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LLC Submittal of NuFuel-HTP2 Fuel and Control Rod Assembly Designs, TR-0816-51127, Revision 2
	ML19211C901
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Site:	NuScale
Issue date:	07/30/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of New Reactors
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ML19211C899	List: ML19211C901;
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	LO-0719-66433 TR-0816-51127-NP, Rev 2
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LO-0719-66433 July 30, 2019 Docket No.52-048 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-0816-51127, Revision 2

REFERENCES:

1. Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, Submittal of NuScale Technical Report NuFuel-HTP2' Fuel and Control Rod Assembly Designs, Revision 1, dated January 7, 2017 (ML17007A001)

NuScale Power, LLC (NuScale) hereby submits Revision 2 of the NuFuel-HTP2' Fuel and Control Rod Assembly Designs (TR-0816-51127) technical report. contains the proprietary version of the report entitled NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-P, Revision 2. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. Enclosure 3 pertains to the NuScale proprietary information to be withheld from the public while Enclosure 4 pertains to the Framatome proprietary information to be withheld from the public. NuScale proprietary is denoted by double curly brackets (i.e., (() while Framatome proprietary information is denoted by straight brackets (i.e., [ ]) in the report. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. Enclosure 2 contains the nonproprietary version of the report entitled NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-NP, Revision 2. This letter makes no regulatory commitments and no revisions to any existing regulatory commitments. If you have any questions, please contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com. Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Bruce Bavol, NRC, OWFN-8H12 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0719-66433 Page 2 of 2 07/29/2019 Enclosure 1: NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-P, Revision 2, proprietary version Enclosure 2: NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-NP, Revision 2, nonproprietary version Enclosure 3: Affidavit of Zackary W. Rad, AF-0719-66434 Enclosure 4: Affidavit of Nathan E. Hottle, Framatome Inc. NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0719-66433 : NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-P, Revision 2, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0719-66433 : NuFuel-HTP2' Fuel and Control Rod Assembly Designs, TR-0816-51127-NP, Revision 2, nonproprietary version 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-0816-51127-NP Rev. 2 Licensing Technical Report NuFuel-HTP2' Fuel and Control Rod Assembly Designs July 2019 Revision 2 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com © Copyright 2019 by NuScale Power, LLC © Copyright 2019 by NuScale Power, LLC i

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report 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. © Copyright 2019 by NuScale Power, LLC ii

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008820. 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. © Copyright 2019 by NuScale Power, LLC iii

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report CONTENTS 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 .................................................................................... 9 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 ......................................................................................................... 27 4.1 Fuel System Damage Criteria ............................................................................. 27 4.1.1 Stress and Loading Limits ................................................................................... 27 4.1.2 Cladding Fatigue ................................................................................................. 32 4.1.3 Fretting ................................................................................................................ 34 4.1.4 Oxidation, Hydriding, and Crud Buildup .............................................................. 36 4.1.5 Fuel Rod Bow ...................................................................................................... 39 4.1.6 Axial Growth ........................................................................................................ 40 4.1.7 Fuel Assembly Distortion Evaluation ................................................................... 41 4.1.8 Fuel Rod Internal Pressure ................................................................................. 42 4.1.9 Assembly Liftoff ................................................................................................... 42 4.2 Fuel Rod Failure Criteria ..................................................................................... 43 4.2.1 Internal Hydriding ................................................................................................ 43 4.2.2 Cladding Collapse ............................................................................................... 43 4.2.3 Overheating of Cladding ...................................................................................... 45 4.2.4 Overheating of Fuel Pellets ................................................................................. 45 © Copyright 2019 by NuScale Power, LLC iv

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report 4.2.5 Excessive Fuel Enthalpy ..................................................................................... 50 4.2.6 Pellet/Cladding Interaction .................................................................................. 50 4.2.7 Bursting ............................................................................................................... 51 4.2.8 Mechanical Fracturing ......................................................................................... 51 4.3 Fuel Coolability .................................................................................................... 51 4.3.1 Cladding Embrittlement ....................................................................................... 51 4.3.2 Violent Expulsion of Fuel ..................................................................................... 51 4.3.3 Generalized Cladding Melting ............................................................................. 51 4.3.4 Fuel Rod Ballooning ............................................................................................ 52 4.3.5 Fuel Assembly Structural Damage from External Forces .................................... 52 4.4 Thermal Hydraulic Evaluation.............................................................................. 59 4.4.1 Core Pressure Drop Evaluation ........................................................................... 59 4.4.2 Guide Tube Boiling .............................................................................................. 63 4.4.3 Control Rod Drop Analysis .................................................................................. 64 5.0 Fuel Assembly Testing ................................................................................................. 66 5.1 Mechanical Testing Summary.............................................................................. 66 5.1.1 Fuel Assembly Lateral Load Deflection (Stiffness) Test (Beginning of Life and End of Life) ................................................................................................... 66 5.1.2 Fuel Assembly Free Vibration (Lateral Pluck) Test (Beginning of Life and End of Life) .......................................................................................................... 66 5.1.3 Fuel Assembly Lateral Impact Test (Beginning of Life and End of Life) .............. 66 5.1.4 Fuel Assembly Lateral Forced Vibration Test (Beginning of Life and End of Life) ..................................................................................................................... 67 5.1.5 Fuel Assembly Axial Stiffness Test (Beginning of Life and End of Life) ............... 67 5.1.6 Fuel Assembly Drop Test (Beginning of Life and End of Life) ............................. 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 © Copyright 2019 by NuScale Power, LLC v

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report 6.2 Control Rod Assembly Evaluation ....................................................................... 74 6.2.1 Cladding Strain .................................................................................................... 74 6.2.2 Cladding Creep Collapse .................................................................................... 75 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 ................................................................... 77 6.2.9 Control Rod Assembly Impact Velocity Limit ....................................................... 78 6.3 Control Rod Assembly Testing............................................................................. 78 7.0 Design Change Process ............................................................................................... 79 8.0 Summary and Conclusions .......................................................................................... 81 9.0 References ..................................................................................................................... 82 9.1 Source Documents .............................................................................................. 82 © Copyright 2019 by NuScale Power, LLC vi

