conservative factors are applied to case-specific key nuclear parameters that vary with time in life and initial conditions before the event. For the DTC, CRA worth, and delayed neutron fraction, a separate multiplier is applied which reflects the relative uncertainty from Table 5-1. To conservatively incorporate uncertainties for the MTC, ((

}}2(a),(c)

Once the nuclear parameter uncertainties have been incorporated into the input file, the final transient calculation is performed. For each statepoint identified as part of the scope, a case is run for each regulating group. The process for creating the input is as follows: The regulating groups are set at the PDIL. The WRSO is identified for each ejected CRA. If a non-ejected CRA is the WRSO, then it is left at the PDIL position after reactor trip. The axial power shape is chosen such that the axial offset is at the highest allowable value. ((

}}2(a),(c)

}}2(a),(c) 5.2.2 Analysis Assumptions and Parameter Uncertainties for Core Response 5.2.2.1 Control Rod Assembly Position The regulating groups of CRAs are placed at the appropriate PDIL. This assumption will maximize the worth of the ejected CRA. The shutdown bank is assumed to be at the all rods out position. Uncertainty for the CRA position is applied.

5.2.2.2 Worst Rod Stuck Out REA is analyzed with the WRSO. This assumes that the highest worth CRA remains stuck out of the core after the trip. The WRSO is determined for each fuel burnup and power level that is analyzed, and is chosen to be in the same quadrant as the ejected CRA. The assumption of a WRSO covers the potential for a postulated ejected CRA to damage a nearby CRDM. The power pulse, minimum critical heat flux ratio, peak enthalpy, and peak temperature occur prior to control rod insertion from reactor trip for prompt critical cases. Therefore, the WRSO assumption has no impact on the limiting results for those cases. Regardless of the impact for a given case, WRSO is assumed in the REA analyses. 5.2.2.3 Input Parameters and Uncertainty Treatment 5.2.2.3.1 Ejected Rod Time The time to eject the CRA from the core is defined by Equation 5-1.

2

Equation 5-1 The acceleration is calculated based on the CRA cross-sectional area and weight of the CRA and control rod driveshaft. The distance is the depth in the core that the CRA is inserted.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 32 5.2.2.3.2 Ejected Rod Location The core is designed with quadrant symmetry, where CRAs 1, 5, 15, and 16 in Figure 5-1 represent all unique CRA positions in the core. If the core design does not exhibit a one-eighth core or quarter-core symmetric pattern then all regulating control rod locations must be explicitly evaluated. Only the CRAs in the regulating bank are eligible for ejection and considered in the REA methodology. Figure 5-1 Control rod assembly layout for the NuScale Power Module 5.2.2.3.3 Reactor Trips The NPM includes reactor trip signals for high power and high power rate. These reactor trip signals are modeled in SIMULATE-3K using the output of the excore detectors as described in Section 5.2.1.2. The values for the reactor trip setpoints are input parameters that are specified based on the NPM design. The example high power rate reactor trip signal used in the sample calculations in Section 6.0 of this report is produced when the core power increases more than 7.5 percent from the initial power level within 30 seconds. The example high power reactor trip signal is produced when the core power exceeds 115 percent of rated power if the initial condition is above 15 percent power; the example low power setpoint is 25 percent of rated power if the initial power level is below 15 percent. 5.2.2.3.4 Reactivity Feedback The MTC and DTC are biased to be as least negative as possible. The effective delayed neutron fraction (eff) is biased to be as small as possible. For the low CRA worth calculations to determine peak pressure, BOC reactivity feedback parameters are used to minimize the power decrease that occurs after the initial power jump. Specific uncertainties applied are listed in Table 5-1. 8 9 7 10 2 6 11 3 1 5 12 4 16 13 15 14

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 33 For events that increase RCS and fuel temperatures, the least negative MTC and DTC are conservative. For events based on reactivity insertion, a smaller eff is conservative. Each time a rod ejection analysis is performed, the example uncertainties defined in Table 5-1 will be verified to ensure they are current and updated, if applicable, consistent with References 8.2.6 and 8.2.10. Table 5-1 Example uncertainties for rod ejection accident calculations Parameter Uncertainty Analysis Delayed neutron fraction 6 percent SIMULATE-3K Ejected CRA worth 12 percent SIMULATE-3K Doppler temperature coefficient 15 percent SIMULATE-3K MTC 2.5 pcm/°F SIMULATE-3K CRA position 6 steps SIMULATE-3K Initial power 2 percent NRELAP5 FH pin peaking nuclear reliability factor ((

}}2(a),(c)

