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LLC Submittal of NuScale Power Module Short-Term Transient Analysis, TR-1016-51669, Revision 1
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LO-0719-66357 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 NuScale Power Module Short-Term Transient Analysis, TR-1016-51669, Revision 1

REFERENCES:

1. Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Technical Reports Supporting the NuScale Design Certification Application (NRC Project No. 0769), dated December 30, 2019 (ML17005A112)

2. NuScale Technical Report, NuScale Power Module Short-Term Transient Analysis, TR-1016-51669, Revision 0, dated December 2016 (ML17005A132)

NuScale Power, LLC (NuScale) hereby submits Revision 1 of the NuScale Power Module Short-Term Transient Analysis, (TR-1016-51669). contains the proprietary version of the report titled NuScale Power Module Short-Term Transient Analysis. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. Enclosure 2 contains the nonproprietary version of the report.

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

If you have any questions, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Marieliz Vera, NRC, OWFN-8H12 : NuScale Power Module Short-Term Transient Analysis, TR-1016-51669-P, Revision 1, proprietary version : NuScale Power Module Short-Term Transient Analysis, TR-1016-51669-NP, Revision 1, nonproprietary version : Affidavit of Zackary W. Rad, AF-0719-66358 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0719-66357 :

NuScale Power Module Short-Term Transient Analysis, TR-1016-51669-P, Revision 1, 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-0719-66357 :

NuScale Power Module Short-Term Transient Analysis, TR-1016-51669-NP, Revision 1, 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

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Technical Report NuScale Power Module Short-Term Transient Analysis July 2019 Revision 1 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical 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.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical 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.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report CONTENTS Abstract ....................................................................................................................................... 1 Executive Summary .................................................................................................................... 2 1.0 Introduction ..................................................................................................................... 3 1.1 Purpose ................................................................................................................. 3 1.2 Scope .................................................................................................................... 3 1.3 Abbreviations ......................................................................................................... 4 2.0 Background ..................................................................................................................... 6 2.1 Regulatory Requirements ...................................................................................... 6 2.2 Regulatory Guidance ............................................................................................. 7 2.3 Modeling Approaches from Literature.................................................................... 9 2.4 NuScale Breach Sizes, Locations, and Exclusion Zones .................................... 11 2.5 Recommended Mechanical Design Transient Analysis Methodology ................. 16 2.6 Recommended Test Data for Benchmarking of Analysis Methodology ............... 17 3.0 Validation Methods of the Short-Term Analysis Methodology.................................. 21 3.1 NuScale Design Basis, Important Loads, and Benchmarking Test Matrix ........... 21 3.2 Heissdampf Reactor Experiments ....................................................................... 23 3.3 Bettis Hydraulic Pressure Pulse Experiment ....................................................... 41 3.4 Marviken Jet Impingement Test Experiment........................................................ 45 4.0 Validation Analysis........................................................................................................ 49 4.1 Thermal Hydraulic Analyses ................................................................................ 49 4.2 Mechanical Dynamic Analyses ............................................................................ 51 5.0 Validation Conclusions ................................................................................................. 56 5.1 NuScale Power Module Modeling Guidelines ..................................................... 56 6.0 NuScale Power Module Asymmetric Cavity Pressurization and Blowdown ........... 60 6.1 NRELAP5 Boundary Conditions for Asymmetric Cavity Pressurization and Blowdown ............................................................................................................ 60 6.2 ANSYS Analysis for Asymmetric Cavity Pressurization and Blowdown .............. 72 7.0 References ..................................................................................................................... 89 7.1 Source Documents .............................................................................................. 89 7.2 Referenced Documents ....................................................................................... 89 Appendix A. NRELAP5 Heissdampf Reactor and Jet Impingement Test Results ......... 92 Appendix B. Pressure Comparison for Bettis Hydraulic Pressure Pulse..................... 107