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report TABLES Table 1-1 Abbreviations ......................................................................................................... 6 Table 3-1 NuScale fuel assembly parameters ..................................................................... 12 Table 3-2 NuScale fuel rod parameters ............................................................................... 13 Table 3-3 NuScale operating conditions .............................................................................. 14 Table 3-4 Fuel assembly materials ...................................................................................... 15 Table 4-1 Summary of results - fuel assembly design margins .......................................... 30 Table 4-2 Stress results in compression (beginning of life) and tension (end of life) .......... 32 Table 4-3 Summary of transients considered in the fuel rod fatigue analysis ..................... 34 Table 4-4 Bounding centerline fuel melt limits ..................................................................... 47 Table 4-5 Bounding transient cladding strain limits ............................................................. 51 Table 4-6 Summary of NuScale fuel assembly damping ratios for when the fuel is located in both the NuScale Power Module and the reactor flange tool .............. 53 Table 4-7 Peak grid impact loads and margins ................................................................... 56 Table 4-8 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 © Copyright 2019 by NuScale Power, LLC vii

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Licensing Technical Report FIGURES Figure 3-1 Fuel assembly general arrangement ................................................................... 16 Figure 3-2 Representative core loading pattern.................................................................... 17 Figure 3-3 Top nozzle ........................................................................................................... 18 Figure 3-4 Bottom nozzle ...................................................................................................... 19 Figure 3-5 Guide tube assembly ........................................................................................... 20 Figure 3-6 Cap screw bottom nozzle connection .................................................................. 21 Figure 3-7 Guide tube quick disconnect top nozzle connection............................................ 22 Figure 3-8 HTP' grid ........................................................................................................... 23 Figure 3-9 HTP' spacer grid characteristics ....................................................................... 24 Figure 3-10 HMP' spacer grid .............................................................................................. 25 Figure 3-11 Fuel rod assembly ............................................................................................... 26 Figure 4-1 Bounding power history envelope ....................................................................... 37 Figure 4-2 Predicted corrosion results .................................................................................. 38 Figure 4-3 Axial corrosion profile .......................................................................................... 39 Figure 4-4 CROV results - cladding stress vs. time ............................................................. 44 Figure 4-5 CROV results - cladding deformation rate vs. time............................................. 45 Figure 4-6 Centerline fuel melt and transient cladding strain bounding envelopes for UO2 fuel .......................................................................................................... 47 Figure 4-7 Centerline fuel melt and transient cladding strain bounding envelopes for 2 wt% Gd2O3 fuel............................................................................................ 48 Figure 4-8 Centerline fuel melt and transient cladding strain bounding envelopes for 4 wt% Gd2O3 fuel............................................................................................ 48 Figure 4-9 Centerline fuel melt and transient cladding strain bounding envelopes for 6 wt% Gd2O3 fuel............................................................................................ 49 Figure 4-10 Centerline fuel melt and transient cladding strain bounding envelopes for 8 wt% Gd2O3 fuel............................................................................................ 49 Figure 4-11 Difference in total pressure drop by channel - best estimate.............................. 61 Figure 4-12 Difference in unrecoverable pressure drop by channel - best estimate.............. 62 Figure 4-13 Cumulative unrecoverable pressure drop - best estimate .................................. 63 Figure 4-14 Control rod position versus time .......................................................................... 65 Figure 4-15 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 NuScale control rod design ................................................................................. 74 © Copyright 2019 by NuScale Power, LLC viii

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Abstract This technical report provides a description of the NuScale fuel assembly (NuFuel-HTP2') and control rod assembly (CRA) designs. The report identifies the applicable regulatory requirements and summarizes the analyses and tests performed to demonstrate that the designs meet the regulatory requirements. The evaluations are focused on the mechanical aspects of the designs, consistent with Section 4.2 of NUREG-0800. Neutronic and thermal-hydraulic performance of the designs are addressed in other parts of the NuScale design certification application. © Copyright 2019 by NuScale Power, LLC 1

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Executive Summary The purpose of this report is to describe the NuFuel-HTP2' fuel and control rod assembly (CRA) design, summarize the key analysis results and evaluate 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 GDC 2, GDC 10, PDC 27, PDC 35, and 10 CFR 50.46. Fuel Assembly The NuScale 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 operations, anticipated operational occurrences, and postulated accidents. Potential failure mechanisms include mechanical loading, fatigue, fretting, oxidation, growth, and distortion. The NuScale criteria and methods have been applied to fuel assembly designs currently in power operation and have been demonstrated to be applicable to the NuScale fuel assembly design. The evaluations presented in this report demonstrate that the fuel assembly design meets all design criteria. 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 design certification application and not in this report. The NuScale criteria and methods have been applied to fuel rod designs currently in power operation and have been demonstrated to be applicable to the NuScale design. The evaluations presented in this report demonstrate that the fuel rod design meets all design criteria. Mechanical testing is performed to characterize the mechanical performance of the fuel assembly. The 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. The evaluations and tests performed are comprehensive and demonstrate the acceptability of the fuel assembly design for use in the NuScale Power Module. © Copyright 2019 by NuScale Power, LLC 2

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Control Rod Assembly The NuScale control rod assembly (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 (Ag-In-Cd) 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 and loads, strain, creep collapse, fatigue, wear, internal pressure, and component melting. The evaluations summarized in this report demonstrate that the CRA design meets all criteria and is acceptable for use in the NuScale Power Module. A CRA testing program is defined to confirm CRA drop times and CRA drop velocity, to assess the propensity for wear, and to assess the impact of the maximum expected fuel assembly distortion and the misalignment of the lead screw and guides. © Copyright 2019 by NuScale Power, LLC 3

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 1.0 Introduction NuScale has developed a 17X17 fuel assembly (NuFuel-HTP2') and control rod assembly (CRA) design for use in the NuScale Power Module based on existing AREVA methods and technology. This design incorporates the full range of AREVA 17X17 operating experience for application to the NuScale operating environment. 1.1 Purpose The purpose of this report is to describe the NuFuel-HTP2' fuel assembly and CRA designs, summarize the key analysis results and demonstrate that the designs meet regulatory requirements. 1.2 Scope NuScale used NRC approved AREVA methodologies to perform the analysis of the NuFuel-HTP2' fuel design. This report evaluates the NuFuel-HTP2' fuel design against the requirements of NUREG-0800 Section 4.2 (Reference 9.1.10). The evaluations are focused on the mechanical aspects of the design. Neutronic and thermal hydraulic performance of the fuel assembly are addressed in other parts of the NuScale design certification application. 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 evalutions 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 9.1.6 and Reference 9.1.9).
The COPERNIC Fuel Rod Design Computer Code (Reference 9.1.1) is used in the fuel performance analysis of the NuFuel-HTP2' fuel design.
The CROV code (Reference 9.1.2) is used to evaluate the cladding creep performance.
The computational procedure for evaluating fuel rod bowing (Reference 9.1.4) is used to evaluate the impacts of fuel rod bow for the NuFuel-HTP2' design.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2

The PWR fuel assembly structural response to externally applied dynamic excitations (Reference 9.1.5) is used to perform seismic structural analysis on the fuel design.