VIPRE-01 5.2.3 Results and Downstream Applicability No explicit acceptance criteria are evaluated in the core response calculations. Instead, the boundary conditions are generated to be used by the system response, subchannel, and fuel response analyses. Applicable acceptance criteria are applied to these downstream analyses. 5.3 System Response The generic non-LOCA methodology is discussed in more detail in the non-LOCA evaluation methodology topical report (Reference 8.2.9); for the system analysis using NRELAP5, REA utilitizes this methodology. However, in order to assess the NuScale-specific criteria outlined in Section 2.2, some deviations or additions to the non-LOCA methodology are implemented. The event-specific analysis is discussed in this section. 5.3.1 Calculation Procedure For the system response, calculations are performed for the purpose of determining the peak RCS pressure analysis and to provide inputs to the subchannel analysis for CHF determination. The mass and energy release from the postulated depressurization is bounded by other RPV releases, which are evaluated for containment peak pressure. This evaluation included the additional energy generated during the REA.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 34 Critical heat flux scoping cases are performed to determine the general trend and to select the cases to be evaluated in the VIPRE-01 subchannel analysis for final confirmation that no MCHFR fuel failures occur. Competing scenario evaluations exist between the peak pressure and the MCHFR calculations. The two scenarios to consider within the system response are as follows: The SIMULATE-3K power response is used to maximize the impact on MCHFR. This tends to be a rapid, peaked power response due to using the maximum possible ejected CRA worth based on insertion to the PDIL. A reduced ejected CRA worth that raises the power quickly to just below both the high power and high power rate trip limits is used through the point kinetics model within NRELAP5, and reactivity feedback mechanisms are used to hold the power at this level. This delays the trip until the transient is terminated by high RCS pressure. These cases do not have an upstream SIMULATE-3K calculation. For calculations using the SIMULATE-3K power response, the power forcing functions from the SIMULATE-3K analysis are converted from percent power into units of MW for input into the NRELAP5 calculations. The initialization and treatment of uncertainties of the system thermal-hydraulic parameters of moderator temperature and system flow are described as follows. The moderator temperature is a function of core power and set by the operating strategy for the plant. In addition to the various safety analysis considerations such as thermal margins, the selection of the moderator temperature operating band is affected by thermodynamic efficiencies and the strategy for normal plant startup and shutdown. In the NRELAP5 analysis, temperature is initialized with a bounding high value. The VIPRE-01 analysis uses the calculated core flow and inlet temperature directly from NRELAP5 as an input forcing function. For hot zero power, the flow rate in NRELAP5 is modeled based on the natural circulation curve of a very low power (e.g., 0.001 percent) and the flow rate in SIMULATE-3K is modeled assuming a conservatively low value (e.g., 5 percent of rated flow). 5.3.1.1 Minimum Critical Heat Flux Ratio The cases that typically provide the most limiting MCHFR results are those where the static ejected CRA worth is close to or in excess of one dollar. These are the cases analyzed with SIMULATE-3K, generally at powers where the CRA is deeper in the core. Parameters with uncertainties and/or biases such as total system flow, inlet temperature, and outlet pressure that are used by the downstream VIPRE-01 calculations are addressed within the NRELAP5 system calculations. Consideration for conservative system conditions in MCHFR analysis includes maximized net RCS heat input; this is performed by maximizing the difference between reactor power and heat removal through the steam generator. high initial RCS temperature; this forces the liquid temperature closer to saturation, which increases the rate at which vapor, and thus pressure, is generated.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 35 Variable (high and low) core pressure: the flow is subject to a sensitivity study of both increased and decreased pressure in the core. This sensitivity study is required for rod ejection due to the unique nature of the rapid power change and possible impacts on core flow. high reactor power before reactor trips; this requires starting at a high power or sustaining a large power run-up, and is related to a large ejected CRA worth and low Doppler and moderator feedback. high RCS pressurization rate; this is caused by high power and high pressurizer level. 5.3.1.2 Reactor Coolant System Pressurization The cases that generate the highest pressures are those following the second scenario described above; operating at a power just below the high-power reactor trip limits until reactor trip on high pressure. Considerations for conservative system conditions in peak pressure analysis include maximized net RCS heat input during the transient; this is performed by maximizing the difference between reactor power and heat removal through the steam generator. low initial pressure and high initial RCS temperature; this forces the liquid temperature closer to saturation, which increases the rate at which vapor, and thus pressure, is generated. low inlet flow; the flow is reduced by a pressure surge arising from within the core. high reactor power prior to reactor trip; this requires starting at a high power or sustaining a large power run-up, and is related to a large ejected CRA worth and low Doppler and moderator feedback. high RCS pressurization rate; this is caused by high power and high pressurizer level. delayed reactor trip and lower reactor trip worth. unavailability of automatic pressure-limiting systems, including pressurizer spray, pressurizer heater control, RPV volume control, and feedwater and steam pressure control. delay of the high-steam superheat reactor trip signal; reactor trip on high pressure is more conservative, and this can be done by increasing the steam pressure. 5.3.2 Analysis Assumptions and Parameter Treatment for System Response 5.3.2.1 Pressure Relief No pressure reduction is assumed. Reference 8.2.2 states that no credit should be taken for any possible pressure reduction because of the failure of the CRDM or CRDM housing.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 36 5.3.2.2 Core Power Initial power is biased high to account for the calorimetric uncertainty (Table 5-1). This calorimetric uncertainty is applied for the HFP cases by increasing the SIMULATE-3K core power response by a factor of 1.02 for an example core power uncertainty of 2%. 5.3.2.3 Direct Moderator and Cladding Heating Direct moderator and cladding heating is modeled in NRELAP5 calculations. Reference 8.2.2 states that prompt heat generation in the coolant should be considered for pressure surge calculations. 5.3.2.4 Core Inlet Temperature Core inlet temperature is assumed to be constant. High initial temperature is conservative for both MCHFR and overpressure (see Sections 5.3.1.1 and 5.3.1.2). 5.3.2.5 Core Flow Low core flow is conservative for both MCHFR and overpressure (see Sections 5.3.1.1 and 5.3.1.2). 5.3.2.6 System Pressure and Pressurizer Level System pressure and pressurizer level are addressed for MCHFR and system pressurization (see Sections 5.3.1.1 and 5.3.1.2). 5.3.3 Results and Downstream Applicability The primary result of the system response is the peak RPV pressure. Scoping of the MCHFR can be performed to determine the generally limiting scenarios as described in Section 4.3.5 of the Non-LOCA Methodology topical report (Reference 8.2.9); final MCHFR calculations for the limiting scenarios are performed by the subchannel analyses. The overall plant response determined by the NRELAP5 calculations is transferred to the subchannel and fuel response analysis for calculation of MCHFR and radial average fuel enthalpy to establish that fuel cladding failure has not occurred. 5.4 Detailed Thermal-Hydraulic and Fuel Response 5.4.1 Subchannel Calculation Procedure The subchannel scope of calculations considers the MCHFR acceptance criteria. A hot channel that applies all the limiting conditions bounding all other channels in the core is modeled. The boundary conditions from NRELAP5 of core exit pressure, system flow, and core inlet temperature and the power forcing function from SIMULATE-3K are applied to the VIPRE-01 model. The MCHFR calculations are performed to verify that CHF is not reached during the event for any rods.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 37 5.4.1.1 VIPRE-01 Deviations from Subchannel Methodology With the rapid nature of the power increase in the REA VIPRE-01 calculations, several deviations from the subchannel methodology described in Reference 8.2.10 are used to increase the convergence and reliability of the final results. These changes are described below. ((

