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LLC, Submittal of NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605, Revision 1
	ML25073A126
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Site:	05200050
Issue date:	03/14/2025
From:	Shaver M; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
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	LO-176338 TR-117605-NP, Revision 1
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LO-176338 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com March 14, 2025 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 NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605, Revision 1

REFERENCE:

NuScale letter to NRC, NuScale Power, LLC Submittal of the NuScale Standard Design Approval Application Part 2 - Final Safety Analysis Report, Chapter 4, Reactor, Revision 0, dated December 28, 2022 (ML22362A079)

NuScale Power, LLC (NuScale) hereby submits Revision 1 of the NuFuel-HTP2 Fuel and Control Rod Assembly Designs, (TR-117605). This revision includes changes made during the Standard Design Approval Application audit. The next revision of the Standard Design Approval Application (Revision 2) will reference this revision of TR-117605. contains the proprietary version of the report entitled, NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605, Revision 1. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavits (Enclosures 3 and 4) support this request. has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirement of 10 CFR Part 810. pertains to the NuScale proprietary information, denoted by double braces (i.e., ((). Enclosure 4 pertains to the Framatome Inc. proprietary information, denoted by brackets (i.e., [ ]). Enclosure 2 contains the nonproprietary version of the report. This letter makes no regulatory commitments and no revisions to any existing regulatory commitments. If you have any questions, please contact Amanda Bode at 541-452-7971 or at abode@nuscalepower.com. I declare under penalty of perjury that the foregoing is true and correct. Executed on March 14, 2025. Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC

LO-176338 Page 2 of 2 03/14/2025 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution: Mahmoud Jardaneh, Chief, New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Stacy Joseph, Senior Project Manager, NRC

NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605-P, Revision 1, Proprietary Version : NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605-NP, Revision 1, Nonproprietary Version : Affidavit of Mark W. Shaver, AF-176339 : Affidavit of Morris Byram, Framatome Inc.

LO-176338 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605-P, Revision 1, Proprietary Version

LO-176338 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605-NP, Revision 1, Nonproprietary Version

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Licensing Technical Report © Copyright 2025 by NuScale Power, LLC i NuFuel-HTP2' Fuel and Control Rod Assembly Designs March 2025 Revision 1 Docket: 52-050 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com © Copyright 2025 by NuScale Power, LLC

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Licensing Technical Report © Copyright 2025 by NuScale Power, LLC ii COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. 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.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Licensing Technical Report © Copyright 2025 by NuScale Power, LLC iii Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928. 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.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Table of Contents © Copyright 2024 by NuScale Power, LLC iv Abstract................................................................... 1 Executive Summary.......................................................... 2 1.0 Introduction.......................................................... 4 1.1 Purpose.............................................................. 4 1.2 Scope................................................................ 4 2.0 Background.......................................................... 7 2.1 Regulatory Requirements................................................ 7 3.0 NuFuel-HTP2' Fuel Assembly Description................................ 8 3.1 Top Nozzle............................................................ 8 3.2 Bottom Nozzle with Mesh Filter Plate....................................... 9 3.3 Zircaloy-4 MONOBLOC' Guide Tubes..................................... 9 3.4 Zircaloy-4 Instrument Tube.............................................. 10 3.5 Zircaloy-4 HTP' Upper and Intermediate Spacer Grids........................ 10 3.6 Alloy 718 HMP' Lower Spacer Grid....................................... 10 3.7 Fuel Rod with Alloy M5 Fuel Rod Cladding................................. 10 4.0 Design Evaluation.................................................... 26 4.1 Fuel System Damage Criteria............................................ 26 4.1.1 Stress and Loading Limits......................................... 26 4.1.2 Cladding Fatigue................................................ 32 4.1.3 Fretting........................................................ 34 4.1.4 Oxidation, Hydriding, and Crud Buildup............................... 35 4.1.5 Fuel Rod Bow................................................... 36 4.1.6 Axial Growth.................................................... 37 4.1.7 Fuel Assembly Distortion Evaluation................................. 38 4.1.8 Fuel Rod Internal Pressure........................................ 39 4.1.9 Assembly Liftoff................................................. 39 4.2 Fuel Rod Failure Criteria................................................ 40 4.2.1 Internal Hydriding................................................ 40 4.2.2 Cladding Collapse............................................... 40 4.2.3 Overheating of Cladding........................................... 43 4.2.4 Overheating of Fuel Pellets........................................ 44 4.2.5 Excessive Fuel Enthalpy.......................................... 47

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Table of Contents © Copyright 2024 by NuScale Power, LLC v 4.2.6 Pellet-Cladding Interaction......................................... 47 4.2.7 Bursting....................................................... 50 4.2.8 Mechanical Fracturing............................................ 50 4.3 Fuel Coolability........................................................ 50 4.3.1 Cladding Embrittlement........................................... 50 4.3.2 Violent Expulsion of Fuel.......................................... 50 4.3.3 Generalized Cladding Melting...................................... 51 4.3.4 Fuel Rod Ballooning.............................................. 51 4.3.5 Fuel Assembly Structural Damage from External Forces................. 51 4.4 Thermal Hydraulic Evaluation............................................ 59 4.4.1 Core Pressure Drop Evaluation..................................... 59 4.4.2 Guide Tube Boiling............................................... 60 4.4.3 Control Rod Drop Analysis......................................... 63 5.0 Fuel Assembly Testing................................................ 66 5.1 Mechanical Testing Summary............................................ 66 5.1.1 Fuel Assembly Lateral Load Deflection (Stiffness) Test................... 66 5.1.2 Fuel Assembly Free Vibration (Lateral Pluck) Test...................... 66 5.1.3 Fuel Assembly Lateral Impact Test.................................. 66 5.1.4 Fuel Assembly Lateral Forced Vibration Test.......................... 67 5.1.5 Fuel Assembly Axial Stiffness Test.................................. 67 5.1.6 Fuel Assembly Drop Test.......................................... 67 5.1.7 Spacer Grid Tests............................................... 67 5.1.8 Top and Bottom Nozzle Tests...................................... 68 5.2 Thermal-Hydraulic Testing Summary....................................... 68 5.2.1 Pressure Drop and Liftoff Testing and Pressure Loss Coefficient Development................................................... 68 5.2.2 Flow-Induced Vibration Testing..................................... 69 6.0 Control Rod Assembly................................................ 70 6.1 Control Rod Assembly Description........................................ 70 6.2 Control Rod Assembly Evaluation......................................... 74 6.2.1 Cladding Strain.................................................. 74 6.2.2 Cladding Creep Collapse.......................................... 75

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 Table of Contents © Copyright 2024 by NuScale Power, LLC vi 6.2.3 Cladding Stress................................................. 75 6.2.4 Cladding Fatigue................................................ 76 6.2.5 Cladding Wear.................................................. 76 6.2.6 Control Rod Internal Pressure...................................... 77 6.2.7 Component Melt Analysis.......................................... 77 6.2.8 Spider Assembly Structural Analysis................................. 78 6.2.9 Control Rod Assembly Impact Velocity Limit........................... 78 6.3 Control Rod Assembly Testing............................................ 78 7.0 Design Change Process............................................... 80 8.0 Summary and Conclusions............................................. 82 9.0 References.......................................................... 83 9.1 Source Documents..................................................... 83

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 List of Tables © Copyright 2024 by NuScale Power, LLC vii Table 1-1 Abbreviations.................................................... 6 Table 3-1 Fuel Assembly Parameters........................................ 12 Table 3-2 Fuel Rod Parameters............................................. 13 Table 3-3 Comparison of Operating Conditions................................. 14 Table 3-4 Fuel Assembly Materials.......................................... 14 Table 4-1 Summary of Results - Fuel Assembly Design Margins................... 29 Table 4-2 Stress Results in Compression and Tension........................... 31 Table 4-3 Summary of Transients Considered in the Fuel Rod Fatigue Analysis....................................................... 33 Table 4-4 Bounding Centerline Fuel Melt Limits................................. 45 Table 4-5 Bounding Transient Cladding Strain Limits............................ 48 Table 4-6 M5 Cladding Stress Intensity Limits................................. 52 Table 4-7 Summary of Fuel Assembly Damping Ratios........................... 52 Table 4-8 Peak Grid Impact Loads and Margins................................ 56 Table 4-9 Component Evaluation Margins..................................... 59 Table 5-1 Pressure Loss Coefficients Derived from Testing....................... 69 Table 6-1 Control Rod Design Parameters.................................... 71 Table 6-2 Control Rod Assembly Materials.................................... 71 Table 6-3 Control Rod Cladding Allowable Stresses............................. 76

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 List of Figures © Copyright 2024 by NuScale Power, LLC viii Figure 3-1 Fuel Assembly General Arrangement................................ 15 Figure 3-2 Representative Core Loading Pattern................................ 16 Figure 3-3 Top Nozzle..................................................... 17 Figure 3-4 Bottom Nozzle.................................................. 18 Figure 3-5 Guide Tube Assembly............................................ 19 Figure 3-6 Cap Screw Bottom Nozzle Connection............................... 20 Figure 3-7 Guide Tube Quick Disconnect Top Nozzle Connection................... 21 Figure 3-8 HTP' Grid..................................................... 22 Figure 3-9 HTP' Spacer Grid Characteristics.................................. 23 Figure 3-10 HMP' Spacer Grid.............................................. 24 Figure 3-11 Fuel Rod Assembly.............................................. 25 Figure 4-1 Predicted Corrosion Results........................................ 36 Figure 4-2 Generalized Cladding Stress Versus Time............................ 42 Figure 4-3 Cladding Deformation Rate Versus Time.............................. 43 Figure 4-4 Centerline Fuel Melt Bounding Envelopes for UO2 Fuel.................. 46 Figure 4-5 Centerline Fuel Melt and Transient Cladding Strain Bounding Envelopes for 8 wt% Gd2O3 Fuel............................................... 47 Figure 4-6 Transient Cladding Strain Linear Heat Generation Rate Limits for UO2 Fuel.......................................................... 49 Figure 4-7 Transient Cladding Strain Linear Heat Generation Rate Limits for 8 wt% Gd2O3 Fuel..................................................... 50 Figure 4-8 Best Estimate Axial Pressure Drop for Limiting Assembly................. 60 Figure 4-9 Internal Guide Tube Temperature for 100 Percent Power................. 62 Figure 4-10 Internal Guide Tube Temperature for 75 Percent Power.................. 63 Figure 4-11 Control Rod Position Versus Time................................... 64 Figure 4-12 Control Rod Velocity Versus Time................................... 65 Figure 6-1 Control Rod Assembly General Arrangement.......................... 72 Figure 6-2 Control Rod Assembly Cut-Away.................................... 73 Figure 6-3 Control Rod Design.............................................. 74