The applicability of the fuel analysis methods to the NuScale design is demonstrated in Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P Revision 1 (Reference 9.1.7) and NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-0716-50351-P (Reference 9.1.8). 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 NuScale CRA design satisfies the applicable design requirements. © Copyright 2019 by NuScale Power, LLC 5

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 1-1 Abbreviations Term Definition AIC silver-indium-cadmium 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 FIV flow-induced vibration FSAR Final Safety Analysis Report 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 cluster control 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 © Copyright 2019 by NuScale Power, LLC 6

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 2.0 Background This report supports final safety analysis report (FSAR) Section 4.2 of the NuScale design certification application. 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 9.1.10). 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:

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 (AOO).
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, the following principal design criteria specified in FSAR Section 3.1 of the NuScale design certification application apply to the fuel mechanical evaluations summarized in this report:

PDC 27, as it relates to ensuring that fuel system damage is never so severe as to prevent control rod insertion when it is required.
PDC 35, 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 NuScale design in Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P Revision 1 (Reference 9.1.7) and NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-0716-50351-P (Reference 9.1.8). © Copyright 2019 by NuScale Power, LLC 7

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 3.0 NuFuel-HTP2' Fuel Assembly Description The NuFuel-HTP2' fuel assembly, shown in Figure 3-1, is a 17x17 pressurized water reactor (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 NuScale Power Module. The NuScale core design is comprised of 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, shown in Figure 3-3, consists of a 304L 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 are made of Alloy 718 and are designed to maintain positive hold-down margin. The leaf spring sets are fastened to the top nozzle with Alloy 718 clamp screws. The upper leaf has an extended tang that engages a © Copyright 2019 by NuScale Power, LLC 8

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 cutout in the top plate of the nozzle to retain the spring leaves in the unlikely event of a leaf or clamp screw failure. 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. 3.2 Bottom Nozzle with Mesh Filter Plate The 304L stainless steel bottom nozzle, shown in 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. 3.3 Zircaloy-4 MONOBLOC' Guide Tubes The MONOBLOC' guide tubes, shown in 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, shown in 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. 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 © Copyright 2019 by NuScale Power, LLC 9

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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, shown in 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 all 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 NuScale design are identical to those used on AREVAs 17x17 PWR product that has extensive operating experience in the United States. 3.6 Alloy 718 HMP' Lower Spacer Grid The HMP' spacer grid, shown in 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 AREVAs 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, shown in 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. 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, over 5 million M5 fuel rods have operated in more than 21,000 fuel assemblies in 84 reactors. The operational experience of M5 cladding covers PWR fuel arrays from 14x14 up to 18x18. © Copyright 2019 by NuScale Power, LLC 10

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 The fuel stack height is 78.74 inches and rests on top of the lower end cap. The fuel rod has an internal stainless steel 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 system 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 design density of the UO2 pellets is 96 percent theoretical density with a possible 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. © Copyright 2019 by NuScale Power, LLC 11

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 3-1 NuScale fuel assembly parameters Parameter Value Fuel rod array 17 x 17 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-0816-51127-NP Rev. 2 Table 3-2 NuScale fuel rod parameters Parameter Value Cladding material M5 Fuel rod length (inch) 85.0 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) 5.31 Fuel pellet outer diameter (inch) 0.3195 Fuel pellet length (inch) 0.400 Fuel pellet surface roughness (µin) [ ]ECI Fuel pellet density (% TD) 96 [ ]ECI Resinter densification limits (24-hour test) ECI [ ] Fuel pellet grain size (m) [ ]ECI Fuel pellet open porosity fraction [ ]ECI [ ]ECI Sorbed gas [ ]ECI Fuel pellet dish volume, nominal/percent (mm3) [ ]ECI Fuel pellet void volume, nominal/percent (cm3) [ ]ECI ECI Plenum spring free length (inch) [ ] 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 © Copyright 2019 by NuScale Power, LLC 13

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 3-3 NuScale operating conditions Parameter NuScale Design Value AREVA 17x17 PWR Value Rated thermal power (MWt) 160 3455 Average coolant velocity (ft/s) 2.7 16 System pressure (psia) 1850 2280 Core tave (°F) 543 584 Core avg linear heat rate, approx. (kW/m) 8.2 18.0 Reactor coolant system (RCS) inlet 497 547 temperature (°F) RCS Reynolds number 76,000 468,000 Fuel assemblies in core 37 193 Fuel assembly loading (kgU) 249 455 Core loading (kgU) 9,213 87,815 © Copyright 2019 by NuScale Power, LLC 14

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 3-4 Fuel assembly materials Component Material Top nozzle AISI 304L stainless steel Bottom nozzle frame AISI 304L 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 AISI 302 stainless steel Fuel pellets UO2 and UO2 plus Gd2O3 Note: Stainless steels are low cobalt. © Copyright 2019 by NuScale Power, LLC 15