}}2(a),(c)

The radial nodalization of the subchannel basemodel is a ((

}}2(a),(c) The phenomenological characteristics of the rod ejection event is unique compared to other events. For a rod that does not experience critical heat flux, the thermal-hydraulics change negligibly while the nuclear physics change dramatically.

Sensitivity studies are used to confirm the radial nodalization of the model accurately maintains the hot channel flow field and results in a conservative MCHFR. Reference 8.2.10 demonstrates ((

}}2(a),(c)

}}2(a),(c)

The default convergence parameters and options for use in VIPRE-01 input is provided in Table 5-2. Table 5-2 Default VIPRE-01 convergence parameters and options Variable Value Description ((

}}2(a),(c) 5.4.2 Analysis Assumptions and Parameter Treatment for Subchannel Response 5.4.2.1 Radial Power Distribution The radial power distribution to be used for the subchannel REA evaluations is a case-specific distribution based on the highest peaked FH rod at the time of peak neutron power as predicted in the SIMULATE-3K analysis. This condition will occur after the ejected CRA is fully out of the core. The peak neutron power will occur after the rod is fully ejected and therefore will represent a skewed power distribution.

With the statistical subchannel methodology defined in Reference 8.2.10, radial peaking uncertainties other than the FH pin peaking NRF are treated within the CHF analysis limit. However, the applicable FH engineering uncertainty for the enthalpy and fuel temperature acceptance criteria (contained in the CHF analysis limit for the MCHFR acceptance criteria) must be applied for these acceptance criteria. Therefore, two options exist for appropriately treating the radial peaking uncertainties for all acceptance criteria.

}}2(a),(c)

Because the analysis is performed for each unique operating cycle core design, the three-dimensional power distribution for each of the screened cases from SIMULATE-3K are analyzed in VIPRE-01. Therefore, design variations, such as enrichment, burnable poison, and loading patterns are captured in these detailed power shapes. The process for transferring the SIMULATE-3K power distribution output at the time of peak power inputs to VIPRE-01 for each case is described. SIMULATE-3K offers choices for editing this information, from fully-detailed powers at each time step, to only a specific assembly at a specific time step. Depending on the preference of the analyst, simple lookup of the required output or an iterative approach may be needed to determine the correct time-step and hot assembly. Once the availability and location of this SIMULATE-3K information is determined, it is transferred to VIPRE-01 input. ((

}}2(a),(c)

The conservative nature of this modeling is described in Section 5.4.1.1. Additionally, as described in Section 6.4.2 of Reference 8.2.10, the radial power distribution more than a few rows removed from the hot subchannel has a negligible impact on the MCHFR results. Analysis of different power distributions of the NPM cores demonstrate that rod powers a few rod rows beyond the hot rod or channel have a negligible impact on the MCHFR. 5.4.2.2 Axial Power Distribution The axial power distribution to be used will be a normalized representation of the SIMULATE-3K assembly-average axial power at time of maximum core neutron power for the assembly containing the highest peak FH rod.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 40 5.4.2.3 Core Inlet Flow Distribution The inlet flow distribution for subchannel analyses is described in Reference 8.2.10. For REA calculations, the limiting inlet flow fraction is applied to the assembly containing the rod with the highest FH as described above. 5.4.2.4 Fuel Heat Transfer Bounding fuel heat transfer inputs are used. Sensitivity studies show that high values are more conservative for REA CHF calculations. Section 6.3.6 discusses the effect of a wide range of heat transfer values on MCHFR. 5.4.3 Fuel Response Calculation Procedure VIPRE-01 is used to calculate the peak radial average fuel enthalpy and maximum rise in order to evaluate acceptance criteria established in Reference 8.2.3. For cladding excess hydrogen the NuScale fuel design uses cladding which is an unlined recrystallization annealed (RXA) fuel cladding. Empirically-based PCMI cladding failure threshold curves for RXA at or above 500°F and below 500°F (from Reference 8.2.3) are applicable to the NuScale fuel design and are shown in Figure 5-2. The most conservative application of these criteria are applied; a limit of 33 cal/g is established so the initial cladding temperature and exposure is not tracked and the excess cladding hydrogen content is not calculated. Figure 5-2 PCMI failure threshold curves for unlined RXA fuel cladding temperatures equal to or above 500 °F, and below 500 °F