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report Appendix C. Thrust Force and Fluid Acceleration Boundary Conditions ..................... 111 Appendix D. ANSYS Heissdampf Reactor Results with Pressure Boundary Condition ................................................................................................. 114 Appendix E. Heissdampf Reactor V29.2 ANSYS Results, Flow Acceleration Boundary Condition ................................................................................ 116 Appendix F. Heissdampf Reactor V31.1 ANSYS Results, Flow Acceleration Boundary Condition ................................................................................ 118 Appendix G. Heissdampf Reactor V32 ANSYS Results, Flow Acceleration Boundary Condition ................................................................................ 124 TABLES Table 1-1 Abbreviations ......................................................................................................... 4 Table 1-2 Definitions .............................................................................................................. 5 Table 2-1 NuScale high-energy pressure boundary breaches ............................................ 11 Table 2-2 Reactor pressure vessel and containment vessel nozzle schedule for breach location identification ............................................................................... 12 Table 2-3 Comparison of Heissdampf reactor and NuScale blowdown properties ............. 18 Table 3-1 Phenomena and parameters for benchmarking .................................................. 23 Table 3-2 Comparison of Heissdampf reactor test conditions (Table 3 and 9 of Reference 7.2.5) .................................................................................................. 24 Table 3-3 Heissdampf reactor geometric input parameters................................................. 27 Table 3-4 Time step as a function of node length and speed of sound ............................... 30 Table 3-5 Measured break opening time (Table 4 of Reference 7.2.5) ............................... 30 Table 3-6 Summary of sensitivity parameters for Heissdampf reactor V31.1...................... 31 Table 3-7 ANSYS validation matrix - flow acceleration boundary condition ....................... 38 Table 3-8 ANSYS validation matrix - acoustic pressure boundary condition ...................... 38 Table 3-9 Dynamic responses for benchmarking ................................................................ 40 Table 3-10 Bettis hydraulic pressure pulse test parameters and geometry (Table I of Reference 7.2.6) .............................................................................................. 42 Table 3-11 Comparison of Marviken test conditions (Tables 2-2 and 2-4 of Reference 7.2.25) ................................................................................................ 45 Table 4-1 Summary of optimal modeling parameters for Heissdampf reactor benchmarking cases ........................................................................................... 50 Table 4-2 Summary of optimal modeling parameters for Marviken jet impingement test 11 benchmarking cases ................................................................................ 50 Table 4-3 Summary of peak pressures (psia) for Bettis hydraulic pressure pulse test simulations .................................................................................................... 51 Table 5-1 Modeling parameters for the NuScale break locations ........................................ 57 Table 5-2 Heissdampf reactor, Marviken, and NuScale high-energy line break comparison .......................................................................................................... 59 Table 6-1 Critical mass flow rate theoretical, expected, and NRELAP5 results .................. 63 Table 6-2 Critical mass flow rate theoretical, expected and NRELAP5 results ................... 69 Table 6-3 Cases for valve opening and break locations ...................................................... 73 Table 6-4 Maximum forces and moments at component interfaces .................................... 77