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 1 Abstract This report describes the NuFuel-HTP2' fuel and control rod assembly design, summarizes the key analysis results, and evaluates the fuel and control rod assembly design performance against regulatory requirements. The evaluations are focused on the mechanical aspects of the designs, consistent with Section 4.2 of NUREG-0800.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 2 Executive Summary This report describes the NuFuel-HTP2' fuel and control rod assembly (CRA) design, summarizes the key analysis results, and evaluates the fuel and CRA design performance against regulatory requirements. The fuel is designed and analyzed in accordance with regulatory requirements in NUREG-0800 Section 4.2 and the regulatory requirements from General Design Criterion (GDC) 2, GDC 10, GDC 27, Principal Design Criterion (PDC) 35, and 10 CFR 50.46. Fuel Assembly The fuel design is a reduced height 17x17 pressurized water reactor design. The assembly contains five spacer grids, 24 MONOBLOC' guide tubes and a top and bottom nozzle. The bottom nozzle contains a mesh debris filter. The top nozzle is removable to allow for reconstitution of fuel rods if needed. The top four grids provide structural support for the fuel assembly and enhance mixing of the coolant. The bottom grid is primarily for structural support. The fuel rod cladding is M5, an advanced zirconium alloy. The fuel is UO2 enriched up to 4.95 weight percent 235U, with gadolinia homogeneously mixed in some fuel pellets. The fuel assembly is analyzed using established design criteria and methods to demonstrate that the fuel assembly is not damaged during normal operation, anticipated operational occurrences, and postulated accidents. Potential failure mechanisms include mechanical loading, fatigue, fretting, oxidation, growth, and distortion. The fuel rod is analyzed using established design criteria and methods to demonstrate that the fuel rod is not damaged during normal operation, anticipated operational occurrences, and postulated accidents. Potential failure mechanisms include internal pressure, internal hydriding, creep collapse, centerline melting, pellet-cladding interaction, and mechanical loading. Certain design criteria (e.g., those associated with loss-of-coolant accidents) are addressed in other portions of the NuScale US460 Final Safety Analysis Report. The design criteria and methods have been applied to fuel assembly and rod designs currently in power operation and have been demonstrated to be applicable to the NuFuel-HTP2' design. The evaluations presented in this report demonstrate that the fuel rod design meets all applicable design criteria. Mechanical testing is performed to characterize the mechanical performance of the fuel assembly. Test results support the design evaluations presented in this report. Thermal-hydraulic evaluations and tests are performed to characterize the hydraulic performance of the fuel assembly and demonstrate that the fuel design meets all applicable design criteria. Evaluations and tests performed are comprehensive and demonstrate the acceptability of the fuel assembly design for use in the NuScale Power Module.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 3 Control Rod Assembly The CRA contains 24 individual control rods fastened to a stainless steel spider hub. The control rod tubes are stainless steel and contain silver indium cadmium (AIC) and boron carbide (B4C) neutron absorbers. The CRA design is analyzed using established design criteria and methods to demonstrate acceptable performance over the design lifetime. Potential failure mechanisms include stresses, strain, creep collapse, fatigue, wear, internal pressure, and component melting. Evaluations summarized in this report demonstrate that the CRA design meets all applicable criteria and is acceptable for use in the NuScale Power Module. Testing is performed to confirm CRA drop times and CRA drop velocity, and to assess the impact of the maximum expected fuel assembly distortion and the misalignment of the lead screw and guides.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 4 1.0 Introduction NuScale Power, LLC (NuScale) has developed a 17x17 fuel assembly, NuFuel-HTP2', and control rod assembly (CRA) design for use in the NuScale Power Module (NPM) based on existing Framatome methods and technology. 1.1 Purpose This report describes the NuFuel-HTP2' fuel assembly and CRA designs, summarizes the key analysis results, and demonstrates the designs meet regulatory requirements. 1.2 Scope This report uses Framatome methodologies to evaluate the NuFuel-HTP2' fuel design against the requirements of NUREG-0800 Section 4.2 (Reference 9.1.1). The evaluations are focused on the mechanical aspects of the design. Neutronic and thermal-hydraulic performance of the fuel assembly are addressed in the NuScale US460 Final Safety Analysis Report. This report also summarizes the CRA design and analyses. Chapter 2 provides the regulatory framework that identifies the requirements against which the fuel design is evaluated. Chapter 3 describes the mechanical design and function of the fuel assembly and provides a brief description of applicable operating experience. Chapter 4 summarizes the results of the fuel assembly design evaluations using NRC-approved methods.

The NuFuel-HTP2' fuel mechanical design is evaluated using generic mechanical design criteria for pressurized water reactor (PWR) fuel designs (Reference 9.1.2, Reference 9.1.3, and Reference 9.1.14).

The COPERNIC Fuel Rod Design Computer Code (Reference 9.1.4) is used in the fuel performance analysis of the NuFuel-HTP2' fuel design.

The CROV code (Reference 9.1.5) is used to evaluate the cladding creep performance.

The computational procedure for evaluating fuel rod bowing (Reference 9.1.6) is used to evaluate the impacts of fuel rod bow for the NuFuel-HTP2' design.

The PWR fuel assembly structural response to externally applied dynamic excitations (Reference 9.1.7) is used to perform seismic structural analysis on the fuel design. The applicability of the fuel analysis methods is demonstrated in Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P-A Revision 1 (Reference 9.1.8), NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-0716-50351-P-A Revision 1 (Reference 9.1.9), and as supplemented for the US460 design Framatome

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 5 Fuel and Structural Response Methodologies Applicability to NuScale, Supplement 1 to TR-0116-20825-P-A, Revision 1, Supplement 1 to TR-0716-50351-P-A, Revision 1, TR-108553-P-A (Reference 9.1.10). Chapter 5 describes the comprehensive testing performed to support the fuel assembly design and analysis. Chapter 6 describes the CRA design and the analyses and testing that support the design. Chapter 7 provides guidance for managing design changes to the fuel and CRA design. Criteria are provided to distinguish changes that require NRC approval. Chapter 8 provides a summary and concludes that the NuFuel-HTP2' fuel and CRA designs satisfy the applicable design requirements.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 6 Table 1-1 Abbreviations Term Definition AOO anticipated operational occurrence ASME American Society of Mechanical Engineers BOL beginning of life CFM centerline fuel melt CFR Code of Federal Regulations CRA control rod assembly CUF cumulative usage factor EFPY effective full-power year EOL end of life FCM fuel centerline melt FSAR Final Safety Analysis Report FIV flow-induced vibration GDC General Design Criteria LHR linear heat rate LOCA loss-of-coolant accident NPM NuScale Power Module NRC Nuclear Regulatory Commission OD outside diameter PDC Principal Design Criterion PHTF portable hydraulic test facility (Richland) PWR pressurized water reactor QD quick disconnect RCCA rod control cluster assembly RCS reactor coolant system RFT reactor flange tool RMS root mean square SRSS square root of the sum of the squares SSE safe shutdown earthquake TCS transient cladding strain

Background

This report supports FSAR Section 4.2. This section identifies the regulatory requirements that form the basis of the fuel design evaluation. 2.1 Regulatory Requirements The mechanical design of the NuFuel-HTP2' fuel assembly is evaluated against criteria established to be consistent with NUREG-0800 Section 4.2 (Reference 9.1.1). These criteria are specified in Section 4.0 of this report. Some of the NUREG-0800 acceptance criteria are addressed in plant transient analyses and are not addressed by this report, as noted in Section 4.0. The following regulatory requirements apply to the fuel mechanical evaluations summarized in this report:

General Design Criterion (GDC) 2, as it relates to designing fuel assemblies to withstand the effects of earthquakes

GDC 10, as it relates to assuring that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs)

GDC 27, as it relates to ensuring that fuel system damage is never so severe as to prevent control rod insertion when it is required

10 CFR 50.46, as it relates to ensuring that fuel assembly and fuel rod damage will not interfere with effective emergency core cooling Additionally, as specified in FSAR Section 3.1, principle design criterion (PDC) 35 applies to the fuel mechanical evaluations as it relates to ensuring fuel damage does not interfere with effective emergency core cooling. Section 4.0 identifies the approved methodologies used for evaluation of each criterion. These methodologies are demonstrated to be applicable to the design in Reference 9.1.8 and Reference 9.1.9 as supplemented for applicability to the US460 design in Reference 9.1.10.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 8 3.0 NuFuel-HTP2' Fuel Assembly Description The NuFuel-HTP2' fuel assembly (Figure 3-1) is a 17x17 PWR design that is approximately one-half the length of typical PWR nuclear plant fuel. Other than the shortened length, the assembly contains design features similar to those of proven HTP' fuel designs. All components of the fuel assembly have relevant operating experience that demonstrates their suitability for use in reactor cores. The assembly is supported by five spacer grids, 24 guide tubes, and a top and bottom nozzle that together provide the structural skeleton for the 264 fuel rods. The fuel rod consists of M5 alloy cladding and uranium dioxide (UO2) pellets with gadolinium oxide (Gd2O3) as a burnable absorber homogeneously mixed within the fuel pellets in select rod locations. Table 3-1 and Table 3-2 list key fuel design parameters. Table 3-3 provides representative operating conditions for the NPM. The core contains 37 fuel assemblies. Sixteen of the fuel assembly locations contain CRAs. Figure 3-2 provides a representative core loading pattern showing the arrangement of the fuel assemblies in the reactor core. The total, nominal height of the fuel assembly is 94 inches (excluding the hold-down spring height). Due to the assembly height and the use of span lengths between spacer grids that are typical for operating PWR plants, the assembly has a total of five spacer grids that provide lateral support for the fuel rods. Four HTP' grids at the intermediate and top spacer locations are welded to the guide tubes, while the HMP' lower grid at the bottom location of the fuel assembly is captured by rings welded to the guide tubes. The fuel assembly materials (Table 3-4) are chosen for their low cobalt content to reduce plant dose, for corrosion resistance, and for desirable structural properties. A summary of the NuFuel-HTP2' components is provided below. 3.1 Top Nozzle The top nozzle (Figure 3-3) consists of a stainless steel frame that interfaces with the reactor upper internals and the core components while providing for reactor coolant flow. The top nozzle flow hole pattern provides low pressure drop while satisfying strength requirements. A through-hole feature allows for insertion of the in-core instrumentation from above the fuel assembly. The top nozzle is attached to the fuel assembly with quick disconnect (QD) features at each of the 24 guide tube locations. The QD features allow for removal of the top nozzle for fuel assembly reconstitution. Two diagonally opposed corners of the top nozzle contain holes for accommodating the upper core plate alignment pins. Mounted on the other two corners are four two-leaf hold-down springs. The spring leaves maintain positive hold-down margin and are fastened to the top nozzle with clamp screws. The upper leaf has an extended tang that engages a cutout in the top plate of the nozzle to retain the spring leaves in the unlikely event of a leaf or clamp screw failure.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 9 On one of the corners with leaf springs is a through-hole that allows for identification of rotational orientation of the assembly from above. Additionally, on the top left side of each face of the nozzle are marks that allow for identification of the orientation of the assembly from the side. The hold-down spring system provides margin to fuel assembly lift off and helps to dampen axial motion of the fuel assembly caused by loss-of-coolant accident (LOCA) and seismic-induced core plate motions. 3.2 Bottom Nozzle with Mesh Filter Plate The stainless steel bottom nozzle (Figure 3-4) consists of a cast frame of ribs. Twenty-four holes allow for connecting the guide tubes to the nozzle using cap screws and a center hole is provided for the instrument tube. A high-strength A-286 alloy mesh filter plate is pinned to the top of the frame to provide debris resistance and is captured by the guide tubes and cap screws. The four corners have concave feet with a radius designed to interface with the alignment pins of the lower core plate. As with the top nozzle, the top left side of each face of the bottom nozzle contains marks to allow identification of assembly orientation from the side. The bottom nozzle height is designed as an anti-straddle feature to prevent the fuel assembly from being successfully set down over a shared fuel pin. This feature permits only one correct nozzle seating on the lower core plate. 3.3 Zircaloy-4 MONOBLOC' Guide Tubes The MONOBLOC' guide tubes (Figure 3-5) have a constant outer diameter. The upper portion of the guide tube has a large internal diameter that allows for rapid insertion of the CRA during a reactor trip. The lower portion of the guide tube has a reduced inner diameter that acts as a dashpot that decelerates the CRA to limit the impact forces on the fuel assembly during a reactor trip. The guide tube has four holes located just above the top of the dashpot to allow for cooling flow for inserted CRAs and outflow during a reactor trip. The outside diameter of the guide tube is constant. The added thickness in the dashpot of the MONOBLOC' guide tube increases the lateral stiffness of the fuel assembly and inhibits fuel assembly distortion and bow. The guide tube lower end plug (Figure 3-6) is threaded to accept a stainless steel cap screw to secure the guide tube to the bottom nozzle. The cap screw has a through-hole that allows for cooling flow to the guide tube as well as outflow during a reactor trip. The design of the dashpot and cap screw hole are consistent with existing PWR designs. The guide tube is connected to the top nozzle with a QD assembly (Figure 3-7). The QD consists of a double-spline sleeve made of Zircaloy-4 attached to the guide tube with multiple spot welds. Machined keyway-type features within the guide tube attachment holes in the top nozzle provide either clearance for removal or restraint for securing the nozzle, based on the radial orientation of the QD features.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 10 3.4 Zircaloy-4 Instrument Tube The Zircaloy-4 instrument tube has constant inner and outer diameters and is located at the center of the 17x17 array. It provides guidance for the in-core instrumentation, which is inserted from the top of the fuel assembly. The instrument tube is not attached to either of the fuel assembly nozzles, but has its axial position fixed by sleeves welded above and below the bottom HMP' grid. 3.5 Zircaloy-4 HTP' Upper and Intermediate Spacer Grids The four HTP' spacer grids (Figure 3-8 and Figure 3-9) that occupy the top four grid positions are formed from interlocking Zircaloy-4 strips that are welded at intersections to form a 17x17 matrix of square cells. Each grid strip includes a pair of strips welded back-to-back to create a doublet. The doublet is formed with flow channels that are angled at the outlets to create a swirling flow pattern. The flow channels are arranged so that there is no net hydraulic torque on the fuel assembly. The shape of the flow channel creates line contacts with the fuel rod that provide increased resistance to grid-to-rod fretting relative to traditional point-contact spacer grid designs. Sideplates are welded to the grid of doublets to complete the spacer grid design. The sideplates include lead-in tabs to eliminate hang-up during fuel movement. The HTP' spacer grids are spot welded to the guide tubes to limit axial movement and maintain alignment with adjacent fuel assemblies. The HTP' grids on the NuFuel-HTP2' fuel design are identical to those used on Framatomes 17x17 PWR product that has extensive operating experience in the United States. 3.6 Alloy 718 HMP' Lower Spacer Grid The HMP' spacer grid (Figure 3-10) resembles the HTP' spacer grid with respect to spring design, rod-to-grid surface contact and manufacturing. The HMP' spacer grid is made from low cobalt, precipitation-hardened Alloy 718 strip material, which provides enhanced strength and relaxation characteristics. The higher strength alloy allows thinner grid strips, resulting in reduced hydraulic resistance. The doublet flow channels are straight (non-mixing) flow channels that provide added reduction in hydraulic resistance. The grids are captured above and below by Zircaloy-4 sleeves spot welded to the guide tubes to maintain axial alignment. The HMP' grid on the NuFuel-HTP2' design is identical to Framatomes 17x17 PWR product that has extensive operating experience in the United States. 3.7 Fuel Rod with Alloy M5 Fuel Rod Cladding The fuel rod design (Figure 3-11) consists of ceramic UO2 pellets contained in seamless M5 zirconium alloy tubing with end caps welded at each end. The M5 cladding material significantly improves the resistance to corrosion compared to other cladding materials.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 11 M5 cladding material was first inserted in a U.S. reactor core in 1995. Twenty-two U.S. reactors have used M5 alloy in more than 7500 fuel assemblies. Globally, more than 5.8 million M5 fuel rods have operated in more than 23,000 fuel assemblies in 84 reactors. The operational experience of M5 cladding covers PWR fuel arrays from 14x14 up to 18x18. The fuel stack height is 78.74 inches and rests on top of the lower end cap. The fuel rod has an internal spring in the upper plenum that axially restricts the position of the fuel stack within the rod, preventing the formation of gaps during shipping and handling while allowing for the expansion of the fuel stack during operation. The void volume of the fuel rod is designed to accommodate fission gas generation during operation to maintain rod internal pressure less than reactor coolant system (RCS) pressure. The lower end cap has a bullet-nose shape to provide a smooth flow transition in addition to facilitating insertion of the rods into the spacer grids during assembly. The upper end cap has a grippable shape that allows for the removal of the fuel rods from the fuel assembly if necessary. The fuel pellet has chamfered edges and is dished on the top and bottom. The chamfers allow for ease of loading and reduce pellet chipping. The dishing and chamfers accommodate pellet swelling during operation and reduce the tendency to produce an hour-glass shape, reducing pellet-to-cladding stress concentrations and the potential for pellet stack gaps. The UO2 pellets have a theoretical density of 96.5 percent with a maximum enrichment up to 4.95 weight percent 235U consistent with current operating plant licensing requirements. The fuel rod design can utilize axial blanket and Gd2O3 fuel configurations.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 12 Table 3-1 Fuel Assembly Parameters Parameter Value Fuel rod array 17x17 Fuel rods per assembly 264 Guide tubes per assembly 24 Instrument tubes per assembly 1 Spacer grids per assembly 5 Fuel assembly height without holddown spring (inch) 94.0 Fuel rod pitch (inch) 0.496 Guide tube outside diameter (inch) 0.482 Guide tube inside diameter - upper region (inch) 0.450 Guide tube inside diameter - dashpot region (inch) 0.397 Instrument tube outside diameter (inch) 0.482 Instrument tube inside diameter (inch) 0.450 HTP' outer/inner strip height (inch) 1.950 / 1.750 HTP' outer/inner strip thickness (inch) [ ]ECI HMP' outer/inner strip height (inch) 1.950 / 1.750 HMP' outer/inner strip thickness (inch) [ ]ECI