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 41 5.4.4 Results and Downstream Applicability The VIPRE-01 analysis is used to demonstrate that no fuel failures are present, using the regulatory criteria outlined in Section 2.1. 5.4.5 Sensitivity Studies Table 5-3 provides a summary and cross-reference of the sensitivity studies associated with the subchannel REA evaluations. Some of the sensitivity studies are mandatory while others are optional. Mandatory sensitivity studies are those that are required in each implementation of the method. Optional sensitivity studies are those that are only required to be performed if a non-default nodalization, parameter, or option is utilized. In general, the purpose of these parametric sensitivity studies is to demonstrate a reliable and converged solution. In addition to the VIPRE-01 assessment of convergence (yes or no), Courant number, and mass and energy errors, several other output parameters are evaluated. The sensitivity studies focus on the MCHFR results because MCHFR is typically the most limiting acceptance criteria. In most cases, there is no difference in sensitivity among the different acceptance criteria. The sensitivity studies are performed to look for trends or differences that would indicate that the solutions are not properly converged. The procedure for each sensitivity study is to change an input parameter or modeling feature to alternate values that cover a range of possible values above and below the reference value. If the input parameter or modeling feature is a binary variable rather than a continuous variable, then the sensitivity study uses the single other value. Regulatory Guide 1.203 (Reference 8.2.30) provides acceptance criteria for the ability of a computer code to model phenomena of interest when compared to experimental data. In the context of code-to-data comparisons, RG 1.203 provides a definition for the phrase Excellent Agreement. This definition is used as a model for developing acceptance criteria for the rod ejection subchannel sensitivity studies. Each sensitivity study should exhibit Excellent Agreement between the reference case and the other cases with different values. Specifically, Excellent Agreement is defined in this context to be when sensitivity cases (as compared to a reference case) exhibit no deficiencies in modeling a given behavior. Major and minor phenomena and trends are correctly predicted and agree closely with the reference case. If Excellent Agreement is not achieved, then the simulation is unreliable and therefore cannot be used. If this occurs, the analyst shall investigate the cause of the issue and make necessary adjustments to analyses or the cycle-specific core design and operating limits. There are three exceptions from the general formulation described above. The first exception is with respect to the fuel heat transfer inputs described in Section 5.4.2.4. Consistent with Table 5-3 (row 1), the purpose of the sensitivity study for fuel heat transfer inputs that is performed in each implementation of the methodology is to determine appropriate inputs to calculate limiting values for the acceptance criteria. The direction of fuel heat transfer inputs that lead to limiting results may change depending on the specific event trajectory. The second exception is the time-step size plot that compares the power as a function of time calculated by SIMULATE-3K versus the discretization of that boundary condition input used in the VIPRE-01 simulation. The acceptance criterion for this plot is that the linear interpolation between two discrete points of the VIPRE-01

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 42 boundary condition must be in Excellent Agreement with the reference. The third exception is for the radial nodalization sensitivity study. It is expected that the single channel model will be more conservative than the other models as shown in Figure 6-11 and discussed in Section 6.3.5. The acceptance criterion for this sensitivity is to demonstrate conservatism of the selected model since the trends for the models may not meet the threshold for excellent agreement. This acceptance criterion includes both an explicit quantitative comparison of the values and a reasonable explanation for observed differences. Table 5-3 Sensitivity studies for rod ejection subchannel evaluations Description Required? Purpose Acceptance Criteria Comments Example Fuel heat transfer inputs Mandatory Determine limiting value for acceptance criteria N/A None Figure 6-12 Time-step size plot Mandatory Confirm valid solution Excellent Agreement Resolve phenomena Figure 6-5 Two-phase flow correlations Mandatory Confirm valid solution Excellent Agreement Courant number Figure 6-7 Axial nodalization Optional - only required if subchannel default nodalization is not used Confirm valid solution Excellent Agreement Courant number, resolve phenomena Figure 6-6 Radial nodalization Optional - only required if subchannel default nodalization is not used Confirm valid solution Demonstrate Conservatism Appropriate and conservative local conditions Figure 6-11 Convergence parameters Optional - only required if subchannel default parameters are not used Confirm valid solution Excellent Agreement Prevent false convergence Figure 6-8 Convergence option deviations Optional - only required if subchannel default options are not used Confirm valid solution Excellent Agreement Prevent false convergence N/A

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 43 5.5 Radiological Assessment An accident radiological calculation is not performed because no fuel failures are predicted. 5.6 Method Summary As introduced in Section 3.2, four separate codes are required to model each rod ejection transient with a high-level flow depicted in Figure 3-1. At each code interface, Table 3-1 provides the information transferred between codes. However, to appropriately bound allowed operations, apply uncertainties, and confirm validity of simulation results, more than one simulation per code is often required. Table 5-4 provides a summary of the simulation types required to fully implement the method. Table 5-4 Summary of simulation types needed to implement method Code Simulation Type Description Downstream Code Purpose SIMULATE5 Worst rod stuck out Valid input SIMULATE5 Pre-ejection conditions SIMULATE-3K SIMULATE-3K Iterative determination of MTC Valid input SIMULATE-3K Confirm rod worth multiplier Valid input SIMULATE-3K Nuclear transient simulation NRELAP5 SIMULATE-3K Re-run transient with power distribution edits VIPRE-01 NRELAP5 Core boundary conditions VIPRE-01 NRELAP5 Peak pressurization case Acceptance criteria VIPRE-01 Core thermal-hydraulics simulation Acceptance criteria VIPRE-01 Sensitivity - fuel heat transfer Limiting value VIPRE-01 Sensitivity - two-phase flow correlations Valid solution VIPRE-01 Sensitivity - axial nodalization if non-default Valid solution VIPRE-01 Sensitivity - radial nodalization if non-default Valid solution VIPRE-01 Sensitivity - convergence parameters if non-default Valid solution VIPRE-01 Sensitivity - convergence options if non-default Valid solution These simulation types are applied for each combination of inputs in the case matrix, which is unique to each discipline as described in this section. The core response