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report Table 6-5 List of component interfaces for force and moment generation .......................... 79 Table 6-6 Maximum forces and moments on containment vessel, reactor pressure vessel, riser, and core barrel assembly ............................................................... 83 Table 6-7 List of component interfaces for force and moment generation .......................... 84 Table 6-8 Summary of largest five forces for RPV, CNV, and RVI ...................................... 85 Table 6-9 Summary of largest five moments for RPV, CNV, and RVI ................................. 86 Table 6-10 Largest forces and moments for CVCS injection line break for RPV, CNV and RVI ....................................................................................................... 87 FIGURES Figure 2-1 Comparison of NuScale Power Module and Heissdampf reactor ....................... 19 Figure 2-2 Test schematic for Bettis hydraulic pressure pulse (Fig. 4 of Reference 7.2.6) ................................................................................................................... 20 Figure 3-1 NRELAP5 models: example nodalization schematic for Heissdampf reactor experiment .............................................................................................. 25 Figure 3-2 Schematic of the Heissdampf reactor pressure vessel and internals (Fig. 4-1 of Reference 7.2.16) ............................................................................. 26 Figure 3-3 Heissdampf reactor discharge nozzle dimensions and sensor locations (Fig. 6 of Reference 7.2.5) .................................................................................. 28 Figure 3-4 ANSYS finite element analysis model of the Heissdampf reactor pressure vessel and internals .............................................................................. 33 Figure 3-5 ANSYS nozzle end nodes ................................................................................... 35 Figure 3-6 ANSYS finite element analysis model sensor locations ...................................... 39 Figure 3-7 ANSYS finite element analysis model of Bettis hydraulic pressure pulse tests ..................................................................................................................... 44 Figure 3-8 Marviken jet impingement test schematic of test configuration (Figures 2-3 and 2-4 of Reference 7.2.25) ........................................................................ 46 Figure 3-9 NRELAP5 models: example schematics for Marviken experiment ..................... 47 Figure 4-1 Heissdampf reactor V32 depressurization propagation from the break location (at 100 ms) ............................................................................................. 52 Figure 4-2 Heissdampf reactor V32 core barrel deformations (displacement scale factor of 200) ....................................................................................................... 53 Figure 6-1 Schematic for the reactor coolant system subcooled blowdown model .............. 61 Figure 6-2 Flow acceleration boundary condition - chemical and volume control system injection pipe break ................................................................................. 64 Figure 6-3 Thrust force boundary condition - chemical and volume control system injection pipe break ............................................................................................. 65 Figure 6-4 Flow acceleration boundary condition - reactor recirculation valve inadvertent opening ............................................................................................. 66 Figure 6-5 Thrust force boundary condition - reactor recirculation valve inadvertent opening ............................................................................................. 67 Figure 6-6 Schematic for the reactor coolant system saturated blowdown model................ 68 Figure 6-7 Flow acceleration boundary condition- spray and degasification line pipe breaks .......................................................................................................... 69 Figure 6-8 Thrust force boundary condition- spray and degasification line pipe breaks .................................................................................................................. 70

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report Figure 6-9 Flow acceleration boundary condition- inadvertent reactor vent valve opening ................................................................................................................ 70 Figure 6-10 Thrust force boundary condition- inadvertent reactor vent valve opening ........... 71 Figure 6-11 Full Model with Two Acoustic Bodies ................................................................. 74 Figure 6-12 Meshes for two acoustic bodies .......................................................................... 75 Figure 6-13 Differential pressure time history across baffle plate for one reactor recirculation valve (RRV1) opening ..................................................................... 88 Figure A-1 Heissdampf reactor Test V31.1 Sensitivity Case A2, mass flow rate .................. 92 Figure A-2 Heissdampf reactor Test V31.1 Sensitivity Case A2, pressure ........................... 93 Figure A-3 Heissdampf reactor Test V31.1 Sensitivity Case A5, mass flow rate .................. 94 Figure A-4 Heissdampf reactor Test V31.1 Sensitivity Case A5, pressure ........................... 95 Figure A-5 Heissdampf reactor Test V31.1 Sensitivity Case B, pressure ............................. 96 Figure A-6 Heissdampf reactor Test V31.1 Sensitivity Case B2, pressure ........................... 97 Figure A-7 Heissdampf reactor Test V31.1 Sensitivity Case Set B, mass flow rate ............. 98 Figure A-8 Heissdampf reactor Test V31.1 Sensitivity Case B2, mass flow rate .................. 99 Figure A-9 Heissdampf reactor Test V31.1 Sensitivity Case Set C, mass flow rate ........... 100 Figure A-10 Heissdampf reactor Test V31.1 Sensitivity Case Set C, pressure .................... 101 Figure A-11 Mass flow rate for Heissdampf reactor Test V31.1............................................ 102 Figure A-12 Mass flow rate for Heissdampf reactor Test V32............................................... 103 Figure A-13 Mass flow rate for Marviken jet impingement test-11 ........................................ 104 Figure A-14 Density for Marviken jet impingement test-11 ................................................... 105 Figure A-15 Thrust force for Marviken jet impingement test-11 ............................................ 106 Figure B-1 Pressure at top and bottom transducers for Run 10S ....................................... 107 Figure B-2 Pressure at top and bottom transducers for Run 10F ....................................... 108 Figure B-3 Pressure at top and bottom transducers for Run 20S ....................................... 109 Figure B-4 Pressure at top and bottom transducers for Run 20F ....................................... 110 Figure C-1 Flow acceleration at the break location for Heissdampf reactor Test V29.2 ................................................................................................................. 111 Figure C-2 Thrust force at the break location for Heissdampf reactor Test V29.2 .............. 111 Figure C-3 Flow acceleration at the break location for Heissdampf reactor Test V31.1 ................................................................................................................. 112 Figure C-4 Thrust force at the break location for Heissdampf reactor Test V31.1 .............. 112 Figure C-5 Flow acceleration at the break location for Heissdampf reactor Test V32 ........ 113 Figure C-6 Thrust force at the break location for Heissdampf reactor Test V32 ................. 113 Figure D-1 Heissdampf reactor Test V31.1 pressure, BP9109 (1330, 90°, 8850)