Single strip thickness of a welded doublet

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 13 Table 3-2 Fuel Rod Parameters Parameter Value Cladding material M5 Fuel rod length (inch) 85.90 Length of total active fuel stack (inch) 78.74 Cladding outer diameter (inch) 0.374 Cladding inner diameter (inch) 0.326 Cladding inner surface roughness (µin) 45 Fuel rod internal pressure (psig) 215 Fuel rod fill gas Helium Fuel rod plenum height (inch) 6.21 Fuel pellet outer diameter (inch) 0.3195 Fuel pellet length (inch) 0.400 Fuel pellet surface roughness (µin) [ ]ECI Fuel pellet density (% TD) 96.5 Resinter densification limits (24-hour test) [ ]ECI Fuel pellet grain size (µm) [ ]ECI Fuel pellet open porosity fraction [ ]ECI Sorbed gas [ ]ECI Fuel pellet dish volume, nominal/percent (mm3) [ ]ECI Fuel pellet void volume, nominal/percent (cm3) [ ]ECI Plenum spring free length (inch) [ ]ECI Plenum spring outer diameter (inch) [ ]ECI Plenum spring wire diameter (inch) [ ]ECI Plenum spring active coils [ ]ECI Plenum spring volume (in3) [ ]ECI Lower end cap height (inch) 0.575 TD = theoretical density UTL = upper tolerance limit LCL = Lower confidence limit UCL = Upper confidence limit

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 14 Table 3-3 Comparison of Operating Conditions Parameter NPM Design Value Framatome 17x17 PWR Value Rated thermal power (MWt) 250 3455 Average coolant velocity (ft/s) 3.6 16 System pressure (psia) 2000 2280 Core average temperature (°F) 540 584 Core average linear heat rate (LHR), approx. (kW/m) 12.8 18.0 RCS inlet temperature (°F) 480 547 RCS Reynolds number 86,000 468,000 Fuel assemblies in core 37 193 Fuel assembly loading (kgU) 251 455 Core loading (kgU) 9,269 87,815 Table 3-4 Fuel Assembly Materials Component Material Top nozzle Stainless steel Bottom nozzle frame Stainless steel Mesh filter plate Alloy 286 Guide tube and QD sleeves Zr-4 Holddown leaf springs Alloy 718 Holddown spring clamp screw Alloy 718 Top connection (quick disconnect) Zr-4 and Alloy 718 Bottom cap screw AISI 316L stainless steel HMP' grid Alloy 718 HTP' grid Zr-4 Fuel rod cladding M5 - cold worked and recrystallized zirconium alloy Fuel rod plenum springs Alloy X-750 Fuel pellets UO2 and UO2 plus Gd2O3 Note: Stainless steels are low cobalt.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 16 Figure 3-2 Representative Core Loading Pattern B-02 B-01 A-01 B-01 A-02 A-01 A-02 B-01 B-01 C-02 B-02 A-01 A-01 C-03 A-01 A-01 B-02 B-01 A-02 A-01 A-02 B-01 C-02 B-01 A-01 B-01 B-02 A-01: Batch A Type 1, 4.50 wt% 235U, with Gadolinia A-02: Batch A Type 2, 4.50 wt% 235U B-01: Batch B Type 1, 4.50 wt% 235U, with Gadolinia B-02: Batch B Type 2, 4.50 wt% 235U C-01: Batch C Type 1, 4.50 wt% 235U, with Gadolinia C-02: Batch C Type 2, 4.50 wt% 235U C-03: Batch C Type 3, 2.65 wt% 235U A - Twice burned, B - Once burned, C - Fresh C-01 C-02 C-02 C-01 C-01 C-01 C-01 C-01 C-01 C-01

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NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 26 4.0 Design Evaluation This section evaluates the NuFuel-HTP2' design against criteria established consistent with Section 4.2 of NUREG-0800 (Reference 9.1.1), to provide assurance that

1. the fuel system is not damaged as a result of normal operation and AOOs.

2. the fuel system damage is never so severe as to prevent control rod insertion when it is required.

3. the number of fuel rod failures is not underestimated for postulated accidents.

4. core coolability is always maintained.

The design criteria are based on Framatome fuel design experience and are consistent with criteria established for previously approved fuel assembly designs. Reference 9.1.8 identifies the NRC-approved codes and methods used to evaluate the fuel performance. Table 2-3 of Reference 9.1.8 associates the method with the corresponding Reference 9.1.1 acceptance criteria. Reference 9.1.9 identifies the method of analysis for the structural response of the fuel assembly to dynamic faulted loads. Reference 9.1.10 confirms the applicability of these methods for use with the SDAA. Use of well-established design criteria and evaluation methods provides assurance of acceptable fuel performance. The results of the fuel performance analyses are applicable for operation in the NPM. 4.1 Fuel System Damage Criteria 4.1.1 Stress and Loading Limits Design Criteria Stress intensities for fuel assembly components shall be less than the stress limits based on ASME Code, Section III criteria (Reference 9.1.2). Buckling of the guide tubes shall not occur during normal operation and AOOs. The cumulative number of strain fatigue cycles on the structural components shall be less than the design fatigue lifetime. NuFuel-HTP2' Design Evaluation In the normal operating analysis, a series of mechanical analyses demonstrate the fuel assembly can withstand stresses and buckling loads from start-up, steady-state operation, shutdown, and AOOs. Each structural component in the fuel assembly is evaluated against the ASME Code Section III, Level A service limits (Reference 9.1.11). The fuel assembly weight, hold-down spring forces, RCS hydraulic loads, thermal loads during plant heat-up, steady-state operation and cooldown, and CRA drop loads (sum of the CRA spring preload and the load from the

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 27 maximum travel of the CRA spring retainer) create the stress states that are evaluated in the analysis. The operating basis earthquake is less than one-third of the safe shutdown earthquake (SSE) ground motion and is enveloped by the SSE analysis. The fatigue analysis evaluates cyclic loading due to normal operation and AOOs combined with the OBE, for a total of 137 transients over the life of the fuel. The NUREG-0800 criterion for a safety factor of 2 on stress amplitude or 20 on the number of cycles is satisfied by the use of the ODonnell-Langer curve (Reference 9.1.12) in the analysis. Guide tube normal operating loads are evaluated for each fuel assembly span. Guide tube spans are separated by spacer grids that begin above the bottom nozzle and extend to the guide tube span below the top nozzle. The loads and material properties of the guide tubes and the other fuel assembly components are based on the bounding outlet temperature in order to conservatively evaluate the loads, design limit, and minimum design margins for fuel assembly components at normal operating conditions. The guide tube buckling analysis evaluates a maximum guide tube eccentricity to create bounding load predictions when the fuel assembly weight, hold-down spring force, and CRA drop loads are simultaneously applied. Guide tube stress calculations consider axial loading conditions that cause tensile stresses. Hydraulic loading is conservatively ignored because it reduces the CRA drop and weight loads. Secondary loads are also considered in the form of spacer grid slip loads. These loads are generated as fuel rod slip is resisted by the grids because of differential thermal expansion and irradiation growth between the fuel rods and guide tubes during normal operation. These frictional loads are evaluated for a condition where rods are unseated, or lifted, at beginning of life (BOL) and are conservative for the seated condition at end of life (EOL). Guide tube primary membrane (Pm), primary membrane + bending (Pm + Pb), and primary + secondary (Pm + Pb + Q) stresses are evaluated against allowable stresses using the ASME Code Level A service limits based on the material yield and ultimate strengths. The guide tube upper sleeve strength is bounded by the strength of the weld connections with the guide tubes. The guide tube upper sleeve seating appendages are also evaluated for bearing stress and shear stress, considering the loads applied through the top nozzle during a CRA drop. For welded or threaded joints and various structural connections in the fuel assembly, the maximum load during normal operation is evaluated against ASME Code Level A service limits. The evaluated connections include the guide tube to spacer grid welds, the guide tube to upper sleeve welds, the guide tube upper sleeve to QD retainer weld, the guide tube to guide tube lower end fitting (GTLEF) weld, and the threaded connection between the GTLEF and the shoulder screw. The threaded connection is evaluated using ASME Code methods.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 28 The bottom nozzle and top nozzle strengths are evaluated considering the maximum operating loads from the fuel assembly weight, hold-down spring force, and CRA drop events and the allowable ASME Code load limit based on prototype component tests. For both the bottom nozzle and top nozzle, testing determined the collapse load for each structural framework. In accordance with the ASME Code, each tested collapse load is multiplied by two-thirds in establishing the allowable load limit. In consideration of corrosion effects, guide tube wall thinning due to corrosion is considered within EOL condition evaluations. This corrosion allowance is applied as a reduction to the inner and outer surface of the guide tube material. Similar wall thinning is not applied to stainless steel components given the minimal levels of corrosion observed for those components in reactor operating conditions. The hold-down spring analysis shows that the spring meets ASME Code limits for stress and strain. Spring stresses do not remain within elastic limits. Rather, the springs shake down to elastic action. Following shakedown, they behave elastically up to a newly established yield point. Table 4-1 provides a summary of the results for normal operation of the NuFuel-HTP2' fuel assembly design. The results presented in Table 4-1 demonstrate that the assembly components meet the acceptance criteria.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 29 Consistent with Reference 9.1.2, the fuel handling qualification includes a 2.5g axial load. The shipping qualification includes 4g axial and 6g lateral loads and grid clamp loads. Loads for shipping and pre-receipt handling are evaluated for fresh fuel conditions. The maximum stresses and loads on the fuel assembly components and structural connections during shipping and handling remain below the specified Table 4-1 Summary of Results - Fuel Assembly Design Margins [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 30 minimum strength, critical buckling loads, and below the ASME Code Level A service limits. For a 4g axial acceleration, the fuel rod plenum spring maintains a force against the fuel stack sufficient to prevent column movement during handling. The evaluation also demonstrates that the fuel rods do not slip through the spacer grids under 4g axial loads. 4.1.1.1 Fuel Rod Cladding Stress and Buckling Design Criterion Fuel rod cladding stress shall not exceed the following stress limits defined in Reference 9.1.3:

[ ]

[ ] The fuel rod shall not buckle based on [ ] criterion during the limiting overpressure transient at BOL. NuFuel-HTP2' Design Evaluation The fuel rod stress and buckling analysis determines the in-core steady-state stress and buckling performance of the fuel rod design. [ ] Pressure and temperature inputs are chosen so that operating conditions for normal operation and AOOs are enveloped. The stress analysis takes into account several sources of cladding stress: pressure differentials, ovality, thermal differentials, flow-induced vibration (FIV), fuel rod growth, and fuel rod to spacer grid (FR-SG) interaction. The following four stress categories are analyzed:

Primary Membrane (Pm) - Pressure stresses

Pm + Primary Bending (Pb) - Pressure, ovality and FIV stresses

Pm + Pb + Local - Pressure, ovality, FIV, and FR-SG stresses

Pm + Pb + Local + Secondary - Pressure, ovality, FIV, FR-SG, growth, and thermal stresses

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 31 At both the inner and outer diameter of the fuel rod, the maximum value of each individual stress is determined. The individual stresses within each stress component (tangential, axial, and radial) are added to find a maximum and minimum stress value. The stress intensity for each category is determined by combining the maximum and minimum stresses. The stress intensity is compared with the allowable stress to determine the margin for the particular stress. The cladding stress results, listed in Table 4-2, show positive margins for all stress categories. The minimum margin occurs on the cladding outer diameter in compression when combining primary membrane + bending + local stresses. The buckling pressure is calculated to be [ ] psi. The buckling pressure is higher than the maximum BOL pressure the fuel rod would experience during an overpressure event (2015 psi). Therefore, the design meets the buckling criterion. The calculated critical bucking load is [ ] lbf. The critical bucking load is greater than the total compressive load of [ ] lbf. Therefore, the Euler buckling criterion is also met. Table 4-2 Stress Results in Compression and Tension [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 32 4.1.2 Cladding Fatigue Design Criterion The fuel rod cumulative usage factor shall not exceed 0.9 (Reference 9.1.3). NuFuel-HTP2' Design Evaluation A bounding analysis of the core design is performed using the COPERNIC fuel rod analysis code (Reference 9.1.4) for both UO2 and UO2-Gd2O3 rods. Plant operations result in fluctuating thermal, pressure, ovality, and pellet-clad contact stresses in the fuel rod cladding. The COPERNIC code predicts changes in cladding diameter, cladding temperature, and fuel rod internal pressure at each time step. These parameters are used to calculate the various stresses used in the fatigue calculation. The transients considered in the fatigue analysis are provided in Table 4-3. The fuel rod life is conservatively assumed to be 10 years. With this assumption, the fuel rod experiences one-sixth of the number of transients identified in Table 4-3 for a sixty year plant design life. The fuel rod behavior during each transient is analyzed using the COPERNIC fuel rod code, which predicts changes in cladding diameter and temperature and the fuel rod internal pressure for each time step. These parameters are used to calculate the various stresses used in the fatigue evaluation. The maximum cumulative usage factor (CUF) for UO2 fuel rods is [ ] and the maximum CUF for UO2-Gd2O3 rods is [ ]. Both of these CUFs are well below the limit of 0.9.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 33 Table 4-3 Summary of Transients Considered in the Fuel Rod Fatigue Analysis Transient Identifier Design Transient Name Events for 60 Year Design Life Condition I Events (Service Level A) A01-HTS Reactor Heatup to Hot Shutdown 200 A02-RCD Reactor Cooldown from Hot Shutdown 200 A03-PAC Power Ascent from Hot Shutdown 700 A04-PWD Power Descent to Hot Shutdown 300 A05-LFW Load Following 19,750 A06-REG Load Regulation 767,100 A07-SSF Steady State Fluctuations 5,000,000 A08-RLI Load Ramp Increase 2000 A09-RLD Load Ramp Decrease 2000 A10-SLI Step Load Increase 3000 A11-SLD Step Load Decrease 3000 A12-LLD Large Step Load Decrease 200 A13-REF Refueling 60 A14-MKP Reactor Coolant System Makeup 175,200 A15-SGI Steam Generator Inventory Control from Hot Shutdown 600 A16-HPD High Point Degasification 440 A17-CEV Containment Evacuation 66,000 A18-CFD Containment Flooding and Drain 400 A19-SLT Secondary Leakage Tests 200 A20-HFT Initial Test Program 20 Condition II Events (Service Level B) B01-FWD Decrease in Feedwater Temperature 180 B02-FWI Increase in Secondary Flow 30 B03-TTX Turbine Trip without Bypass 90 B04-TTB Turbine Trip with Bypass 180 B05-LOP Loss of Normal Alternating Current Power 60 B06-SVC Inadvertent Main Steam Isolation Valve Closure 30 B07-IOD Inadvertent Operation of the Decay Heat Removal System 15 B08-TRP Reactor Trip from Full Power 125 B09-DRP Control Rod Misoperation 60 B10-PZM Inadvertent Pressurizer Spray 15 B11-COP Cold Overpressure Protection 30 B12-CMT Chemical and Volume Control System Malfunctions 30 Condition III Events (Service Level C) (only one of the following is considered) C01-VNT Spurious Emergency Core Cooling System (ECCS) Valve Actuation 5 C02-BDN Inadvertent Opening of a Reactor Safety Valve 5 C03-SML Chemical and Volume Control System Pipe Break 5 C04-SGT Steam Generator Tube Failure 5 D01-SLB Steam Piping Failures 1 D02-FLB Feedwater Piping Failures 1 D03-CRE CRA Ejection 1 T01-PHT Primary Hydrostatic Test 10

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 34 4.1.3 Fretting Design Criterion Fuel rod failures due to fretting shall not occur, as verified by fretting tests (Reference 9.1.2). NuFuel-HTP2' Design Evaluation Fretting and vibration performance is validated by the 1000-hour life and wear test performed at the Richland portable hydraulic test facility (PHTF), in addition to other relevant HTP' vibration and fretting tests.

The 1000-hour life and wear test performed at the Richland PHTF was run for 1032 hours at a temperature of 300 degrees F at or above the target Reynolds number of 52,000. At the conclusion of the test, [ ] fuel rods were examined for grid-to-rod fretting performance. The test results show no wear abnormalities with results [ ] well within the performance base for historical test results of proven in-reactor designs. The fretting results are based on a conservative flow configuration with a test-to-reactor momentum flux ratio of approximately [ ]. The test assembly replicated the EOL condition [ ] The HMP' grid for the EOL test assembly was relaxed to [ ] of the unirradiated grid-to-rod support.

The predicted vibration response amplitude for the NuFuel-HTP2' design is [ ], which is less than the maximum measured rod amplitude of [ ] in the Hermes-T vibration and wear test performed for the 17x17 HTP' fuel transition, which considered the effects of bundle-to-bundle cross flow.

Fretting results for autoclave testing (8005 hours) of the 17x17 HTP' grid design, which is identical to the HTP' grid used on the NuFuel-HTP2' design, show low wear [ ] for conservative imposed vibration amplitudes [ ].

Fretting results for autoclave testing (1000 hours) of the Advanced W17 HTP' design with intermediate flow mixers, where the HTP' and HMP' grids are similar to the NuFuel-HTP2' design, show low wear [ ] for conservative imposed vibration amplitudes and [ ]. T02-SHT Secondary Hydrostatic Test 10 T03-CHT Containment Hydrostatic Test 10 Table 4-3 Summary of Transients Considered in the Fuel Rod Fatigue Analysis (Continued) Transient Identifier Design Transient Name Events for 60 Year Design Life

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 35 The predicted small vibration amplitudes for the NuFuel-HTP2' design are a result of the lower NPM axial and cross flow velocities relative to those of a typical PWR. The NPM has a nominal flow rate of 395 gpm per fuel assembly compared to a best-estimate flow of approximately 2050 gpm per fuel assembly for a forced circulation PWR using a 17x17 fuel assembly design. The maximum calculated local cross flow velocity for NuFuel-HTP2' fuel assemblies is (( }}2(a),(c),ECI ft/sec for a uniform core compared to approximately [ ] ft/sec for a conventional PWR using a 17x17 fuel assembly design. The robust fretting characteristics of the NuFuel-HTP2' design provide confidence in the FIV performance of the fuel. The grid design provides line contact between the fuel rods and the spacer grid with large contact surfaces to mitigate wear. The grid-to-rod support also provides for higher damping to help suppress FIV and fretting wear. The grid-to-rod support conditions in the NuFuel-HTP2' design are similar to the grid-to-rod support of other HTP' fuel assemblies, of which more than 18,000 have been introduced into operating reactors globally. Thus, the HTP' test results and operating experience pertaining to rod vibration and fretting is applicable for evaluating the NuFuel-HTP2' design. Based on the minimal fretting wear measured during the life and wear test in the Richland PHTF, the small predicted rod vibration amplitudes, and the extensive favorable operating and test experience with the HTP' fuel design, the NuFuel-HTP2' design is not expected to experience FIV or wear issues in the NPM, and fuel rod failures due to fretting do not occur. There are no limitations with respect to time or burnup for the conditions evaluated. 4.1.4 Oxidation, Hydriding, and Crud Buildup Design Criterion The fuel rod cladding peak oxide thickness shall not exceed a best-estimate predicted value of 100 microns. Hydrogen pickup is controlled by the corrosion limit. Crud buildup is limited by inclusion in the oxidation measurement (Reference 9.1.4). NuFuel-HTP2' Design Evaluation A bounding analysis is performed using the COPERNIC fuel rod analysis code (Reference 9.1.4) for UO2 rods with and without Gd2O3. The corrosion of the fuel rods is modeled in order to calculate the oxide thickness that develops on the outer surface of the rods during operation. A bounding input power history, expressed in terms of effective full-power hours is used that bounds the individual rod power histories of the UO2 and Gd2O3 rods in the equilibrium cores. The corrosion analysis primarily depends on the amount of energy transfer through the cladding and the irradiation time. It shows little sensitivity to the fuel rod design characteristics inside the rods. The use of a bounding power history envelope makes the analysis equally applicable to all fuel rod types.

] }}2(a),(c),ECI micrometers (Figure 4-1), which is well below the design limit of 100 micrometers.

Framatome and industry operating experience show that crevice corrosion is not a current cause of PWR fuel failures. Framatome PWR fuel designs and materials have shown no susceptibility to fuel reliability concerns driven by crevice corrosion. Specific to the M5 cladding, there are no significant crevices on the fuel cladding surface that could shield an area from the RCS flow, so there is no significant risk of crevice corrosion for the M5 fuel rod cladding. 4.1.5 Fuel Rod Bow Design Criterion There is no specific design criterion for fuel rod bow. Fuel rod bowing is evaluated with respect to the mechanical and thermal-hydraulic performance of the fuel assembly (Reference 9.1.6). NuFuel-HTP2' Design Evaluation Fuel rod bow is the deviation from straightness of the fuel rods in the fuel assembly. The presence of fuel rod bow is identified by the deviation in water channel gap from nominal conditions. The primary effects of rod bow are a decrease in the critical heat flux ratio and an increase in local power peaking. Secondary effects of fuel rod bow Figure 4-1 Predicted Corrosion Results (( [

] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 37 can include fuel clad fretting at 100 percent gap closure, although the probability of rod-to-rod contact is minimal. The NuFuel-HTP2' design does not introduce changes from current Framatome PWR fuel assembly designs that might adversely impact rod bow. Operating plant rod bow data for current Framatome PWR designs continue to be adequately covered by the existing rod bow correlation methodology. The NuFuel-HTP2' design is within the current experience base with regards to fuel rod bending stiffness and operating temperature and is less limiting regarding end grid slip loads and span length. The mechanical rod bow analysis concludes that fuel assembly performance is within current models and experience. Rod bow penalties are derived for both linear heat generation rate (LHGR) and critical heat flux based on the NRC-approved methodology for quantifying fuel rod bowing. 4.1.6 Axial Growth Design Criteria For the fuel assembly, the axial clearance between core plates and the top and bottom nozzles shall allow sufficient margin for fuel assembly growth during the assembly lifetime. For the fuel rod, adequate clearance shall be maintained between the fuel rod and the top and bottom nozzles to accommodate the differences in the growth of fuel rods and the growth of the fuel assembly (Reference 9.1.2). NuFuel-HTP2' Design Evaluation The fuel assembly and its components grow during operation. There are two components of the growth: thermal expansion and irradiation growth. The minimum clearance between the fuel rods and the top and bottom nozzles and the clearance between the fuel assembly and core plates at the EOL condition are determined using worst case fuel rod and fuel assembly growth models and worst case initial dimensions. [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 38 [ ] The analysis shows that the allowable average assembly fast fluence is (( }}2(a),(c),ECI n/cm2 and the allowable maximum average fuel rod fast fluence is (( }}2(a),(c),ECI n/cm2. The minimum core plate to fuel assembly gap is [ ] inches at the allowable assembly fast fluence. The minimum fuel rod shoulder gap is [ ] inches (at a fuel assembly fluence of (( }}2(a),(c),ECI n/cm2 during hot operation). A separate growth-based analysis was performed to demonstrate that adjacent fuel assemblies would maintain adequate spacer grid alignment given differing burnup profiles. The analysis applied a burnup differential of [ ] GWd/mtU between adjacent fuel assemblies. The recommended allowable offset based on typical fuel assembly design requirements is that flat vertical surfaces of spacers on adjacent assemblies maintain a positive overlap at the worst-case conditions. The analysis demonstrates a maximum offset of [ ] inch, which maintains a positive overlap for the spacer grids that have a height of 1.750 inches. 4.1.7 Fuel Assembly Distortion Evaluation The NuFuel-HTP2' fuel has features, including spacer grids, structural connections, and guide tube diameters, similar to current Framatome 17x17 fuel but with a shorter overall length. The shorter length increases the lateral stiffness of the fuel assembly. As validation, fuel assembly lateral stiffness tests were performed for the NuFuel-HTP2' fuel design in-air at BOL and EOL conditions. Similar lateral stiffness tests have been conducted for current Framatome 17x17 fuel designs at EOL and BOL conditions. The test results show that the fuel assembly lateral stiffness is more than [ ] times greater than that of the current Framatome 17x17 fuel design. With this level of lateral stiffness and significantly lower hydraulic loads on the fuel assembly due to natural circulation flow, the fuel assembly has a high level of resistance to fuel assembly distortion. Differential fuel rod and guide tube growth rates, coupled with spacer grid slip loads, can contribute to fuel assembly distortion during operation. The fuel design has the same structural components and fuel rod cladding diameters and material (producing similar slip loads and growth rates) as the current Framatome 17x17 PWR designs; thus, the guide tube stresses from fuel rod and guide tube differential growth are bounded by Framatome design experience. The differential growth stresses imparted to the guide tubes are reduced compared to Framatome experience because of the reduced number of spacer grids over which tensile loads may accumulate on the guide tubes and the reduced length of the fuel rods and guide tubes, which results in lower differences in growth. These design characteristics ensure that the fuel rod and guide tube growth differential effects are within Framatomes recent PWR 17x17 fuel design experience.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 39 Operating experience of the current Framatome 17x17 fuel assembly design demonstrates little in-reactor fuel distortion as evidenced by the absence of incomplete rod insertions and slow-to-settle observations where full insertion of the control rod is delayed. Therefore, control rod drop concerns related to assembly distortion are not expected for the NuFuel-HTP2' fuel assembly design. 4.1.8 Fuel Rod Internal Pressure Design Criterion The internal gas pressure of the peak fuel rod in the reactor shall remain below a value that would cause the fuel-cladding gap to increase because of outward cladding creep during steady-state operation and extensive departure from nucleate boiling (critical heat flux) propagation to occur (Reference 9.1.4). NuFuel-HTP2' Design Evaluation A bounding analysis is performed using the COPERNIC fuel rod analysis code (Reference 9.1.4) for UO2 rods with and without Gd2O3. The maximum fuel rod internal pressure is conservatively compared to a design limit equal to nominal system pressure (2000 psia). Meeting this criterion demonstrates that the fuel-clad gap does not increase due to cladding outward creep during steady-state operation because a greater pressure external to the rod prevents outward creep and fuel-clad liftoff. This criterion also ensures that extensive departure from nucleate boiling propagation does not occur. The COPERNIC pressure calculation is based on a best-estimate prediction plus an uncertainty allowance to take into account code uncertainties and manufacturing variations. The analysis considers steady-state and Condition I (normal operation) and Condition II (AOO) transients over the full burnup range. The transients are modeled in the COPERNIC input with appropriate axial flux shapes. The maximum calculated internal pressure over the burnup history (for both UO2 and Gd2O3 rods) is (( }}2(a),(c),ECI psia compared to a limit of 2000 psia. 4.1.9 Assembly Liftoff Design Criterion The fuel assembly shall not lift off from the lower core plate under normal operating conditions and AOOs (Reference 9.1.2). NuFuel-HTP2' Design Evaluation The fuel assembly lift-off analysis evaluates an (( }}2(a)(c) AOO. To bound conditions for this event, the maximum flow rates at each corresponding power level are (( }}2(a)(c). The maximum hydraulic lift is

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 40 [ ] lbf, resulting in a lift margin of [ ] lbf. There are large margins against lift-off at all normal operating, startup, and transient (AOO) conditions. Because the NPM relies on natural circulation of the coolant without mechanical pumps, the assumed flow rates envelope all operating conditions and AOOs. 4.2 Fuel Rod Failure Criteria 4.2.1 Internal Hydriding Design Criterion The fabrication limit for total hydrogen inside a fuel rod assembly is maintained at a minimal level to limit internal hydriding (Reference 9.1.2). NuFuel-HTP2' Design Evaluation Fuel rod internal hydriding is controlled by fabrication limits for fuel pellet moisture. These controls, typical for Framatome fuel manufacturing, limit the total hydrogen content, including moisture, to [ ] ppm by weight, before rod final closure welding. 4.2.2 Cladding Collapse Design Criterion The predicted creep collapse life of the fuel rod shall exceed the maximum expected incore life (Reference 9.1.5 and Reference 9.1.3). NuFuel-HTP2' Design Evaluation The cladding creep collapse analysis is performed using the methodology of Reference 9.1.5, extended to M5 applications in Reference 9.1.3. A bounding analysis of the core design is performed using the COPERNIC fuel rod analysis code and the CROV creep ovalization code for both UO2 and Gd2O3 rods. Reference 9.1.5 establishes the three collapse criteria to be analyzed:

bifurcation buckling pressure

yield stress

deformation rate COPERNIC simulates the performance of the fuel rod throughout the lifetime of the rod to generate the parameters required to perform the creep collapse analysis:

rod internal pressure

interior and exterior cladding temperatures

coolant temperature and fast flux

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 41 The CROV code applies the initialization parameters from COPERNIC along with the fuel rod geometry (i.e., outside diameter, wall thickness, and ovality) to simulate the cladding creep-down deformations versus time. Consistent with Reference 9.1.5, when the ovality creep rate of the cladding exceeds 0.1 mils/hr, or the generalized stress within the cladding exceeds the yield stress, the cladding is considered to have failed. In addition, the bifurcation buckling pressure must not be exceeded. The three collapse criteria determine the predicted creep collapse life of the fuel, which must exceed the maximum expected incore life. The CROV analysis demonstrates that the bifurcation buckling pressure limit is not exceeded. A bifurcation buckling pressure limit of approximately (( }}2(a),(c),ECI psi is calculated that approximates the limit from the CROV analysis. The CROV analysis does not explicitly calculate the margin to the limit; however, the maximum pressure differential applied to the fuel rod in the CROV runs occurs at BOL and is approximately (( }}2(a),(c),ECI psi, which indicates that the bifurcation buckling pressure collapse criterion is satisfied with significant margin. Figure 4-2 and Figure 4-3 demonstrate that the fuel rod does not collapse as a result of stress beyond the yield point over the expected three cycle incore life.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 43 All three collapse criteria are met. Therefore, the predicted creep collapse life of the fuel rod exceeds the maximum expected incore life of the fuel rod and cladding creep collapse does not occur. 4.2.3 Overheating of Cladding Overheating of cladding is evaluated in the FSAR Chapter 15 transient analyses and is not addressed in this report. Figure 4-3 Cladding Deformation Rate Versus Time (( }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 44 4.2.4 Overheating of Fuel Pellets Design Criterion Fuel melting during normal operation and AOOs shall be precluded (Reference 9.1.4). NuFuel-HTP2' Design Evaluation The fuel centerline melt (FCM) analysis is performed using the methodology of Reference 9.1.4. A bounding analysis of the core design is performed using the COPERNIC fuel rod analysis code for both UO2 and UO2-Gd2O3 rods. The COPERNIC code predicts the transient LHRs where the onset of FCM occurs. [ ] [ ] Using the calculated LHR limits for each of the fuel rods, a bounding envelope is created, as shown in Table 4-4. Figure 4-4 and Figure 4-5 show the bounding FCM LHR limits for each fuel rod type. (( [ ] }}2(a),(c),ECI

] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 47 4.2.5 Excessive Fuel Enthalpy Excessive fuel enthalpy from a reactivity initiated accident is addressed in FSAR Chapter 15 analyses. 4.2.6 Pellet-Cladding Interaction Design Criteria As stated in NUREG-0800 Section 4.2, there is no generic criterion for fuel failure resulting from pellet-cladding interaction or pellet-cladding mechanical interaction. Cladding strain and fuel melt criteria are applied as a surrogate. The maximum uniform hoop strain (elastic plus plastic) shall not exceed 1 percent. Steady-state creep-down and irradiation growth are excluded (Reference 9.1.4). The fuel melt criterion is stated in Section 4.2.4. Figure 4-5 Centerline Fuel Melt and Transient Cladding Strain Bounding Envelopes for 8 wt% Gd2O3 Fuel (( [ ] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 48 NuFuel-HTP2' Design Evaluation The transient cladding strain analysis is performed using the methodology of Reference 9.1.4. A bounding analysis of the core design is performed using the COPERNIC fuel rod analysis code for both UO2 and UO2-Gd2O3 rods. The COPERNIC code predicts the transient LHRs where the cladding uniform hoop strain equals 1 percent. [ ] Using the calculated LHR limits for each of the fuel rods, a bounding envelope is created, as shown in Table 4-5. Figure 4-6 and Figure 4-7 show the transient cladding strain LHR limits for each fuel rod type. Table 4-5 Bounding Transient Cladding Strain Limits (( [ ] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 49 Figure 4-6 Transient Cladding Strain Linear Heat Generation Rate Limits for UO2 Fuel (( [ ] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 50 4.2.7 Bursting Swelling and rupture of the cladding relates to the ECCS performance evaluation in FSAR Chapter 15 analyses and is not addressed in this report. 4.2.8 Mechanical Fracturing The seismic and LOCA load analysis summarized in Section 4.3.5 addresses externally applied forces on the fuel rod. 4.3 Fuel Coolability 4.3.1 Cladding Embrittlement Cladding embrittlement relates to the ECCS performance evaluation and is not addressed in this report. 4.3.2 Violent Expulsion of Fuel Because reactivity initiated accidents are addressed in FSAR Chapter 15 analyses, they are not addressed in this report. Figure 4-7 Transient Cladding Strain Linear Heat Generation Rate Limits for 8 wt% Gd2O3 Fuel (( [ ] }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 51 4.3.3 Generalized Cladding Melting As stated in NUREG-0800 Section 4.2, criteria for cladding embrittlement in Section 4.3.1 are more stringent than generalized cladding melting criteria. Therefore, additional specific criteria are not used. 4.3.4 Fuel Rod Ballooning Burst strain and flow blockage caused by ballooning of the cladding relates to the ECCS performance evaluation and is not addressed in this report. 4.3.5 Fuel Assembly Structural Damage from External Forces Design Criterion The fuel assembly shall withstand the loads from an SSE and a LOCA. Specific acceptance criteria for fuel assembly components are identified in Reference 9.1.7. NuFuel-HTP2' Design Evaluation The external load analysis is performed using the methodology of Reference 9.1.7 as modified by Reference 9.1.10. The analysis demonstrates positive margins to all applicable criteria. Cladding Faulted The faulted limits applied to the fuel rod cladding are shown in the far right column of Table 4-6. The table also shows the limits used in the previous faulted analysis that were based on Table 3-3 of the Framatome M5 Topical report (Reference 9.1.3). The middle column of values is derived from Table XIII-3110-1 of Reference 9.1.13. They are simplified for application to the cladding allowable stresses. The applied M5 cladding stress intensity limits are adopted from Revision 2 of the Framatome M5 Topical report (Reference 9.1.14).