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 4.0 Design Evaluation This section evaluates the NuFuel-HTP2' design against criteria established consistent with Section 4.2 of NUREG-0800 (Reference 9.1.10), 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, and (4) core coolability is always maintained. The design criteria are based on AREVA fuel design experience and are consistent with criteria established for previously approved fuel assembly designs. Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P Revision 1 (Reference 9.1.7) identifies the NRC-approved codes and methods used to evaluate the fuel performance. Table 2-3 of Reference 9.1.7 associates the method with the corresponding SRP acceptance criteria. NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-0716-50351-P (Reference 9.1.8) identifies the method of analysis for the structural response of the fuel assembly to dynamic faulted loads. 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 NuScale Power Module (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 9.1.6) Buckling of the guide tubes shall not occur during normal operation and AOOs. The cumulative number of strain fatigue cycles on the structural components should be significantly less than the design fatigue lifetime. NuFuel-HTP2' Design Evaluation In the normal operating analysis, a series of mechanical analyses are performed on the NuScale fuel assembly to verify that it 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 limit (Reference 9.1.3). The fuel assembly weight, hold-down spring forces, RCS hydraulic loads, thermal loads during plant heat-up, steady-state operation and cooldown, and control rod assembly (CRA) drop loads (sum of the CRA spring preload and the load from the maximum travel of the CRA spring retainer) create the stress states that are evaluated in the analysis. The operating basis earthquake is less than 1/3 of the safe © Copyright 2019 by NuScale Power, LLC 27

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 shutdown earthquake 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 182 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 9.1.13) 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 due to 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 weld, and the threaded connection between the guide tube lower end fitting and the shoulder screw. The threaded connection is evaluated using ASME code methods. 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 load limit based on prototype component tests. For both the bottom nozzle and top nozzle, testing determined the collapse load for each © Copyright 2019 by NuScale Power, LLC 28

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 structural framework. In accordance with the ASME Code, each tested collapse load is multiplied by 2/3 in establishing the allowable load limit. 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. © Copyright 2019 by NuScale Power, LLC 29

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Consistent with Reference 9.1.6, the fuel handling qualification includes a 2.5g axial load. The shipping qualification includes 4g axial and 6g lateral loads and grid clamp loads. All 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 minimum strength, critical buckling loads, and/or 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 will 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 9.1.9:

        *    [
                                        ]

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 all normal and AOO 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-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

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2

Pm + Pb + Local + Secondary - Pressure , ovality, FIV, FR-SG , growth, and thermal stresses 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. Table 4-2 Stress results in compression (beginning of life) and tension (end of life) 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 (1910 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. 4.1.2 Cladding Fatigue Design Criterion The fuel rod cumulative usage factor shall not exceed 0.9. (Reference 9.1.9) NuFuel-HTP2' Design Evaluation A bounding analysis of the NuScale core design is performed using the COPERNIC fuel rod analysis code (Reference 9.1.1) 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 © Copyright 2019 by NuScale Power, LLC 32

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 will experience 10/60 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. © Copyright 2019 by NuScale Power, LLC 33

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 4-3 Summary of transients considered in the fuel rod fatigue analysis Number Description Number of Cycles Condition I Events (Service Level A) 1 Reactor heatup to hot standby 200 2 Reactor cooldown from hot standby 200 3 Power ascent from hot standby 500 4 Power descent to hot standby 500 5 Load following 19,750 6 Load regulation 767,100 7 Steady-state fluctuations 5,260,000 8 Load ramp increase 2000 9 Load ramp decrease 2000 10 Step load increase 3000 11 Step load decrease 3000 12 Large step load decrease 200 Condition II Events (Service Level B) 13 Turbine trip without bypass 60 14 Loss of all AC power 60 15 Inadvertent main steam isolation valve closure 30 16 Inadvertent operation of decay heat removal system 15 17 Reactor trip from reduced power 180 18 Reactor trip from full power 120 19 Inadvertent control rod assembly drop 60 20 Inadvertent pressurizer spray 15 21 Inadvertent opening of a reactor safety valve 20 Condition III Events (Service Level C) (only one of the following is considered) 22 Spurious reactor vent valve actuation 15 23 Spurious reactor recirculation valve actuation 15 24 Small break loss-of-coolant accident 5 25 Steam generator tube rupture 5 4.1.3 Fretting Design Criterion Fuel rod failures due to fretting shall not occur, as verified by fretting tests. (Reference 9.1.6) 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. © Copyright 2019 by NuScale Power, LLC 34

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2

The 1000-hour life and wear test performed at the Richland PHTF was run for 1032 hours at a temperature of 300°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 significantly 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 very 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 very conservative imposed vibration amplitudes and [

].

The predicted small vibration amplitudes for the NuFuel-HTP2' design are a consequence of much lower axial and cross flow velocities due to the natural circulation (no mechanical pumps) of the NuScale reactor coolant system. The NuScale reactor has a nominal flow rate of 259 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 NuScale fuel assemblies is [ ] 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 flow-induced vibration 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 © Copyright 2019 by NuScale Power, LLC 35

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 NuScale reactor core, and fuel rod failures due to fretting will 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 9.1.1) NuFuel-HTP2' Design Evaluation A bounding analysis is performed for NuScale core designs using the COPERNIC fuel rod analysis code (Reference 9.1.1) 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. An input power history envelope, Figure 4-2, expressed in terms of effective full power hours is used that bounds the individual rod power histories of all the UO2 and Gd2O3 rods in the NuScale 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; therefore, only a UO2 rod is explicitly modeled. The use of a bounding power history envelope makes the analysis equally applicable to all fuel rod types. The predicted oxide thickness is shown in Figure 4-2. The maximum oxide thickness reaches [ ] micrometers, which is well below the design limit of 100 micrometers. © Copyright 2019 by NuScale Power, LLC 36

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-2 Predicted corrosion results The EOL axial corrosion profile is shown in Figure 4-3. The maximum oxide thickness occurs in the lower half of the core, slightly below the middle of the pellet stack. This behavior is because the axial power distributions are consistently peaked towards the lower half of the core, causing the coolant to reach near-saturation conditions under bounding conditions at a core elevation lower than for typical PWR designs. © Copyright 2019 by NuScale Power, LLC 38

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-3 Axial corrosion profile 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 9.1.4) 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. The secondary effects of fuel rod bow 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 any changes from current AREVA PWR fuel assembly designs that might adversely impact rod bow. Operating plant rod bow data for current AREVA PWR designs continue to be adequately covered by the existing rod bow correlation methodology. © Copyright 2019 by NuScale Power, LLC 39

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 (CHF) based on the NRC-approved methodology for quantifying fuel rod bowing and its effects, demonstrated to be applicable to the NuScale design in Reference 9.1.7. 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 9.1.6) 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 growth analysis is performed for the life of the fuel, in which a representative equilibrium fuel cycle yields an average fuel assembly burnup of (( }}2(a)(c),ECI GWd/mtU with a maximum fuel rod burnup of (( }} 2(a)(c),ECI GWd/mtU. Significant margin exists for this representative cycle to allow alternate fuel cycle designs with higher burnups consistent with the approved methodology in Reference 9.1.6. The minimum clearance between the NuScale 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-0816-51127-NP Rev. 2 [

The analysis demonstrates significant margin for a representative equilibrium fuel cycle with a maximum average assembly fluence of ((

        }}2(a)(c),ECI and a maximum fuel rod fluence of ((
        }}2(a)(c),ECI. Significant margin exists to allow alternate fuel cycle designs with higher burnups consistent with the approved methodology in Reference 9.1.6.