(Fig. 4-5 of Reference 7.2.16) ........................................................................... 114 Figure D-2 Heissdampf reactor Test V31.1 pressure, KP0009 (1307, 90°, 8850)

(Fig. 4-12 of Reference 7.2.16) ......................................................................... 114 Figure D-3 Displacement comparison for Heissdampf reactor Test V32, KS1030 (1307, 90°, 2265) (Fig. A-118 of Reference 7.2.16) .......................................... 115 Figure D-4 Displacement comparison for Heissdampf reactor Test V32, KS1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) ........................................ 115 Figure E-1 Displacement at upper part of the core barrel for V29.2, KS1008 (1330, 90°, 8410) (Fig. 26 of Reference 7.2.5) ............................................................. 116 Figure E-2 Outside reactor pressure vessel displacement for V29.2, BS0106 (1590, 90°, 7350) (Fig. 28 of Reference 7.2.5) .................................................. 116

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report Figure E-3 Outside reactor pressure vessel displacement for V29.2, BS0107 (1590, 180°, 7350) (Fig. 28 of Reference 7.2.5) ................................................ 117 Figure E-4 Outside reactor pressure vessel displacement for V29.2, BS0108 (1590, 270°, 7350) (Fig. 28 of Reference 7.2.5) ................................................ 117 Figure F-1 Outside reactor pressure vessel for V31.1, BS0106 (1590, 90°, 7350)

(Fig. 28 of Reference 7.2.5) .............................................................................. 118 Figure F-2 Outside reactor pressure vessel for V31.1, BS0107 (1590, 180°, 7350)