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 52 4.3.5.1 Analysis Inputs 4.3.5.1.1 Lateral Model The lateral fuel assembly model is developed using the model parameters from Section 6.1 of Reference 9.1.7. Damping coefficients specific to the NuFuel-HTP2' design are summarized in Table 4-7 and defined in Reference 9.1.9. The analysis uses values that are more conservative than those listed in Table 4-7. Table 4-6 M5 Cladding Stress Intensity Limits [ ] Table 4-7 Summary of Fuel Assembly Damping Ratios [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 53 The following model parameters are established through design-specific characterization testing:

[ ] [ ] [ ]

[ ] [ ] [ ] [ ] The application of the free and forced vibration test data from prototypical BOL and EOL NuFuel-HTP2' fuel assemblies to define the fuel assembly dynamic characteristics is described in Appendix A of Reference 9.1.9. Dynamic crush testing was performed on prototypical NuFuel-HTP2' spacer grids in both a non-irradiated (BOL) and simulated-irradiated (EOL) condition to define the external spacer grid stiffness and damping characteristics. Grid corrosion is not directly modeled because the effect of oxidation on spacer grids is a small increase in strength and not accounting for it is conservative. The [ ] is demonstrated in Appendix A of Reference 9.1.9. The dynamic crush test is also used to establish a grid load limit. [ ] The grid loads are reported in Table 4-8. Table 4-7 Summary of Fuel Assembly Damping Ratios (Continued) [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 54 Fuel assembly lateral impact testing was performed on prototypical NuFuel-HTP2' fuel assemblies in both a BOL and EOL condition to establish the internal stiffness and damping parameters for the spacer grid. 4.3.5.1.2 Vertical Model The vertical fuel assembly model is developed using the model parameters from Section 6.2 of Reference 9.1.7. The following model parameters are established through design-specific characterization testing:

[ ] [ ] [ ]

[ ] [ ] [ ] [ ] Spacer grid slip load tests define the grid-to-fuel rod slider slip load at BOL conditions. In this test, a prototypical NuFuel-HTP2' spacer grid is loaded with cladding segments and a uniform load is applied across all of the cladding segments. The load at which the fuel rods are observed to begin slipping through the grid is identified as the global slip load. To simulate the EOL condition, [ ] The Alloy 718 lower end grid is [ ] An axial stiffness test was performed on full-scale prototypical NuFuel-HTP2' fuel assemblies in both the BOL and EOL condition. The axial stiffness of the fuel assembly is measured at key axial locations (e.g., spacer grids and top nozzle). The measurements of location-specific axial stiffness are used to benchmark the stiffness of the grid-to-fuel rod slider elements in the vertical model. In the event of a fuel assembly drop, two impact mechanisms require characterization. The nozzle-to-core plate gap stiffness and damping are established by a dynamic drop test of the fuel assembly. In this test, full-scale prototypical NuFuel-HTP2' fuel assemblies in both the BOL and EOL condition were dropped onto a rigid surface from varying heights. Impact loads, assembly velocity, and assembly position were recorded. [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 55 [ ] The fuel rod-to-nozzle gap element is defined using dynamic drop test results performed on the EOL fuel assembly. In this test, [ ] [ ] 4.3.5.1.3 Excitation Inputs The excitation inputs for the external load analysis are NPM core plate displacement time histories for the SSE and LOCA events. The core plate displacement time histories include both horizontal and vertical motions. The SSE input motions are the result of an evaluation of four soil types. For each soil type, all 6 modules are analyzed. To account for the effect of uncertainty in the reactor module dynamic analysis, each time history variation is analyzed with three different scaled time intervals: the reference interval and plus or minus 10 percent. The frequency shift due to the 10 percent variation of the time scale is considered to be effectively equivalent to the broadening of spectral peaks that is done when generating in-structure response spectra. Considering the defined variations, a total of 12 time histories are considered in the analysis. The LOCA time histories are derived from bounding high energy line breaks in the primary coolant system and inadvertent or spurious operation of reactor coolant pressure boundary valves. Core plate motions are the combined dynamic response due to asymmetric cavity pressurization of the containment, depressurization of the reactor pressure vessel, and thrust force at the break or valve location. The LOCA events for the NPM also result in vertical hydraulic forces acting on the reactor internals. These forces are considered in addition to the core plate motions as a source of excitation for the fuel, as described in Reference 9.1.7.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 56 4.3.5.2 Analysis Results 4.3.5.2.1 Lateral Analysis The horizontal excitation of the full reactor core is considered in the analysis through a series of two dimensional row models with lengths of three, five, and seven fuel assemblies (Figure 3-2). Excitations in both horizontal directions are considered. The peak impact loads for the base case seismic inputs and the margin to the grid impact load limit from all cases are summarized in Table 4-8. The peak impact loads for SSE and LOCA in a given direction are combined and margin is calculated against the impact load limit. The positive margin for these impact loads confirms that the spacer grid does not experience plastic deformation that exceeds the limit established in the Framatome methodology (Reference 9.1.7). Thus, the requirements for core coolability and control rod insertion are met. 4.3.5.2.2 Vertical Analysis The single assembly vertical model is subjected to vertical core plate displacement time histories corresponding to the SSE and LOCA events. The maximum seismic impact load for the base case inputs, [ ] The maximum LOCA impact load is [ ] Component loads for the guide tubes, fuel rods, hold-down spring, nozzles, and guide tube connections are extracted from the vertical analysis for further load analysis. 4.3.5.2.3 Stress Analysis The lateral and vertical analysis results are used as inputs to load and stress evaluations of the non-grid fuel assembly components. Lateral and vertical loads are combined, along with steady-state normal operating loads, for each component for comparison to its respective acceptance criteria. As defined in Section 8.1.2 of Reference 9.1.7, [ ] Table 4-8 Peak Grid Impact Loads and Margins [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 57 [ ] The results of the component evaluations are presented in Table 4-9. Fuel rods The accident loads on the fuel rod consist of the lateral bending stress, the axial stress from vertical loads, and an additional loading corresponding to impacts to the fuel assembly. The impact induced stress is a localized bending stress resulting from the portion of the grid impacts that are passed through the fuel rods (i.e., internal impact loads). The overall combined stress is the SRSS of the individual components combined with steady-state stresses to determine the maximum stress intensity. The faulted stress intensity limits in Table 4-6 are based on Service Level C limits defined in Reference 9.1.13. As such, the [ ] If the calculated maximum stress intensity is [ ] The fuel rods are also evaluated for buckling under compressive loads. Guide tubes The accident loads on the guide tube consist of [ ] The overall combined stress is calculated as the SRSS of the individual components combined with steady-state stresses to determine a maximum stress state. The allowable stress limits are based on ASME Code Level C service limits, consistent with Reference 9.1.7. The guide tubes are also evaluated for buckling under compressive loads.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 58 Guide tube-to-spacer connections The vertical load acting on the guide tube-to-spacer grid connection is [ ] The combined load is calculated as the SRSS of the individual components. The allowable strength of the guide tube-to-grid connections is based on the ASME Code Level D service limit and is established through testing. Guide tube-to-nozzle connections The loads considered in the evaluation of the guide tube-to-nozzle connections are [ ] The allowable load for the top nozzle connection is based on the ASME Code Level C service limit and is established through strength testing. The allowable load for the bottom nozzle connection is based on ASME Code Level D service limits because the bottom nozzle cannot affect control rod insertion. Top and bottom nozzles The loads considered in the evaluation of the nozzles are the axial loads from the vertical analysis. The allowable strength of the top and bottom nozzle is established by testing. Because of the robustness of the nozzles, the testing is not carried to the extent of failure, and thus the allowable load and resulting margin are artificially low.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 59 4.4 Thermal Hydraulic Evaluation 4.4.1 Core Pressure Drop Evaluation An evaluation is performed of the fuel assembly pressure drop characteristics. The total pressure drop, including the recoverable and unrecoverable constituents, is determined for the limiting fuel assembly in the reactor core for rated core power and system flow conditions. Pressure loss coefficients based on testing are reported in Table 5-1 and are used in the evaluation. The results of the pressure drop analysis indicate that the total pressure drop is dominated by the recoverable loss because of elevation change. For the full power case, the channel with the largest pressure drop has an elevation loss that is approximately (( }}2(a),(c),ECI percent of the total pressure drop. The overall core total pressure drop for rated power conditions is (( }}2(a),(c),ECI Figure 4-8 shows the total pressure drop axial profile for the limiting assembly in the core. As observed in the figure, the elevation pressure drop is gradual and linear between each grid or nozzle while the component pressure loss increases the pressure drop slope at the component axial location. The results are based on Table 4-9 Component Evaluation Margins [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 60 operation at 100 percent rated power conditions with a bounding radial and axial core power distribution. 4.4.2 Guide Tube Boiling The guide tube boiling analysis determines the water temperature profile inside the guide tubes of all assemblies, with the limiting conditions found in those fuel assemblies that contain CRAs. The analysis considers the limiting CRA positions and power levels (i.e., the normal parked all rods out position and the power dependent insertion limit position). The design criterion is that long term bulk boiling in the guide tube is precluded. This criterion is satisfied by demonstrating that the coolant flow inside the guide tube remains below the saturation temperature under conservative thermal-hydraulic conditions and assembly relative power. Figure 4-8 Best Estimate Axial Pressure Drop for Limiting Assembly (( }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 61 The guide tube coolant temperatures are dependent on the coolant flow rates inside the guide tubes, the amount of radiation heating of the control rod, the direct heating of the water inside the guide tube, and the direct heating of the guide tube itself. Heat transfer with the surrounding assembly subchannels is modeled with conduction through the guide tube wall and convection with axial fluid flow. Conservative analyses are performed by evaluating normal operation as a function of power and RCS flow rate. Core geometry parameter uncertainties are considered in the areas of (( }}2(a),(c),ECI The guide tube heating analysis determines the coolant flow distribution in the core, including rodded and unrodded guide tubes as a function of elevation. Flow enters the guide tube through holes in the side of the guide tube and from below the bottom nozzle by way of the through-hole in the cap screw. Analysis results for the 100-percent full-power case and 75-percent power case indicate that bulk boiling does not occur at any location within the guide tube under steady-state conditions for the equilibrium core and control rod power dependent insertion limit depths. Results are shown for 100-percent and 75-percent power in Figure 4-9 and Figure 4-10, respectively. The 100-percent power case is limiting (( }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 63 4.4.3 Control Rod Drop Analysis The control rod drop analysis predicts the insertion rate and impact velocity of the CRA during a reactor trip. The calculated impact velocity is compared to the maximum acceptable impact velocity for the CRA spring in Section 6.2.9. When the CRA is dropped into a fuel assembly, water in the guide tube is displaced through several flow paths. The rate of displacement depends on the number, size, and location of the holes along the guide tube. The fuel assembly design has 24 guide tubes, each containing two pairs of side flow holes at the entrance to the dashpot. In addition, water is forced out through the top annulus of the guide tube and through the hole in the cap screw at the bottom of the guide tube assembly. Figure 4-10 Internal Guide Tube Temperature for 75 Percent Power (( }}2(a),(c),ECI

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 64 A best-estimate, mechanistic model is used to evaluate the maximum impact velocity based on the guide tube and control rod geometry, nonlinear coefficients for drag loss (hydraulic and mechanical), and the equation of motion. Drag coefficients were developed with control rod drop measurements from a 17x17 PWR plant with a similar fuel geometry and from CRA drop testing. Coolant flow velocity through the guide tube is conservatively assumed to be zero to maximize the impact velocity. The CRA impact velocity limit is defined in Section 6.2.9 based on the CRA spring design. The control rod drop analysis predicts an impact velocity of 2.09 ft/sec, which is below the impact velocity limit, and a drop time of 1.14 sec, where drop time is defined as the time between the start of rod movement and the time of full insertion. Figure 4-11 shows axial position versus time for the CRA drop based on a 75.982 inch travel distance for the control rod from its initial position to full insertion. Figure 4-12 shows CRA velocity versus time. Figure 4-11 Control Rod Position Versus Time 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Position (Inches) Time (s)

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 65 Figure 4-12 Control Rod Velocity Versus Time 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Velocity (ft/s) Time (s)

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 66 5.0 Fuel Assembly Testing 5.1 Mechanical Testing Summary A comprehensive test program was conducted at Framatomes Richland Test Facility to characterize the mechanical performance of the NuFuel-HTP2' fuel design. The test results are used in the fuel assembly normal operation and seismic analyses to determine the acceptability of the design for in-reactor operation. Prototypical fuel assemblies for BOL and EOL conditions were fabricated and mechanically tested. The BOL assembly spacer grids are in the as-fabricated BOL condition. The EOL assembly simulates the EOL conditions with grid cells relaxed and fuel rods seated on the bottom nozzle. The fuel assembly characterization tests and their use in modeling and analysis are further described in Reference 9.1.7. 5.1.1 Fuel Assembly Lateral Load Deflection (Stiffness) Test The lateral load deflection (stiffness) test is performed to characterize the static, lateral structural response of the fuel assembly at BOL and EOL. The assembly is secured in prototypical upper and lower core plate interfaces. The test is performed by laterally deflecting the center of the test assembly at the second HTP' spacer grid from the bottom to a displacement along one axis. The force required to deflect the assembly and the corresponding displacement are recorded continuously for the complete loading and unloading cycle. 5.1.2 Fuel Assembly Free Vibration (Lateral Pluck) Test The free vibration (lateral pluck) test is performed to characterize the dynamic, lateral response of the fuel assembly over a large range of amplitudes at BOL and EOL. The assembly is secured in prototypical upper and lower core plate interfaces. The test is performed by laterally deflecting at the second HTP' spacer grid from the bottom of the test fuel assembly to a given displacement, and obtaining the response of the assembly when the applied force is suddenly released. Deflection versus time is measured and is used to establish the fuel assembly first mode and damping. 5.1.3 Fuel Assembly Lateral Impact Test The lateral impact test is performed to characterize the dynamic, lateral impact behavior of the fuel assembly at BOL and EOL. The test is performed in two phases. In the first phase, the test consists of deflecting the fuel assembly to a given displacement at the second HTP' spacer grid from the bottom and then suddenly releasing the load allowing it to impact on a baffle plate at the third spacer grid from the bottom location. In the second phase, the test consists of deflecting the fuel assembly to a given displacement at the third spacer grid location from the bottom and then suddenly releasing the load allowing it to impact on a baffle plate at the second spacer grid from the bottom location. The assembly response is obtained in