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 44 (SIMULATE5 and SIMULATE-3K) develop a case matrix that parameterizes initial power level, cycle exposure, and ejected rod location. The cases that correspond to the limiting peak powers for each initial power level are provided for system response (NRELAP5). In the system response, each case is run with initial reactor coolant system temperature bias. Transient core boundary conditions for each case are provided for use in detailed core thermal-hydraulics and fuel response (VIPRE-01). Additionally, simulations to check that the peak reactor coolant system pressure is found are performed. VIPRE-01 simulations of transients are performed to determine the limiting case. Upon determination of the limiting case, the required sensitivity studies as dictated by Section 5.4.5 are performed. As a result of completion of the simulation types for the cases in each case matrix, limiting values for each acceptance criteria may be compared to the criteria defined in Table 7-1.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 45 6.0 Sample Rod Ejection Calculations Examples of implementing calculations, as well as sensitivity studies, are presented to provide context. The assumptions and inputs used in the examples in Section 6.0 are not required to be used in the applications of the methodology. Instead, application-specific sensitivity studies or other evaluations are required to determine the appropriate assumptions and input for a given core design, consistent with the methodology in Section 5.0. Table 6-1 lists each of the example figures provided in this section and identifies the purpose of the figure. Specifically, Table 6-1 identifies whether the figure is a sub-step from an example calculation or an example of a sensitivity study as described in Section 5.4.5. Table 6-1 Sensitivity studies for rod ejection subchannel evaluations Figure Number Figure Purpose Figure Title Figure 6-1 Example result Power response at 55 percent power, end of cycle Figure 6-2 Example result Power response at 100 percent power, beginning of cycle Figure 6-3 Example result Power response for peak reactor coolant system pressure evaluation Figure 6-4 Example result Pressure response for peak reactor coolant system pressure evaluation Figure 6-5 Example sensitivity study (Table 5-3) Time step effect on power forcing function Figure 6-6 Example sensitivity study (Table 5-3) Effect of axial node size (inches) on critical heat flux Figure 6-7 Example sensitivity study (Table 5-3) Effect of VIPRE-01 two-phase flow model options on critical heat flux Figure 6-8 Example sensitivity study (Table 5-3) Effect of VIPRE-01 damping factors on critical heat flux Figure 6-9 Example input Example case-specific core radial power distribution at time of peak power Figure 6-10 Example input Example case-specific hot assembly radial power distribution at time of peak power Figure 6-11 Example sensitivity study (Table 5-3) Radial nodalization sensitivity MCHFR comparison Figure 6-12 Example sensitivity study (Table 5-3) Effect of heat transfer inputs on critical heat flux

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 46 Figure 6-1 shows an example of the power response at 55 percent and EOC, which is the highest power case of an example core design and operational limits. The large CRA worth, which is effectively a prompt critical reactivity insertion, results in a rapid power increase. This power increase is quickly turned around by the negative MTC and DTC feedback. The reactor trip signal is given early in the transient, as soon as the two operating detectors show a 15 percent power increase, and a delay of two seconds is assumed. After the large, narrow pulse, with a pulse width at half height of 0.12 seconds, a nearly steady state power of around 56 percent is reached due to the uncertainty treatment until the CRAs start moving. Figure 6-1 Power response at 55 percent power, end of cycle In comparison, Figure 6-2 shows an example of the power response of an REA occurring at 100 percent and BOC. At these conditions, the low ejected worth results in a power response of smaller magnitude compared to the prompt response in Figure 6-1. The module protection system limits are not reached and the long term power comes to a new equilibrium steady state power around 106 percent. 0 100 200 300 400 500 600 700 0 0.5 1 1.5 2 2.5 3 3.5 4 Power (percent) Time (s) 55% Power EOC, Inner Bank Power

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 47 Figure 6-2 Power response at 100 percent power, beginning of cycle 6.1 Rod Ejection Accident Sample Analysis System Pressure Response Results Figure 6-3 provides the power response for an example peak RCS pressure evaluation. Figure 6-4 provides the peak RCS pressure response with this example power forcing function. This calculation, as noted in the NRELAP5 methodology presented in Section 5.3, uses reactivity insertion and feedback inputs that allow the reactor power to jump to a level that is just below the trip setpoints for high reactor power and high power rate. The power is then held at this level until the reactor trip on high reactor pressure is reached. Pressure increases because of the increased power and also because of the loss of AC power assumed at the time of reactor trip. The peak pressure reached during this example REA is 2076 psia. 98 100 102 104 106 108 110 112 114 0 0.5 1 1.5 2 2.5 3 3.5 4 Power (percent) Time (s) 100 percent Power BOC, Inner Group Power

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 48 Figure 6-3 Power response for peak reactor coolant system pressure evaluation Figure 6-4 Pressure response for peak reactor coolant system pressure evaluation 0 20 40 60 80 100 120 0 20 40 60 80 100 120 Reactor Power (percent) Time (sec) 1800 1850 1900 1950 2000 2050 2100 0 20 40 60 80 100 120 Peak RCS Pressure (psia) Time (sec)

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 49 6.2 NRELAP5 Minimum Critical Heat Flux Ratio Impacts Table 6-2 provides an evaluation of sensitivity calculations performed for the MCHFR in NRELAP5. The data shows the comparative effect on the MCHFR in terms of a percent difference from a nominal example case, based on the EOC 50 percent SIMULATE-3K core response. Table 6-2 NRELAP5 MCHFR impacts from sensitivity evaluation Parameter Change MCHFR Impact RCS average temperature Tavg +10°F (( Loss of offsite power Loss of offsite power initiated concurrent with REA RCS Flow Minimum design flow at 50% power

}}2(a),(c),ECI 6.3 VIPRE-01 Sensitivities Section 5.4.5 defines the mandatory and optional subchannel sensitivity analyses. The following sections provide example results for the subchannel sensitivity analyses.