(Fig. 28 of Reference 7.2.5) .............................................................................. 118 Figure F-3 Outside reactor pressure vessel displacement for V31.1, BS0108 (1590, 270°, 7350) (Fig. 28 of Reference 7.2.5) ................................................ 119 Figure F-4 Pressure for V31.1, BP9109 (1330, 90°, 8850) (Fig. 4-5 of Reference 7.2.16) ............................................................................................................... 119 Figure F-5 Pressure for V31.1, BP9117 (1330, 270°, 8850) (Fig. 4-6 of Reference 7.2.16) ............................................................................................................... 120 Figure F-6 Pressure for V31.1, BP9133 (1330, 88°, 5505) (Fig. 4-7 of Reference 7.2.16) ............................................................................................................... 120 Figure F-7 Pressure for V31.1, BP9140 (1330, 90°, 2300) (Fig. 4-8 of Reference 7.2.16) ............................................................................................................... 121 Figure F-8 Pressure for V31.1, BP8301 (0, 0°, 10370) (Fig. 4-10 of Reference 7.2.16) ............................................................................................................... 121 Figure F-9 Differential pressure for V31.1, KP0009 (1307, 90°, 8850) (Fig. 4-12 of Reference 7.2.16) .............................................................................................. 122 Figure F-10 Hoop strain for V31.1, at core barrel outside diameter KA2009 (1330, 90°, 8850) (Fig. 4-15 of Reference 7.2.16) ........................................................ 122 Figure F-11 Axial strain for V31.1, at core barrel outside diameter KA3008 (1330, 90°, 8850) (Fig. 4-16 of Reference 7.2.16) ........................................................ 123 Figure G-1 Outside reactor pressure vessel displacement for V32, BS0106 (1590, 90°, 7350) (Fig. 8 of Reference 7.2.23) ............................................................. 124 Figure G-2 Outside reactor pressure vessel displacement for V32, BS0116 (1590, 90°, 5550) (Fig. A-47 of Reference 7.2.16) ....................................................... 124 Figure G-3 Core barrel displacement for V32, KS1013 (1307, 90°, 7195) (Fig. A-112 of Reference 7.2.16) ................................................................................... 125 Figure G-4 Core barrel displacement for V32, KS1030 (1307, 90°, 2265) (Fig. A-118 of Reference 7.2.16) ................................................................................... 125 Figure G-5 Core barrel displacement for V32, KS1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) ................................................................................... 126 Figure G-6 Core barrel hoop strain for V32, KA2008 (1330, 90°, 8845) (Fig. A-66 of Reference 7.2.16) .............................................................................................. 126 Figure G-7 Core barrel axial strain for V32, KA3009 (1330, 90°, 8825) (Fig. A-71 of Reference 7.2.16) .............................................................................................. 127 Figure G-8 Sensitivity study: core barrel displacement for V32, KA1030 (1307, 90°,

2265) (Fig. A-118 of Reference 7.2.16) ............................................................. 127 Figure G-9 Sensitivity study: core barrel displacement for V32, KA1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) ................................................... 128 Figure G-10 Sensitivity study: core barrel hoop strain for V32, KA2008 (1330, 90°,

8845) (Fig. A-66 of Reference 7.2.16) .............................................................. 128

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Licensing Topical Report Figure G-11 Sensitivity study: core barrel axial strain for V32, KA3009 (1330, 90°,

8825) (Fig. A-71 of Reference 7.2.16) .............................................................. 129

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Abstract The Short-Term Transient Analysis technical report is prepared to supplement the information contained in the Final Safety Analysis Report relative to the dynamic analyses performed to evaluate structural response of the NuScale Power Module (NPM). Short-term transients are events caused by the failure or actuation of piping and valves, and include high-energy line breaks. These events results in system internal pressure waves and asymmetric cavity pressurization waves exterior to the pipe break or valve outlet.

These events require special treatment due to the rapidly changing thermal hydraulic conditions and the resulting dynamic mechanical loads. In addition to the rapid nature of the transients, fluid-structure interactions are influential and consideration is required.

This technical report provides an overview of the analytical methods used to simulate the short-term transient mechanical loads, the benchmarking performed to validate the analysis methods, and the analysis of the short-term transient events for the NPM. Resulting loads to be used as input to component analyses are presented in conclusion.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Executive Summary This document identifies transients involving breaches in pressure boundaries inside and outside containment. Events that involve a breach in a high-energy pressure boundary require special treatment from a mechanical analysis standpoint due to the rapidly changing thermal hydraulic conditions and the resulting dynamic mechanical loads. In addition to the rapid nature of these transients, the fluid-structure interaction (FSI) must be modeled, at least to the extent that a reasonably bounding loading profile can be assured. For American Society of Mechanical Engineers (ASME) Service Level B, C and D events that include these dynamic mechanical loads, the stress analysis must confirm the structural design adequacy and ability, with no loss of safety function, of the reactor vessel internals (RVI), and portions of the reactor coolant pressure boundary (RCPB) that are not compromised, to withstand the loads from breaches in high-energy pressure boundaries in combination with the safe shutdown earthquake as specified in the NuScale Power Module (NPM) component design specifications.

Various commercially available software applications can be used for transient analysis of complicated structures. NuScale has reviewed recent industry experience with similar applications to develop its short-term transient analysis methodology. To provide consistency with other NuScale thermal hydraulic and mechanical applications, the codes NRELAP5 and ANSYS are selected for the dynamic analysis methodology. The thermal hydraulic code NRELAP5 is used to generate boundary conditions (BCs) for the mechanical analysis, and ANSYS is used to simulate the FSI and calculate resultant time history loads. This approach is consistent with literature references.