[

The minimum clearance between the fuel rod and the top and bottom nozzles, i.e., total shoulder gap (top plus bottom clearance), is [ ] inches, at EOL hot conditions. 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 AREVA 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 NuScale fuel design in-air at BOL and EOL conditions. Similar lateral stiffness tests have been conducted for current AREVA 17x17 fuel designs at EOL and BOL conditions. The test results show that the NuScale fuel assembly lateral stiffness is more than [ ] times greater than that of the current AREVA 17x17 fuel design. With this level of lateral stiffness and significantly lower hydraulic loads on the fuel assembly due to natural circulation flow, the NuScale 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 NuScale fuel design has the same structural components and fuel rod cladding diameters and material (producing similar slip loads and growth rates) as the current AREVA 17x17 PWR designs; thus, the guide tube stresses from fuel rod and guide tube differential growth are bounded by AREVA design experience. The differential growth stresses imparted to the guide tubes are reduced compared to AREVA experience due to 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 AREVAs recent PWR 17x17 fuel design experience. Operating experience of the current AREVA 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. © Copyright 2019 by NuScale Power, LLC 41

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 4.1.8 Fuel Rod Internal Pressure Design Criterion The internal gas pressure of the peak fuel rod in the reactor shall be limited to a value below that which would cause: (1) the fuel-cladding gap to increase due to outward cladding creep during steady-state operation, and (2) extensive DNB (critical heat flux) propagation to occur. (Reference 9.1.1) NuFuel-HTP2' Design Evaluation A bounding analysis is performed for NuScale core designs using the COPERNIC fuel rod analysis code (Reference 9.1.1) for UO2 rods with and without Gd2O3. The maximum fuel rod internal pressure is conservatively compared to a design limit equal to system pressure (1850 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 DNB 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 [ ] psia compared to a limit of 1850 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 9.1.6) 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 [       ] 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 NuScale plant design relies on natural circulation of the coolant without any mechanical pumps, the assumed flow rates envelope all operating conditions and AOOs.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 9.1.6) NuFuel-HTP2' Design Evaluation Fuel rod internal hydriding is controlled by fabrication limits for fuel pellet moisture. These controls, typical for AREVA 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. (References 9.1.2 and 9.1.9) NuFuel-HTP2' Design Evaluation The cladding creep collapse analysis is performed using the methodology of Reference 9.1.2, extended to M5 applications in Reference 9.1.9. A bounding analysis of the NuScale core design is performed using the COPERNIC fuel rod analysis code and the CROV creep ovalization code for both UO2 and Gd2O3 rods. Reference 9.1.2 establishes the three collapse criteria to be analyzed: bifurcation buckling pressure, yield stress, and 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. The CROV code applies the initialization parameters from COPERNIC along with the fuel rod geometry (outside diameter, wall thickness, and ovality) to simulate the cladding creep-down deformations versus time. Consistent with Reference 9.1.2, 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 NuScale bifurcation buckling pressure limit of [ ] psi is calculated that approximates the limit from the CROV analysis. The CROV analysis does not explicitly © Copyright 2019 by NuScale Power, LLC 43

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 [ ] psi, which indicates that the bifurcation buckling pressure collapse criterion is satisfied with significant margin. Figure 4-4 demonstrates that the fuel rod will not collapse as a result of stress beyond the yield point over the expected three cycle incore life. Figure 4-4 CROV results - cladding stress vs. time Figure 4-5 demonstrates that the fuel rod will not collapse as a result of deformation rate exceeding the limit of 0.1 mil/hr over the expected incore life. © Copyright 2019 by NuScale Power, LLC 44

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-5 CROV results - cladding deformation rate vs. time 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 will not occur. 4.2.3 Overheating of Cladding Overheating of cladding is evaluated in the NuScale Chapter 15 transient analyses and is not addressed in this report. 4.2.4 Overheating of Fuel Pellets Design Criterion Fuel melting during normal operation and AOOs shall be precluded. (Reference 9.1.1) © Copyright 2019 by NuScale Power, LLC 45

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 NuFuel-HTP2' Design Evaluation The centerline fuel melt (CFM) analysis is performed using the methodology of Reference 9.1.1. A bounding analysis of the NuScale core design is performed using the COPERNIC fuel rod analysis code for both UO2 and UO2-Gd2O3 rods. The COPERNIC code predicts the transient linear heat rates (LHRs) where the onset of fuel centerline melting occurs. [

[

Using the calculated LHR limits for each of the fuel rods, a bounding envelope is created, as shown in Table 4-4. Figure 4-6, Figure 4-7, Figure 4-8, Figure 4-9, and Figure 4-10 show the bounding envelopes for both transient cladding strain (TCS) and CFM limits for each fuel rod type. [

                                               ] The LHR limits are used in the NuScale core design.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 4-4 Bounding centerline fuel melt limits Figure 4-6 Centerline fuel melt and transient cladding strain bounding envelopes for UO2 fuel © Copyright 2019 by NuScale Power, LLC 47

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-7 Centerline fuel melt and transient cladding strain bounding envelopes for 2 wt% Gd2O3 fuel Figure 4-8 Centerline fuel melt and transient cladding strain bounding envelopes for 4 wt% Gd2O3 fuel © Copyright 2019 by NuScale Power, LLC 48

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-9 Centerline fuel melt and transient cladding strain bounding envelopes for 6 wt% Gd2O3 fuel Figure 4-10 Centerline fuel melt and transient cladding strain bounding envelopes for 8 wt% Gd2O3 fuel © Copyright 2019 by NuScale Power, LLC 49