6.3.1 Computational Time Steps Figure 6-5 provides a comparison between the time step size and power forcing functions used by VIPRE-01 and NRELAP5. VIPRE-01 assumes a time step of ((

}}2(a),(c) seconds, and the markers on the VIPRE-01 trendline are the actual VIPRE-01 time steps; VIPRE-01 linearly interpolates the power between these points.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 50 (( }}2(a),(c) Figure 6-5 Time step effect on power forcing function 6.3.2 Code Axial Node Lengths Figure 6-6 provides a comparison of various axial nodalizations used in VIPRE-01 compared to the resulting CHF value. The largest difference in the MCHFR from the nodalization used in the VIPRE-01 basemodel is ((

}}2(a),(c) As described in Section 5.4.1.1, the axial nodalization used must be shown to result in a reliable and converged solution on an application-specific basis.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 51 (( }}2(a),(c) Figure 6-6 Effect of axial node size (inches) on critical heat flux 6.3.3 Two-Phase Flow Correlation Options Figure 6-7 provides a comparison of the profile-fit model (EPRI) against the non-profile fit subcooled void model (HOMO). This provides additional evidence for robustness of the time step size used and any potential violations of the Courant limit. The MCHFR occurs at the same time step, and all time steps are within ((

}}2(a),(c) in CHFR.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 52 (( }}2(a),(c) Figure 6-7 Effect of VIPRE-01 two-phase flow model options on critical heat flux 6.3.4 Numerical Solution Damping Factors Figure 6-8 shows a comparison of damping factors used in solving the VIPRE-01 numerical solution. ((

}}2(a),(c)

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 53 (( }}2(a),(c) Figure 6-8 Effect of VIPRE-01 damping factors on critical heat flux 6.3.5 Radial Power Distribution and Nodalization Figure 6-9 provides an example core radial power distribution, while Figure 6-10 provides the hot assembly radial power distribution from the limiting statepoint at time of peak power. In the default radial nodalization, these power distribution inputs would be used to represent all fuel rods in the core. ((

}}2(a),(c)

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 54 (( }}2(a),(c) Figure 6-9 Example case-specific core radial power distribution at time of peak power (( }}2(a),(c) Figure 6-10 Example case-specific hot assembly radial power distribution at time of peak power If the default radial nodalization is not used, a sensitivity study is required as described in Section 5.4.5. An example of the single channel radial nodalization for a different case with a peak power of roughly 300% rated power is provided. For this sensitivity study, three different nodalization schemes are examined of ((

}}2(a),(c) The results from this sensitivity study are plotted in Figure 6-11.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 55 (( }}2(a),(c) Figure 6-11 Radial nodalization sensitivity MCHFR comparison As expected from the reasoning provided in Section 5.4, the timing and magnitude of the decrease in MCHFR as the power increases and then is turned around by the Doppler feedback is close for the three cases, with the ((

}}2(a),(c) This sensitivity provides an example justification that the single channel radial nodalization is appropriate for this particular case. As noted above, each implementation of the single channel model for a limiting case requires a similar sensitivity to confirm applicability.

6.3.6 Fuel Rod Heat Transfer Figure 6-12 provides an example sensitivity study as described in Section 5.4.5 for the purpose of determining limiting values for acceptance criteria. Figure 6-12 shows the comparison of high and low heat transfer inputs, specifically fuel rod gap conductance values of ((

}}2(a),(c) BTU/hr-ft2-°F and the effect on CHF. (( 
}}2(a),(c)

}}2(a),(c) Therefore, the fuel heat transfer input sensitivity analyses required by Section 5.4.5 consider the difference in impact for different acceptance criteria.

(( }}2(a),(c) Figure 6-12 Effect of heat transfer inputs on critical heat flux

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 57 7.0 Summary and Conclusions This report described the methodology for the evaluation of an REA in the NPM. This methodology was developed to demonstrate compliance with the requirements of GDC 13 and GDC 28, and the acceptance criteria and guidance in Regulatory Guide 1.236 and SRP Sections 4.2 and 15.4.8. NuScale intends to use this methodology for REA analysis of NPM designs. The methodology presented is not generic for different core designs, therefore cycle-specific analysis must be performed for each core design. The methodology described herein uses a variety of codes and methods. The three-dimensional neutronic behavior is analyzed using SIMULATE5 and SIMULATE-3K; the reactor system response is analyzed using NRELAP5; and the subchannel TH behavior and fuel response, including transient fuel enthalpy and temperature increases, is analyzed using VIPRE-01. The software is validated for use to evaluate the REA. This report includes the identification of important phenomena and input and specifies appropriate uncertainty treatment of the important input for a conservative evaluation. The methodology is discussed and demonstrated by the execution of sample calculations and sensitivity analyses. Section 6.0 of this report provides sample REA sensitivity calculations. These data provide confirmation that the method for satisfying the regulatory acceptance criteria outlined in Section 2.1 are appropriate. The regulatory acceptance criteria are maximum RCS pressure. Results from the sample analysis using the NRELAP5 system code that evaluates the peak NPM pressure due to the power pulse from a worst-case rod ejection demonstrates that the maximum system pressure is well below the criteria of 120 percent of design pressure. fuel cladding failure. Transient enthalpy rise is well below the criteria for HZP, intermediate, and HFP conditions considering fuel rod differential pressure at HZP and cladding excess hydrogen with a wide margin. The subchannel model also predicts that the peak fuel centerline temperature is well below the incipient melting point. For the limiting critical heat flux (CHF) cases VIPRE-01 predicts ample margin to CHF. core coolability. The results associated with core coolability of peak radial average fuel enthalpy are met with ample margin. Incipient fuel melt is precluded by a wide margin. fission product inventory. The fission product inventory effects are not applicable, because no fuel rod failure is allowed and the highest rod differential pressure is assumed for the HZP requirement of transient fuel enthalpy rise. Sample REA analysis quantitative results compared to the regulatory acceptance criteria are summarized below in Table 7-1.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 58 Table 7-1 Summary of NuScale criteria and sample evaluation results Parameter Criteria Sample Evaluation Results - Limiting Case Maximum RCS pressure 120% design 2076 psia (94.4% design) HZP fuel cladding failure (average enthalpy) < 100 cal/g 34.6 cal/g FGR effect on cladding differential pressure 2.3.4 (item 2) N/A CHF fuel cladding failure MCHFR > CHF analysis limit 1.47 Cladding excess hydrogen-based PCMI failure < 33 cal/g 11.9 cal/g Incipient fuel melting cladding failure < incipient fuel melt limit 2162 °F Peak radial average fuel enthalpy for core coolability < 230 cal/g 84.0 cal/g Fuel melting for core cooling < incipient fuel melt limit 2162°F Fission product inventory 2.3.4 N/A