Benchmarking is required for the dynamic models used to develop the time histories to be applied as input to component stress analyses. The benchmarking results give confidence that the methodology provides acceptable simulation of the dynamic mechanical loads for the events identified. The benchmarking cases used for this analysis are integral and separate effects tests that provide experimental results of both the thermal hydraulic phenomena and the resulting mechanical loads. The results of the thermal hydraulic and mechanical dynamic analyses give a level of confidence that the dynamic loads associated with a short-term transient event are acceptably modeled for the NuScale design using the methodologies described. Results of the benchmarking evaluation demonstrate results are not sensitive to the thermal hydraulic BCs, and an accurate structural response can be generated with known modeling simplifications. As summarized in Section 3.0, the parameters important to dynamic analysis compare favorably with experimental results. Based on the overall favorable agreement, application of a biasing margin on the thermal hydraulic or dynamic analysis results for the NuScale design is not necessary.

Structural loads that result from the high energy breaches are specified as time history forces, moments, in-structure displacements and accelerations, and acoustic pressures. Specified loads bound the loads due to inadvertent or spurious valve operation and piping breaks, and consider both the blowdown and asymmetric cavity pressurization of the containment pressure vessel simultaneously. Resultant loads specified in this technical report contribute to the basis for the mechanical design of the NPM.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 1.0 Introduction 1.1 Purpose The purpose of this report is to document the methodology for analyzing high-energy pressure boundary breaches due to the failure or actuation of piping and valves inside and outside containment. Breaches in the pressure boundary of high-energy systems require special treatment in mechanical design due to the large hydraulic forces that are rapidly generated during a depressurization event.

This report summarizes the high-energy breaches that are analyzed in the NuScale design and the planned analytical methods for simulating associated dynamic mechanical loads.

The analytical methods are benchmarked using experimental results to ensure they are capable of accurately simulating the blowdown phenomena. This report provides quantitative comparisons of the performance of the short-term transient methodology with experimental high-energy line break (HELB) test data. The report demonstrates that the evaluation methodology is applicable to the high energy breaches that must be analyzed in the NuScale design, and provides an overview of the NuScale short-term transient analysis and results.

1.2 Scope The NuScale Power Module (NPM) design includes flow jet diffusers for some valves that discharge fluid into containment. Essential structures, systems, and components are those required to shut down the reactor and mitigate the consequences of a postulated short term transient event. The purpose of jet diffusers, which are provided for the reactor vent valves (RVVs) and potentially the reactor safety valves (RSVs), is to prevent the valve discharge fluid jet from damaging nearby essential SSC.

As an engineering conservatism, the flow jet diffusers were not credited in the short-term transient methodology, benchmarking, or analysis. The reasons for this are as follows.

First, the system loads associated with the unconstrained postulated high energy breaches are low for the NuScale design, as shown in this report, and they do not need to be further reduced.

Secondly, the flow diffusers are designed to significantly disrupt the effluent flow at the valve discharge outlets, compared to the unconstrained flow that has been traditionally analyzed. This disruption reduces the acoustic energy, which minimizes the effects of the pressure wave propagating in the containment. Similar to the internal pressure wave discussion, the system loads associated with asymmetric cavity pressure wave events are demonstrated to be low, so there is little benefit in benchmarking the performance of the flow diffusers to credit a further reduction in the loads associated with asymmetric cavity pressurization waves exterior to the valve outlet.

Additionally, the short-term transient methodology relies on experimental data to benchmark the thermal hydraulic and mechanical analysis methods, to ensure that the simulation of the loads for postulated pipe breaks and valve actuations are sufficiently bounding. Significant research has been performed in the subject area of HELBs, and

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 there are numerous experiments and standard problems available for benchmarking.

However, there is not a significant body of research to benchmark the reduced flow conditions associated with the NuScale high-energy protective devices.