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 4.2.5 Excessive Fuel Enthalpy Excessive fuel enthalpy from a reactivity initiated accident is addressed in the NuScale 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 [ ]. Steady-state creep-down and irradiation growth are excluded. (Reference 9.1.1) The fuel melt criterion is stated in Section 4.2.4. NuFuel-HTP2' Design Evaluation The transient cladding strain (TCS) analysis is performed using the methodology of Reference 9.1.1. A bounding analysis of the NuScale 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 [ ]. [

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, Figure 4-7, Figure 4-8, Figure 4-9, and Figure 4-10 show the bounding envelopes for both TCS and CFM limits for each fuel rod type. The LHR limits are used in the NuScale core design. © Copyright 2019 by NuScale Power, LLC 50

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 4-5 Bounding transient cladding strain limits 4.2.7 Bursting Swelling and rupture of the cladding relates to the emergency core cooling system performance evaluation that is part of the NuScale FSAR Chapter 15 analyses and is not addressed in this report. 4.2.8 Mechanical Fracturing The seismic and loss-of-coolant accident (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 emergency core cooling system performance evaluation and is not addressed in this report. 4.3.2 Violent Expulsion of Fuel Because reactivity initiated accidents are addressed in the NuScale FSAR Chapter 15 analyses, they are not addressed in this report. 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. © Copyright 2019 by NuScale Power, LLC 51

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 4.3.4 Fuel Rod Ballooning Burst strain and flow blockage caused by ballooning of the cladding relates to the emergency core cooling system 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 a safe shutdown earthquake or LOCA. Specific acceptance criteria for fuel assembly components are identified in Reference 9.1.5. NuFuel-HTP2' Design Evaluation The external load analysis is performed using the methodology of Reference 9.1.5 and Reference 9.1.8. The analysis demonstrates positive margins to all criteria. 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 9.1.5. Damping coefficients specific to the NuFuel-HTP2' design are summarized in Table 4-6 and defined in Reference 9.1.8. © Copyright 2019 by NuScale Power, LLC 52

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 4-6 Summary of NuScale fuel assembly damping ratios for when the fuel is located in both the NuScale Power Module and the reactor flange tool The following model parameters are established through design-specific characterization testing:

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 [

                                                           ] 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 9.1.8.

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. The [

                     ] is demonstrated in Appendix A of Reference 9.1.8. The dynamic crush test is also used to establish a grid load limit. [
                                         ] The grid loads are reported in Table 4-7.

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 9.1.5. 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 © Copyright 2019 by NuScale Power, LLC 54

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 is 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. [

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 NuScale Power Module (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 a single soil-structure profile (S7). For the certified seismic design response spectrum, considering cracked and uncracked concrete conditions and two bounding reactor module configurations, a total of four variations are examined. 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 15%. The frequency shift due to the 15% 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. © Copyright 2019 by NuScale Power, LLC 55

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 NuScale design 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 9.1.5. In addition to the excitation inputs considered for the NPM, seismic core plate motions were also evaluated for the condition where the reactor core is stored in the reactor flange tool (RFT) during refueling. The RFT evaluation is based on the same limiting seismic motion considered for the NPM analysis and considers the same uncertainties, in the form of frequency shifts. There is no LOCA excitation considered for the RFT analysis. 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 (refer to the reactor core configuration in 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 4-7. The peak impact loads for SSE and LOCA in a given direction are combined by the square root of the sum of the squares (SRSS) method and margin is calculated against this SRSS impact load. The positive margin for these impact loads confirms that the NuScale spacer grid will not experience plastic deformation that exceeds the limit established in the AREVA methodology (Reference 9.1.5). Thus, the requirements for core coolability and control rod insertion are met. A lateral seismic analysis of the fuel while the core resides in the RFT is also performed. In this configuration, the core is already contained in a safe shutdown condition and post-LOCA coolability is not a factor. Therefore, the only focus of the RFT seismic analysis is to evaluate the structural integrity of the fuel rod cladding in order to ensure that this fission product barrier is not breached. The results of the RFT lateral analysis provide the load inputs to be considered for this fuel rod evaluation. Table 4-7 Peak grid impact loads and margins © Copyright 2019 by NuScale Power, LLC 56

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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. A bounding vertical seismic load of [ ] is also calculated for the fuel residing in the RFT. This load is considered in calculating the component stresses for the fuel rod in the RFT. 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 9.1.5, [

The results of the component evaluations are presented in Table 4-8. 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. [

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 Level C service limits, consistent with Reference 9.1.5.

The guide tubes are also evaluated for buckling under compressive loads. 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 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 Level C service limit and is established through strength testing. The allowable load for the bottom nozzle connection is based on ASME 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. Due to 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. © Copyright 2019 by NuScale Power, LLC 58

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Table 4-8 Component evaluation margins 4.4 Thermal Hydraulic Evaluation 4.4.1 Core Pressure Drop Evaluation An evaluation is performed of the fuel assembly pressure drop characteristics. The recoverable and unrecoverable pressure drop is determined for each of the 37 fuel assemblies in the reactor core over a wide range of 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 due to elevation change. For the full power best-estimate case, the channel with the greatest unrecoverable pressure drop has an © Copyright 2019 by NuScale Power, LLC 59

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 elevation loss approximately [ ] times greater than the unrecoverable pressure drop. The overall core total pressure drop for the best estimate case is [

Figure 4-11 shows the difference in total pressure drop by channel relative to the average total pressure drop for the core. As shown in the figure, each assembly is represented by one channel. Figure 4-12 shows the difference in the unrecoverable pressure drop by channel relative to the average unrecoverable pressure drop for the core. Figure 4-13 shows the cumulative unrecoverable pressure drop in the channels with the maximum and minimum unrecoverable pressure drop. The results are based on operation at 100 percent power, 100 percent flow with best-estimate assembly-by-assembly distributions for core inlet flow, core inlet temperature, and assembly radial power. © Copyright 2019 by NuScale Power, LLC 60

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 4-13 Cumulative unrecoverable pressure drop - best estimate 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 - the normal parked position and the power dependent insertion limit. 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 conditions for flow, temperature, pressure, and assembly power. 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 adjacent subchannel, by conduction through the guide tube wall and convection, is modeled, consistent with the temperature gradient. © Copyright 2019 by NuScale Power, LLC 63

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Conservative analyses are performed by evaluating normal operation as a function of power and RCS flow rate. Core parameter uncertainties are considered in the areas of [