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 67 the form of deflection versus time measured at two spacer elevations corresponding to the pull and impact locations. Instrumentation is also used to monitor out of plane movement and twist during the test. A load cell attached between the baffle plate and the support measures the test assembly impact force. 5.1.4 Fuel Assembly Lateral Forced Vibration Test The lateral forced vibration test is used to characterize the dynamic, lateral response of the fuel assembly at BOL and EOL. This test complements the free vibration test by providing information on higher modes of the fuel assembly natural frequency, but is typically limited to smaller amplitudes than the free vibration test. This test is performed by applying a dynamic horizontal motion to the test assembly. The fuel assembly is installed on the seismic test stand, and secured in a prototypical support fixture, in order to achieve fixed-fixed end boundary conditions. The input is applied at the first and second intermediate spacer grids. These locations are selected in order to be able to excite all modes of interest. For each mode, the fuel assembly response is measured by accelerometers and displacement sensors attached at the HTP' grid locations. [ ], the evolution of frequency versus vibration amplitude is analyzed. 5.1.5 Fuel Assembly Axial Stiffness Test The axial stiffness test is performed to characterize the static, axial structural response of the fuel assembly at BOL and EOL. The test is performed in the same fixture used for the free and forced vibration testing. The fuel assembly is secured at the top and bottom plates with a simulated core plate fixture. A jack screw is mounted between the simulated core plate and the upper support structure. A load cell is mounted between the lower support plate and the floor plate. The jack is used to apply the load, and the load cell measures the applied load. The axial deflection of the fuel assembly, under load, is measured at key locations with respect to a fixed reference. Instrumentation is also used to monitor lateral movement of the fuel assembly. 5.1.6 Fuel Assembly Drop Test The drop test is performed to characterize the dynamic, axial structural response of the fuel assembly at BOL and EOL. The test fuel assembly is suspended a specified distance above a plate attached to a load cell. The assembly is released and allowed to fall onto the plate and load cell. The displacement of the fuel assembly bottom nozzle is measured throughout the test. Because the use of this data is to calibrate the impact behavior of the vertical seismic model, it is necessary to collect data on force and displacement as a function of time. 5.1.7 Spacer Grid Tests The mechanical performance of the spacer grids was confirmed through a series of structural tests on prototype grids.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 68 Dynamic crush tests are performed on HTP' spacer grids at unirradiated and simulated-irradiated conditions. The tests determine the through-grid stiffness and damping values for the lateral seismic models and the crushing load limits for the grids. The static crush characteristics (static stiffness and elastic load limit) are used to establish allowable grid clamping loads applied during shipping. Grid slip load testing defines the grid-to-fuel rod slip load at non-irradiated (BOL) conditions for both the HTP' and HMP' grids. Grid slip load testing is not performed at irradiated (EOL) conditions because the fuel rods are not actively restrained in the grid. The slip load values are used in the fuel assembly evaluation. 5.1.8 Top and Bottom Nozzle Tests Strength testing of the bottom nozzle is performed to establish the axial load limit for evaluation. A prototypical bottom nozzle is tested at room temperature in static axial compression by applying a load to 24 springs on the guide tube positions. The spring stiffness is set to be equal to the guide tubes stiffness in order to simulate the load distribution of the guide tubes. A maximum room temperature test load is applied without collapse of the structure. This tested maximum load is used to demonstrate the structural adequacy in the design evaluation by comparison with the normal operating and faulted loads. Strength testing of the top nozzle is also performed to establish the axial load limit for evaluation. A prototypical top nozzle is tested at room temperature in static axial compression by applying a load to the top of the top nozzle, which is set on 24 springs at the guide tube positions. The spring stiffness is set to be equal to the guide tubes stiffness in order to simulate the real load distribution of the guide tubes. A room temperature test load is applied that exceeds the design load and resulted in no plastic deformation of the structure. This tested maximum load is used to demonstrate the structural adequacy in the design evaluation by comparison with the shipping and handling, normal operating, and faulted loads. 5.2 Thermal-Hydraulic Testing Summary 5.2.1 Pressure Drop and Liftoff Testing and Pressure Loss Coefficient Development Pressure drop and liftoff testing was performed on a full-scale NuFuel-HTP2' prototype fuel assembly in the PHTF at the Framatome Richland Test Facility. The testing configuration simulated the upper and lower core supports in the NPM. The test data obtained from the pressure drop and liftoff testing are used to develop pressure loss coefficients for subsequent thermal-hydraulic and mechanical analyses. The pressure loss coefficients for the spacer grids and the overall loss for the NuFuel-HTP2' fuel assembly are dependent on the Reynolds number. The coolant

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 69 flow in the NPM is driven by natural circulation. At nominal flow conditions, the Reynolds number is approximately 86,000. Pressure drop data are reduced to Reynolds number dependent values and are used in the development of the pressure drop coefficient correlation. The pressure loss coefficients for assembly components and the overall fuel assembly are given in Table 5-1. In addition to the pressure drop test, a hydraulic liftoff test was performed in the PHTF to acquire data to develop a correlation to be used in fuel assembly hydraulic lift analyses. The liftoff test is performed on a prototypical fuel assembly at six different temperatures for characterization over a range of Reynolds numbers. At each temperature, the flow is adjusted to obtain a conservative lift point, which is defined as the flow and temperature state at which the assembly is barely seated. Pressure drop measurements are taken at the conservative lift point and at a state point where the assembly lifts measurably. An overall pressure loss coefficient for use in hydraulic lift analyses is determined and is given in Table 5-1. 5.2.2 Flow-Induced Vibration Testing A 1000-hour life and wear test was performed at the Richland PHTF to validate the fretting performance of the fuel assembly structure. The life and wear test is run for 1032 hours at a temperature of 300 degrees F at or above the target Reynolds number of 52,000. At the conclusion of the test, [ ] fuel rods are examined for grid to rod fretting. The test results are summarized in Section 4.1.3. Table 5-1 Pressure Loss Coefficients Derived from Testing [ ]

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 70 6.0 Control Rod Assembly 6.1 Control Rod Assembly Description The CRA includes 24 individual control rods fastened to a one-piece cast stainless steel spider and coupling hub (Figure 6-1). Table 6-1 provides major CRA design parameters. The top end of each individual control rod attaches to the spider by a nut, pin, and weld combination. Compared to Framatomes standard rod control cluster assembly (RCCA) design, the individual control rods are shortened to match the reactor core height, but retain the same basic design features and materials. Framatomes standard RCCA design has been implemented in twelve 17x17 PWRs in the United States, comprising over 600 individual assemblies. The combination of the pin, nut, upper end plug, and spider boss form a flex joint, which provides flexibility to accommodate potential misalignment between the CRA and fuel assembly guide tubes. The upper end plug has a reduced diameter shank and lower shoulder to provide lateral clearance with the interior diameter of the spider finger. The clearance allows for elastic deflection of the upper end plug for any misaligned control rod, fuel assembly, upper internals, or fuel handling equipment. A preloaded helical spring is assembled into a skirt internal to the bottom of the spider hub and provides for energy absorption during a CRA trip. The spring is preloaded and maintained within the hub by a retaining ring and tension bolt. During a refueling outage or after a reactor trip, the spring retaining ring rests on the fuel assembly top nozzle. The CRA interfaces with the control rod drive mechanism coupling through a cavity at the top of the CRA spider (Figure 6-2) that matches the male coupling dimensions on the drive shaft, similar to current designs in operation. A 302 stainless steel plenum spring is used within the individual rods to restrain motion of the absorber materials within the cladding during shipping and handling. The absorber material is a combination of B4C pellets and silver-indium-cadmium (AIC) bar. The lower AIC absorber is located in the higher flux region because it has lower swelling under irradiation than B4C. A stack support resides within the annulus of the lowermost AIC absorber to reduce compressive loads on the bottom segment of AIC, thus reducing thermal creep of the AIC during operation. The control rod cladding is 304L stainless steel tubing with stainless steel end plugs welded to each end, encapsulating the rod internals to complete the rod assemblies (Figure 6-3). The design differences between the Framatome standard RCCA and the NuScale CRA are the length of the rod components, the cladding material, and spider spring and retainer modifications. Framatome operating experience has identified two life-limiting phenomena for control rod assemblies: cladding strain and cladding wear. Strain behavior of the CRA is similar given the diametral equivalence of the rods. The minor difference in cladding material should not be a significant factor in the allowable cladding strain. Cladding wear is expected to be acceptable on the CRA given the low axial flow rates of the NPM.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 71 The changes to the spider spring and retainer design relative to the standard Framatome RCCA design are made to increase the spring preload and decrease the spring solid height. These changes allow the CRA spider to absorb the energy from a higher impact force than is experienced in typical PWR applications. Control rod assembly materials exposed to reactor coolant are either low carbon stainless steels or Alloy 718, all of which are resistant to corrosion from reactor coolant exposure. These materials have been used extensively and successfully in operating PWRs. The B4C and AIC absorber materials are encapsulated in a 304L stainless steel tube that is welded at both ends, which protects the absorbers from coolant interaction. Control rod integrity is confirmed by inspection during initial refueling outages. Table 6-2 identifies CRA component materials. Table 6-1 Control Rod Design Parameters Parameter Value CRA total weight (lb) 43 CRA total height (inch) 94.37 Control rod length - short/medium/long (inch) 87.065 / 87.425 / 87.875 Control rod outer diameter (inch) 0.381 Control rod inner diameter (inch) 0.344 Control rod bottom end plug length (inch) 1.913 B4C outer diameter (inch) 0.333 B4C stack length (inch) 62.0 AIC outer diameter (inch) 0.336 AIC stack length (inch) 12.0 Height of CRA spider assembly (inch) 10.387 CRA shaft outer diameter (inch) 1.804 Table 6-2 Control Rod Assembly Materials Component Material Spider 304L stainless steel Rod end plugs 308L stainless steel Cladding 304L stainless steel Solid spacer, lock pin, nuts, tension bolt 304L stainless steel Spring retainer 17-4 PH stainless steel Spider spring Alloy 718 Control rod plenum spring 302 stainless steel Absorber materials 80% Ag - 15% In - 5% Cd and B4C Stack support Alloy X750

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 72 Figure 6-1 Control Rod Assembly General Arrangement ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6 ,1&+(6

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 73 Figure 6-2 Control Rod Assembly Cut-Away 63,'(50$&+,1,1* 187 63,'(5635,1* 7(16,21%2/7 635,1*5(7$,1(5 &21752/52'

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 74 6.2 Control Rod Assembly Evaluation This section evaluates the CRA design against typical criteria to demonstrate acceptable performance under all conditions of operation over a 20 effective full power year (EFPY) design lifetime. 6.2.1 Cladding Strain The potential for control rod cladding strain is primarily a result of swelling of the control rod absorber material caused by neutron fluence. The analysis considers elevation-specific fluence values and thermal expansion of the absorber and cladding. The calculated cladding and absorber temperatures are based on the flux predicted for various axial positions along the control rods. Volumetric swelling rates of the B4C pellets and the AIC absorber are based on models benchmarked to measurements from in-reactor control components. The strain calculation is performed at the following axial elevations:

bottom of the annular AIC bar, where the fluence is highest

bottom of the solid AIC bar

bottom of the B4C pellet stack Cladding strain is limited to [ ] percent to maintain ductility for irradiated 304L stainless steel cladding. The strain calculation determines that the cladding strain limit is met for greater than 20 EFPY of operation. Figure 6-3 Control Rod Design ,1&+(6 ,1&+(6 ,1&+(6 '(6&5,37,21 &21752/52' 833(5(1'3/8* /2:(5(1'3/8* &/$,1* $118/$5$,&$%625%(5 $,&$%625%(552' 62/,'63$&(5 67$&.6833257 3/(180635,1* %&3(//(7 ,7(0 ;6+25752' ;0(',8052' ;/21*52' &21752/52'$66(0%/<

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 75 6.2.2 Cladding Creep Collapse Cladding creep collapse evaluations include short-term and long-term collapse analyses. In the short-term collapse analysis, the differential pressure across the control rod cladding must not exceed the critical buckling pressure. Two critical buckling pressures are calculated:

bifurcation buckling pressure of a perfectly circular shell (Pcr), to confirm the elastic stability of the cladding

yield-point buckling pressure (Pyp) accounting for initial tubing ovality The calculated buckling pressures at hot and cold conditions exceed the system pressure of 2000 psia and a conservative maximum reactor system pressure of 2310 psia, providing margin to the buckling criteria. In the long-term collapse analysis, changes in cladding ovalization and cladding stress over time are predicted using the CROV creep ovalization code. The analysis assumes no cladding support by the control rod internals. Creep collapse is evaluated at the lowest tip of the control rod because this region experiences the largest fast flux. The CROV analysis demonstrates that after 20 EFPY, the cladding ovality remains within acceptance limits, the cladding stress remains below the material yield strength, and the maximum cladding diameter as a result of ovalization is less than the inner diameter of the guide tube dashpot. 6.2.3 Cladding Stress Control rod cladding stresses are categorized, calculated, and compared to service limits in accordance with the ASME Code. Cladding stresses are calculated based on

differential pressure.

differential temperature.

cladding ovality.

loads from control rod drive mechanism stepping and reactor trip.

bending due to misalignment.

seismic conditions.