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 59 8.0 References 8.1 Source Documents 8.1.1 American Society of Mechanical Engineers, Quality Assurance Program Requirements for Nuclear Facility Applications, ASMENQA-1-2008, ASME NQA-1a-2009 Addenda, as endorsed by Regulatory Guide 1.28, Revision 4. 8.1.2 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Title 10, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, (10 CFR 50 Appendix B). 8.1.3 NuScale Power, LLC, Quality Assurance Program Description, MN-122626, Revision 0. 8.2 Referenced Documents 8.2.1 U.S. Code of Federal Regulations, Part 50, Title 10, Domestic Licensing of Production and Utilization Facilities (10 CFR 50). 8.2.2 U.S. Nuclear Regulatory Commission, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, Regulatory Guide 1.236, June 2020. 8.2.3 U.S. Nuclear Regulatory Commission, Standard Review Plan, Fuel System Design, NUREG-0800, Section 4.2, Rev. 3, March 2007. 8.2.4 U.S. Nuclear Regulatory Commission, Standard Review Plan, Spectrum of Rod Ejection Accidents (PWR), NUREG-0800, Section 15.4.8, Rev. 3, March 2007. 8.2.5 NuScale Power, LLC, NuScale Power Critical Heat Flux Correlations, TR-0116-21012-P-A, Revision 1. 8.2.6 NuScale Power, LLC, Nuclear Analysis Codes and Methods Qualification, TR-0616-48793-P-A, Revision 1. 8.2.7 NuScale Power, LLC, Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P-A, Revision 1. 8.2.8 NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422-P, Revision 3. 8.2.9 NuScale Power, LLC, Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416-P Revision 4.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 60 8.2.10 NuScale Power, LLC, Statistical Subchannel Analysis Methodology, TR-108601-P, Revision 3. 8.2.11 BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, January 2004. 8.2.12 Hetrick, D. L., Dynamics of Nuclear Reactors, ANS, Illinois, pp. 64 and 166, 1993. 8.2.13 EPRI Technical Report 1003385, Three-Dimensional Rod Ejection Accident Peak Fuel Enthalpy Analysis Methodology, November 2002. 8.2.14 U.S. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to VIPRE-01 Mod 02 for PWR and BWR Applications, EPRI-NP-2511-CCMA, Revision 3, October 30, 1993. 8.2.15 CASMO5: A Fuel Assembly Burnup Program Users Manual, SSP-07/431 Revision 7. Studsvik Scandpower, December 2013. 8.2.16 SIMULATE5 Advanced Three-Dimensional Multigroup Reactor Analysis Code, SSP-10/438 Revision 4. Studsvik Scandpower, December 2013. 8.2.17 SIMULATE-3K Extended Fuel Pin Model, SSP-05/458 Revision 1. Studsvik Scandpower, March 2008. 8.2.18 SIMULATE-3K Input Specification, SSP-98/12 Revision 17. Studsvik Scandpower, September 2013. 8.2.19 SIMULATE-3K Models and Methodology, SSP-98/13 Revision 9. Studsvik Scandpower, September 2013. 8.2.20 R. McCardell, et.al., Reactivity Accident Test Results and Analyses for the SPERT III E-Core - A Small, Oxide-Fueled, Pressurized Water Reactor, IDO-17281. March 1969. 8.2.21 G. Grandi, Validation of CASMO5 / SIMULATE-3K Using the Special Power Excursion Test Reactor III E-Core: Cold Start-Up, Hot Start-Up, Hot Standby and Full Power Conditions. Proceedings of PHYSOR 2014, Kyoto, Japan, September 28-October 3, 2014. 8.2.22 H. Finnemann, A. Galati. NEACRP 3-D LWR Core Transient Benchmark Final Specifications, NEACRP-L-335 Revision 1. EOCD Nuclear Energy Agency, January 1992. 8.2.23 G. Grandi, Effect of the Discretization and Neutronic Thermal Hydraulic Coupling on LWR Transients. Proceedings of NURETH-13, Kanazawa City, Japan, September 27-October 2, 2009.

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 61 8.2.24 LWR Core Reactivity Transients, SIMULATE-3K Models and Assessments, SSP-04/443 Revision 3. Studsvik Scandpower, July 2011. 8.2.25 U.S. Nuclear Regulatory Commission, Phenomenon Identification and Ranking Tables (PIRTs) for Rod Ejection Accidents in Pressurized Water Reactors Containing High Burnup Fuel, NUREG/CR-6742 (LA-UR-99-6810), September 2001. 8.2.26 Safety Evaluation Report on EPRI NP-2511-CCM VIPRE-01, May 1986. 8.2.27 NuScale Power, LLC, Response to RAI 9306, Question 15.04.08-1, June 4, 2018, ADAMS Accession Nos. ML18155A627 (package) and ML18155A628 (public version). 8.2.28 NuScale Power, LLC, Supplemental Response to NRC Request for Additional Information No. 9306 (eRAI No. 9306), Question 15.04.08-1, February 21, 2019, ADAMS Accession Nos. ML19052A611 (package) and ML19052A612 (public version). 8.2.29 U.S. Nuclear Regulatory Commission, Final Safety Evaluation for NuScale Power, LLC Topical Report TR-0716-50350, Revision 1, "Rod Ejection Accident Methodology," dated June 3, 2020 (ML20157A223). 8.2.30 U.S. Nuclear Regulatory Commission, Transient and Accident Analysis Methods, Regulatory Guide 1.203, December 2005 (ML053500170).