The guide tube flow rate analysis determines the coolant flow in the guide tube 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 into the guide tube 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 will not occur at any location within the guide tube under steady state conditions. The results indicate that the 100 percent power case is limiting. 4.4.3 Control Rod Drop Analysis The control rod drop analysis predicts the insertion rate and impact velocity of the control rod assembly 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 NuScale 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. A best-estimate, mechanistic model is used to evaluate the 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 are based on control rod drop measurements from a 17x17 PWR plant with a similar fuel geometry. Coolant flow velocity through the guide tube is conservatively assumed to be zero. 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 2.32 ft/sec, which is below the impact velocity limit, and a drop time of 1.08 sec, where drop time is defined as the time between the start of rod movement and the time of full insertion. Figure 4-14 shows axial position versus time for the CRA drop based on a starting position of 79.280 inches from the top of the fuel assembly bottom nozzle to the lower tip of the control rod. Figure 4-15 shows CRA velocity versus time. For reactivity insertion use in the plant safety analyses, the best estimate curve of Figure 4-14 is conservatively modified to bound potential plant conditions. © Copyright 2019 by NuScale Power, LLC 64

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 80 70 60 Position, inches 50 40 30 20 10 0 0 0.2 0.4 0.6 0.8 1 1.2 Time, seconds Figure 4-14 Control rod position versus time 14 12 10 Velocity, ft/sec 8 6 4 2 0 0 0.2 0.4 0.6 0.8 1 1.2 Time, seconds Figure 4-15 Control rod velocity versus time © Copyright 2019 by NuScale Power, LLC 65

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 5.0 Fuel Assembly Testing 5.1 Mechanical Testing Summary A comprehensive test program was conducted at AREVAs 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 beginning-of-life (BOL) and end-of-life (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 9.1.5. 5.1.1 Fuel Assembly Lateral Load Deflection (Stiffness) Test (Beginning of Life and End of Life) This test is performed to characterize the static, lateral structural response of the fuel assembly. 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 (Beginning of Life and End of Life) This test is performed to characterize the dynamic, lateral response of the fuel assembly over a large range of amplitudes. 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 (Beginning of Life and End of Life) This test is performed to characterize the dynamic, lateral impact behavior of the fuel assembly. 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 © Copyright 2019 by NuScale Power, LLC 66

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 is obtained in 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 (Beginning of Life and End of Life) This test is used to characterize the dynamic, lateral response of the fuel assembly. 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/or 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/or 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 (Beginning of Life and End of Life) This test is performed to characterize the static, axial structural response of the fuel assembly. 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 any lateral movement of the fuel assembly. 5.1.6 Fuel Assembly Drop Test (Beginning of Life and End of Life) This test is performed to characterize the dynamic, axial structural response of the fuel assembly. 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 were confirmed through a series of structural tests on prototype grids. © Copyright 2019 by NuScale Power, LLC 67

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 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 AREVA NP Richland Test Facility. The testing configuration simulated the upper and lower core supports in the NuScale Power Module. 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 flow in the © Copyright 2019 by NuScale Power, LLC 68

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 NuScale reactors is driven by natural circulation and at nominal flow conditions, the Reynolds number is approximately 80,000. 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. This pressure loss coefficient incudes a correction for the assembly residual momentum factor based on the liftoff test results. 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°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 © Copyright 2019 by NuScale Power, LLC 69

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 6.0 Control Rod Assembly 6.1 Control Rod Assembly Description The NuScale CRA includes 24 individual control rods fastened to a one-piece cast stainless steel spider and coupling hub (see 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 AREVAs standard rod cluster control assembly (RCCA) design, the individual control rods are shortened to match the NuScale reactor core height, but retain the same basic design features and materials. AREVAs 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 (see 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 (Ag-In-Cd) bar. A stack support resides within the annulus of the lowermost Ag-In-Cd absorber to reduce compressive loads on the Ag-In-Cd, thus reducing thermal creep of the Ag-In-Cd during operation. The control rod cladding is 304 stainless steel tubing with stainless steel end plugs welded to each end, encapsulating the rod internals to complete the rod assemblies (see Figure 6-3). The only design differences between the AREVA standard RCCA and the NuScale CRA are the length of the rod components, the cladding material, and spider spring and retainer modifications. AREVA operating experience has identified two life-limiting phenomena for control rod assemblies: cladding strain and cladding wear. Strain behavior of the NuScale 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 NuScale CRA given the low axial flow rates of the NuScale reactor. The changes to the spider spring and retainer design relative to the standard AREVA RCCA design are made to increase the spring preload and decrease the spring solid © Copyright 2019 by NuScale Power, LLC 70

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 height. These changes allow the CRA spider to absorb the energy from a higher impact force than is experienced in typical PWR applicactions. 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 Ag-In-Cd 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 Ag-In-Cd outer diameter (inch) 0.336 Ag-In-Cd 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 © Copyright 2019 by NuScale Power, LLC 71

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 Figure 6-3 NuScale control rod design 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 Ag-In-Cd 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 Ag-In-Cd bar, where the fluence is highest
Bottom of the solid Ag-In-Cd 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, for swelling rates corresponding to 20 EFPY of operation, the absorbers do not contact the cladding and the cladding strain limit is not exceeded.

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 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 with substantial margin the system pressure of 1850 psia and the maximum hydrostatic test pressure of 2640 psia respectively.

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
Flow-induced vibration
Shipping and handling conditions (evaluated in isolation from other conditions)
Stuck rod condition (evaluated in isolation from other conditions)

The design stress intensity, Sm, is 2/3 of the cladding yield strength [

                                      ]. Table 6-3 defines the allowable stresses for each of the four ASME stress categories. The calculated stresses result in a minimum design margin of

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 1.47, which occurs for primary membrane stresses at cold conditions, where margin is defined as the allowable stress divided by the calculated stress. 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 [ ] 6.2.4 Cladding Fatigue The control rod cladding is analyzed for fatigue from reactor trips, stepping loads, and flow-induced vibration bending loads, assuming no wear. The analysis conservatively assumes infinite cycles from flow-induced vibration 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
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 NuScale design to determine an allowable wear-based design life. Prototype testing using a full-scale CRA © Copyright 2019 by NuScale Power, LLC 76