FIV.

shipping and handling conditions (evaluated in isolation from other conditions).

stuck rod condition (evaluated in isolation from other conditions).

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 76 The design stress intensity, Sm, is two-thirds of the cladding yield strength [ ]. Table 6-3 defines the allowable stresses for each of the four ASME Code stress categories. The calculated stresses result in a minimum design margin of [ ], which occurs for primary membrane stresses at hot conditions, where margin is defined as the allowable stress divided by the calculated stress. 6.2.4 Cladding Fatigue The control rod cladding is analyzed for fatigue from stepping loads and FIV bending loads, assuming no wear. The analysis conservatively assumes infinite cycles from FIV and control rod stepping. The analysis concludes that the fatigue stress is below the endurance stress limit of the cladding material based on the ASME Code fatigue curve (12 ksi), and therefore fatigue failure does not occur. 6.2.5 Cladding Wear Control rod cladding wear limits are determined by reducing the cladding wall thickness in mechanical analyses until the margin to acceptance limits is reduced to zero. This method is consistent with the industry approach for PWRs. Because of the potential for leaching of the B4C pellets and subsequent impact on shutdown capability if the cladding barrier is breached, a [ ] reduction in the minimum cladding wall thickness is conservatively applied to calculate the wear limits. The calculated wear limits address circumferential wear and azimuthally localized wear. The following limits are calculated:

maximum wear depth of [ ] inch, independent of geometry Table 6-3 Control Rod Cladding Allowable Stresses Stress Category Temperature, °F ASME Allowable Stress, psi [ A: Primary membrane 70 Pm Sm 650 B: Primary membrane + bending 70 Pm + Pb 1.5Sm 650 C: Primary and secondary membrane + bending 70 Pm + Pb + Q 3.0Sm 650 D: Faulted 70 Pm Sm 650 Pm + Pb 2.25Sm 70 650 ]

minimum cross-sectional area of [ ] inch2 remaining after wear (uniform circumferential wear)

minimum cross-sectional area of [ ] inch2 remaining after wear (azimuthally localized wear, considering [ ]) Wear limits are used in conjunction with wear rates specific to the NPM design to determine an allowable wear-based design life. After initial irradiation and operation of the CRA design, inspections are performed so that actual rod wear rates can be compared with the predetermined wear limits to demonstrate acceptable performance. The operating environment for the CRAs is expected to be less severe with respect to rod wear than the environment typical of operating PWRs. Axial flow rates in the reactor core and in the guide tubes are significantly lower and cross flows at and above the fuel assembly top nozzles are very low (approximately (( }}2(a),(c),ECI feet/second). The absence of outlet flow nozzles in the upper internals reduces the cross flows compared to a typical PWR. These flow conditions create a more benign flow environment, reducing mechanical interactions with the guide cards and fuel assemblies. Based on this assessment, the CRA design lifetime is not expected to be limited by control rod wear. 6.2.6 Control Rod Internal Pressure The control rod internal pressure analysis predicts the maximum internal rod pressure using a conservative model that calculates the depletion in the B4C pellets and release of helium to the rod plenum volume. The calculation includes helium backfill, residual, and sorbed gases in the determination of the final maximum internal pressure. The AIC material is not a source of gases. During normal operation, the CRAs are positioned such that the B4C pellets are located above the active fuel. Over the lifetime of the CRA, there is very low depletion, which results in insignificant helium production. In addition, the large porosity of the B4C absorber material provides sufficient volume to accommodate any helium produced. However, the analysis conservatively assumes [ ] percent release and retention of the helium due to depletion of the B4C pellets. For a 20 EFPY control rod assembly design life, a conservatively calculated 10B depletion of [ ] percent creates a predicted maximum rod internal pressure of [ ] psia, which meets the criterion of being less than RCS pressure (2000 psia). 6.2.7 Component Melt Analysis The control rod is analyzed to ensure that each component remains below the melt temperature. The analysis uses conservative values for heating rates and gap conductance. The worst case calculated temperatures for all rod components are well below the material melt limits.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 78 6.2.8 Spider Assembly Structural Analysis The spider assembly structural analysis evaluates the static and fatigue stresses in the CRA spider assembly and the control rod to spider connections. The following loads are analyzed:

reactor trip and CRA stepping

stuck rod

shipping and handling

hydraulic load during reactor trip The following elements of the spider assembly are analyzed:

spider arm

CRA spring

spring retainer

spring tension bolt

spring flange

spring housing

spider coupling splines

upper end plug, nut, and spider connection Calculated stresses are compared to ASME Code minimum material strength values. The results demonstrate positive margins for all components, validating the structural integrity of the spider assembly and connections during normal operation and faulted conditions. 6.2.9 Control Rod Assembly Impact Velocity Limit The kinetic energy absorption capacity of the CRA spring is analyzed to determine the CRA maximum allowable impact velocity during a reactor trip. The design has a longer and heavier CRA driveline than is typical for PWRs and has lower axial flow rates, resulting in a higher CRA impact velocity. The spider spring is designed to absorb the kinetic energy of the CRA during a reactor trip using the available spring retainer travel to prevent the CRA spider hub from impacting the fuel assembly top nozzle. The maximum allowable impact velocity is [ ] ft/sec. Section 4.4.3 describes a calculation performed that determined an impact velocity of 2.09 ft/sec. 6.3 Control Rod Assembly Testing The CRA is similar to existing 17x17 control rod assemblies except for the shorter length. The CRA drive shaft is longer than typically used in the industry. Prototype testing was

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 79 performed to confirm CRA drop times, to assess the propensity for wear, and to assess the impact of the maximum expected misalignment of the fuel assembly guide tubes, guide cards, and riser supports predicted to occur during a concurrent LOCA and seismic event. The combination of misalignments resulting in the slowest drop produced a drop time of approximately 1.2 seconds. Section 4.4.3 describes the calculation performed to determine the impact velocity.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 80 7.0 Design Change Process The following defines the mechanical design change process for the NuFuel-HTP2' fuel design. The design change process is modeled after the NRC-approved change process defined in Reference 9.1.2 and is in conformance to the NuScale design control requirements and 10 CFR 50 Appendix B Quality Assurance Program. The fuel design criteria for the NuFuel-HTP2' design are distributed across several sources. Reference 9.1.8, Table 7-1 identifies the mechanical design criteria and source of applicability to the NuFuel-HTP2' design. The design criteria sources are

Framatome methodologies that are applicable to the NuFuel-HTP2' design provided in Reference 9.1.4, Reference 9.1.5, Reference 9.1.6, Reference 9.1.7, Reference 9.1.2, Reference 9.1.3, and Reference 9.1.14.

NuScale methodologies provided in Reference 9.1.8, Reference 9.1.9, and Reference 9.1.10.

the remaining criteria that are evaluated by the methodologies as approved in the FSAR. Applicable design criteria are summarized in Section 4.0. Compliance to these criteria is demonstrated by

documenting the fuel system and fuel assembly design drawings.

performing analyses with the NRC-approved models and methods described in Section 4.0.

using lead test assemblies, prototypic testing and engineering analyses, where appropriate, to demonstrate that analytical methods and acceptance criteria remain applicable and to demonstrate in-reactor performance.

continuing irradiation surveillance programs, including post irradiation examinations, to confirm fuel assembly performance.

using the quality assurance procedures, quality control inspection program, and design control requirements set forth in the NRC-approved NuScale Quality Assurance Program. Design changes can be made without NRC review and approval if the following conditions are met:

demonstrating compliance with the approved criteria as defined above

changes in plant technical specifications are not required

the applicability of NRC-approved methodologies as described in Section 4.0 is demonstrated to be valid

burnup limits are within those approved by the NRC

a change in the attachment of the spacer grid to the guide tubes

a change in the strip thickness of the spacer grid

a change in cladding thickness Examples of changes that would not meet these conditions:

new cladding material

a spacer grid with a new mixing behavior or new rod support mechanism

a change that would alter the fuel behavior relative to the NRC-approved models, for example rod growth, assembly growth, or clad corrosion

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 82 8.0 Summary and Conclusions This report describes the NuFuel-HTP2' fuel assembly design and corresponding CRA design. The designs incorporate features with extensive operating experience and are evaluated using NRC-approved evaluation methods. The testing and design evaluations demonstrate that the designs meet regulatory criteria and perform acceptably in the NPM.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Revision 1 © Copyright 2024 by NuScale Power, LLC 83 9.0 References 9.1 Source Documents 9.1.1 NUREG-0800, U.S. NRC Standard Review Plan Section 4.2, Revision 3, Fuel System Design, March 2007. 9.1.2 EMF-92-116(P)(A), Rev. 0, Generic Mechanical Design Criteria for PWR Fuel Designs, February 2015. 9.1.3 BAW-10227P-A, Rev. 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, June 2003. 9.1.4 ANP-10231PA-01, COPERNIC Fuel Rod Design Computer Code, January 2004. 9.1.5 ANP-10084PA-03, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse (CROV computer code), October 1980. 9.1.6 XN-75-32 (P) (A), Supplements 1-4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983. 9.1.7 ANP-10337P-A, Rev. 0, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, April 2018. 9.1.8 TR-0116-20825-P-A, Applicability of AREVA Fuel Methodology for the NuScale Design, June 2016, Revision 1. 9.1.9 TR-0716-50351-P, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, December 2019, Revision 1. 9.1.10 TR-108553-P-A, Framatome Fuel and Structural Response Methodologies Applicability to NuScale, Supplement 1 to TR-0116-20825-P-A, Revision 1, Supplement 1 to TR-0716-50351-P-A, Revision 1, October 2022, Revision 0. 9.1.11 ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Subsection NG, Core Support Structures, 2010 Edition with 2011a Addenda, July 1, 2011. 9.1.12 ODonnell, W.J. and B.F. Langer, Fatigue Design Basis for Zircaloy Components, Nuclear Science and Engineering, Volume 20, pp. 1-12, September 1964. 9.1.13 ASME Boiler and Pressure Vessel Code, Section III, Division 1, 2019 Edition. 9.1.14 BAW-10227P-A, Rev. 2, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, January 2023.

LO-176338 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 Mark W. Shaver, AF-176339

AF-176339 Page 1 of 2

NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Director 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 NuFuel-HTP2 Fuel and Control Rod Assembly Designs. NuScale has performed significant research and evaluation to develop a basis for this process 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, NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-117605, Revision 1. 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.

AF-176339 Page 2 of 2 (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 § 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 March 14, 2025. Mark W. Shaver

LO-176338 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 Morris Byram, Framatome Inc.

A F F I D A V I T

1.

My name is Morris Byram. I am Product Manager, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3.

I am familiar with the Framatome information contained in Enclosure 1 to the NuScale Power, LLC letter Number LO-176338 with subject, NuScale Power, LLC Submittal of NuFuel-HTP2TM Fuel and Control Rod Assembly Designs, TR-117605-P, Revision 1, and referred to herein as Document. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6.

The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary: (a) The information reveals details of Framatomes research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability. (e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome. The information in this Document is considered proprietary for the reasons set forth in paragraph 6(b), 6(c), and 6(d) above.

7.

In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct. Executed on: (3/13/2025) (NAME) Email: morris.byram@framatome.com Phone: 434-221-1082 BYRAM Morris Digitally signed by BYRAM Morris Date: 2025.03.13 13:32:49 -07'00'}}

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