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 62 Appendix A. NRC Acceptance of NuScale Validation of SIMULATE-3K The NRC reviewed NuScales benchmark of SIMULATE-3K against a selection of SPERT-III cold startup tests for each statepoint, generally corresponding to the highest static worth for the statepoint (Reference 8.2.21). NuScale compared the SPERT-III conditions with the NPM operating parameters and demonstrated that the SPERT-III test conditions were generally representative of the NPM core designs from a reactivity-initiated accident perspective (References 8.2.27 and 8.2.28). Table A-1 provides the comparison of conditions to SPERT-III, Table A-2 summarizes the SPERT-III cases selected, and Table A-3 provides the results. Additional comparison results are shown in Figure A-1 through Figure A-5. The NRC determined that the NuScale results demonstrated generally good agreement between the results predicted by SIMULATE-3K and the SPERT-III experimental results. Table A-1 Range of comparison for SPERT-III Parameter Units SPERT-III NuScale Power Module Reactor Type PWR PWR Fuel Material Uranium dioxide Uranium dioxide UO2 Enrichment w/o 4.8 4.95 Clad Material Stainless Steel Zircaloy Alloy (M5) Active Fuel Length in 38.3 78.74 Core Diameter in ~26 ~68 Rated Power MWt 20 160-250 Rated Flow kg/s 1,260 680-820 Design Core Exit Temperature F 650 590-610 Design Pressure psia 2,515 1,850-2,000

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 63 Table A-2 Summary of selected SPERT-III cases Test # Statepoint Condition Initial Coolant Temperature (°F) Reactivity Insertion ($) 43 Cold Startup 78 1.210 70 Hot Startup 250 1.210 60 Hot Startup 500 1.230 81 Hot Standby 500 1.170 86 Full Power 500 1.170

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 64 © Copyright 2023 by NuScale Power, LLC Table A-3 Tabulated results and comparisons of selected SPERT-III cases Test # Peak Power (MW) [Exp. Uncertainty = +/-15%] Integrated Energy (MW-sec) [Exp. Uncertainty = +/-17%] Reactivity Compensation ($) [Exp. Uncertainty = +/-11%] S3K SPERT-III Difference S3K SPERT-III Difference S3K SPERT-III Difference 43 (( 280 (( (( 6 (( (( 0.22 (( 70 280 6.3 0.22 60 410

}}2(a),(c) 8.5

}}2(a),(c)

}}2(a),(c) 0.24 
}}2(a),(c) 81 330 86 
}}2(a),(c) 610 
}}2(a),(c)

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 70 Additionally, the NRC reviewed NuScales verification analysis of the NEACRP REA benchmark performed by Studsvik Scandpower with SIMULATE-3K (Reference 8.2.27). This analysis was performed under NuScales approved 10 CFR Part 50, Appendix B, quality assurance program. The results of this analysis are presented below. Table A-4 provides a comparison of the SIMULATE-3K results obtained by NuScale against the NEACRP benchmark reference solutions. Table A-4 NEACRP Benchmark Results Comparison Parameter Case NEACRP S3K Critical Boron Concentration (ppm) A1 567.7 (( A2 1160.6 B1 1254.6 B2 1189.4 C1 1135.3 C2 1160.6 Reactivity Release (pcm) A1 822 A2 90 B1 831 B2 99 C1 958 C2 78 Maximum Power (%) A1 117.9 A2 108.0 B1 244.1 B2 106.3 C1 477.3 C2 107.1 Time of Maximum Power (s) A1 0.56 A2 0.10 B1 0.52 B2 0.12 C1 0.27 C2 0.10 Final Power (%) A1 19.6 A2 103.5 B1 32.0 B2 103.8 C1 14.6 C2 103.0

}}2(a),(c)

Rod Ejection Accident Methodology TR-0716-50350-NP Rev. 3 © Copyright 2023 by NuScale Power, LLC 71 Parameter Case NEACRP S3K Final Average Doppler Temperature (°C) A1 324.3 (( A2 554.6 B1 349.9 B2 552.0 C1 315.9 C2 553.5 Final Maximum Centerline Temperature (°C) A1 673.3 A2 1691.8 B1 559.8 B2 1588.1 C1 676.1 C2 1733.5 Final Coolant Outlet Temperature (°C) A1 293.1 A2 324.6 B1 297.6 B2 324.7 C1 291.5 C2 324.5

}}2(a),(c)

After review the NRC determined that the results demonstrated good agreement between NuScales SIMULATE-3K results and the NEACRP benchmark reference solutions. Based on NuScales analysis results, the NRC found that NuScale demonstrated that SIMULATE-3K can successfully model the NEACRP benchmarks for reactivity-initiated accidents. The NRC concluded that the NuScale validation of SIMULATE-3K against the SPERT-III experiments and the NEACRP benchmark suite, as discussed above, were acceptable and demonstrated that SIMULATE-3K can be used in its methodology to accurately model a reactivity-initiated accident (Reference 8.2.29).

LO-152611 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Affidavit of Carrie Fosaaen, AF-152612

AF-152612 Page 1 of 2 NuScale Power, LLC AFFIDAVIT of Carrie Fosaaen I, Carrie Fosaaen, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the process by which NuScale develops its rod ejection accident methodology. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled, Rod Ejection Accident Methodology, TR-0716-50350, Revision 3. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC §

AF-152612 Page 2 of 2 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on October 20, 2023. Carrie Fosaaen that the foregoing is trtrtrtrrtrtrtruee uee ue uee uee and corr Carrie Fosaaen}}

ML23293A292

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