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 is performed to assess CRA susceptibility to wear. 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 NuScale control rods 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 [ ] 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 releases the 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 Ag-In-Cd 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. The helium generation is conservatively calculated based on the assumption that the lower portion of the B4C pellet stack is subject to the fluence at the lower tip of the pellet stack. For a 20 EFPY CRA design life, the predicted 10B depletion is [ ] percent over the lower portion of the pellet stack. The maximum rod internal pressure at EOL is predicted to be [ ] psia, which meets the criterion of being less than RCS pressure (1850 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. 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: © Copyright 2019 by NuScale Power, LLC 77

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2

Reactor trip / 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 / 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 seismic loads. 6.2.9 Control Rod Assembly Impact Velocity Limit The kinetic energy absorption capacity of the CRA spring is analyzed to determine the maximum impact velocity of the CRA on the fuel assembly during a reactor trip. The NuScale design has a longer and heavier CRA driveline than is typical for PWRs and has significantly lower axial flow rates, resulting in a higher CRA impact velocity. The spider spring is designed to absorb all of 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 acceptable impact velocity is [ ] ft/sec. Section 4.4.3 describes a calculation performed to determine the impact velocity. The calculated velocity is less than the maximum acceptable velocity, justifying the spring design. Testing will confirm that the calculated maximum velocity is conservative. 6.3 Control Rod Assembly Testing The NuScale 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 will be 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 predicted to occur during a concurrent LOCA and seismic event. © Copyright 2019 by NuScale Power, LLC 78

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 7.0 Design Change Process Reference 9.1.6 provides generic design criteria for fuel assembly designs and a process to demonstrate compliance. Compliance to these criteria is demonstrated by the following:

documenting the fuel system and fuel assembly design drawings
performing analyses with NRC-approved models and methods
confirming the adequacy of significant new design features using prototype tests or lead test assemblies prior to full reload implementation
continuing irradiation surveillance programs, including past 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 quality assurance program The generic design change process described in Reference 9.1.6 and the specific design criteria in this report will be used to justify fuel design changes for the NuScale design without requiring NRC review and approval. Acceptable changes to the NuFuel-HTP2' fuel design will meet all of the following conditions:
The change does not result in an un-reviewed safety question.
Changes in plant technical specifications are not required.
The applicability of NRC-approved methodologies is demonstrated to be valid.
Burnup limits are within those approved by the NRC.

Reference 9.1.11 contains examples of design changes that could be made using the change process defined in Reference 9.1.6 without NRC review and approval. NRC concurrence with this process is documented in Reference 9.1.12. The following list summarizes the examples provided in Reference 9.1.11. Both minor design changes and new fuel designs fall within the scope of the design change process such that as long as the criteria continue to be met, subject to the limitations above, design changes can be made without NRC review and approval. Examples of minor design changes are:

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
the first NuScale use of an assembly design feature previously irradiated in a different lattice
a change in enrichment

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2

a change in gadolinia-bearing rod locations Examples of design changes that constitute new fuel designs are:
new cladding material
a spacer grid with a new functional 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-0816-51127-NP Rev. 2 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 will perform acceptably in the NuScale Power Module. © Copyright 2019 by NuScale Power, LLC 81

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 9.0 References 9.1 Source Documents 9.1.1 ANP-10231PA-01, COPERNIC Fuel Rod Design Computer Code, January 2004. 9.1.2 ANP-10084PA-03, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse (CROV computer code), October 1980. 9.1.3 ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Subsection NG, Core Support Structures, 2010 Edition with 2011a Addenda, July 1, 2011. 9.1.4 XN-75-32 (P) (A), Supplements 1-4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983. 9.1.5 ANP-10337P-A, Rev. 0, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, April 2018. 9.1.6 EMF-92-116(P)(A), Revision 0, Generic Mechanical Design Criteria for PWR Fuel Designs, February 2015. 9.1.7 TR-0116-20825-P, Applicability of AREVA Fuel Methodology for the NuScale Design, Revision 1. 9.1.8 TR-0716-50351-P, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, Revision 1. 9.1.9 BAW-10227P-A,Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, June 2003. 9.1.10 NUREG-0800, U.S. NRC Standard Review Plan Section 4.2 Rev. 3, Fuel System Design, March 2007. 9.1.11 Letter, James F. Mallay (Framatome ANP) to Document Control Desk (NRC), EMF-92-116(P), Generic Mechanical Design Criteria for PWR Fuel Designs, NRC:99:029, July 9, 1999. 9.1.12 Letter, Stuart A. Richards (NRC) to James F. Mallay (Framatome ANP), Siemens Power Corporation Re: Request for Concurrence on Safety Evaluation Report Clarifications (MA6160), November 3, 2000. 9.1.13 ODonnell, W.J. and B.F. Langer, Fatigue Design Basis for Zircaloy Components, Nuclear Science and Engineering, Volume 20, pp. 1-12, September 1964. 9.1.14 TR-0916-51502, "NuScale Power Module Seismic Analysis," April 2019, Revision 2. © Copyright 2019 by NuScale Power, LLC 82

NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-0816-51127-NP Rev. 2 9.1.15 TR-1016-51669, "NuScale Power Module Short-Term Transient Analysis," December 2016. © Copyright 2019 by NuScale Power, LLC 83

LO-0719-66433 : Affidavit of Zackary W. Rad, AF-0719-66434 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, 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 method by which NuScale develops its fuel design. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled NuFuel-HTP2' Fuel and Control Rod Assembly Design, TR-0816-51127, Revision 2. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon AF-0719-66434 Page 1 of 2

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 July 30, 2019. Zackary W. Rad AF-0719-66434 Page 2 of 2

LO-0719-66433 : Affidavit of Nathan E. Hottle, Framatome Inc. NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

                                       ) ss.

CITY OF LYNCHBURG )

1. My name is Nathan E. Hottle. I am Manager, Product Licensing, 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 the following document: TR-0816-51127-P Rev. 2, "NuFuel-HTP2' Fuel and Control Rod Assembly Designs,"

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 Inc. 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 Framatome's 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 paragraphs 6(b), 6(c) and 6(d) above.

7. In accordance with Framatome's 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.

SUBSCRIBED before me this - - - - -

                                    , 2019.

Heidi Hamilton Elder NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 12/31/2022 Reg. # 7777873}}

ML19211C901

Text

SUBJECT:

REFERENCES:

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