Lastly, the inadvertent actuation block feature of the emergency core cooling system also limits the severity of blowdowns resulting from the inadvertent actuation of one emergency core cooling system valve. Each valves inadvertent actuation block prevents the valve from opening until the differential pressure between the reactor pressure vessel (RPV) and the containment vessel (CNV) decreases below a threshold that reduces the discharge mass flow rate and the accompanying blowdown loads. Conservatively, this analysis ignores the inadvertent actuation block feature and models emergency core cooling system valve blowdowns at the high differential pressure reactor coolant system conditions.

Based on these considerations, the short-term transient methodology, benchmarking, and analysis conservatively neglect these protective devices.

1.3 Abbreviations Table 1-1 Abbreviations Term Definition ASME American Society of Mechanical Engineers BC boundary condition CNV containment vessel CRDM control rod drive mechanism CVCS chemical and volume control system DHRS decay heat removal system FSI fluid-structure interaction GDC General Design Criteria HDR Heissdampf reactor HELB high-energy line break JIT jet impingement test LBB leak-before-break LOCA loss-of-coolant accident NPM NuScale Power Module PWR pressurized water reactor PZR pressurizer RCPB reactor coolant pressure boundary RCS reactor coolant system RPV reactor pressure vessel RRV reactor recirculation valve RSV reactor safety valve RVI reactor vessel internals RVV reactor vent valve SG steam generator SSC structures, systems, and components

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 1-2 Definitions Term Definition Acoustic pressure The difference between the local pressure and equilibrium pressure due to a sound wave.

ANSYS Engineering simulation software used for coupled fluid-structure interaction (FSI) analysis.

Benchmarking Analysis performed to demonstrate that the results of a simulation or hand calculation provide acceptable agreement with experimental results.

Blowdown load A hydraulic load that develops as a result of the transient flow and pressure fluctuations following a valve actuation or a breach in a high-energy pressure boundary.

Essential boundary condition BC in which a dependent variable, such as pressure or temperature, is (BC) applied to a domain boundary. Also called Dirichlet BC.

Essential SSC Essential structures, systems, and components are those required to shut down the reactor and mitigate the consequences of a postulated short term transient event.

Forcing function An externally generated force that acts on a system, and is only a function of time.

Natural BC BC in which a derivative of a dependent variable, such as pressure or temperature, is applied to a domain boundary. Also called Neumann BC.

NRELAP5 NuScale proprietary version of RELAP5-3D thermal hydraulic analysis code.

Pressurization load A hydraulic load that develops as a result of a postulated high-energy pipe break or valve opening.

RELAP5-3D Thermal hydraulic analysis code used for simulation of transients and postulated accidents in light water reactor systems.

Subcompartment A fully or partially enclosed volume within the containment that houses or adjoins high-energy piping systems and restricts the flow of fluid to the main containment volume in the event of a postulated pipe rupture.

Terminal end The extremities of a piping run that connect to structures, components (e.g., vessels, pumps, valves) or piping anchors that act as rigid constraints to piping motion and thermal expansion.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 2.0 Background This short-term transient analysis is provided as a technical report to facilitate the submittal of proprietary and public information to the NuScale docket, in support of the NRC review of the technical information and results.

2.1 Regulatory Requirements Requirements related to analyzing loads due to HELBs are from the General Design Criteria (GDC) of 10 CFR 50, Appendix A and the requirements related to the emergency core cooling modeling in 10 CFR 50, Appendix K. Applicable requirements that are implemented in the short term transient methodology are provided below. Section 2.2 addresses the guidance pertaining to the requirements below.

2.1.1 10 CFR 50 Appendix A, GDC 4 The GDC 4 was considered in the design of SSC to be protected from short term transient dynamic effects. Blowdown and asymmetric cavity pressurization are dynamic effects of a high energy valve actuation or pressure boundary breach that are the subject of this methodology.

2.1.2 10 CFR 50 Appendix A, GDC 14 Compliance with GDC 14 requires that the reactor coolant pressure boundary (RCPB) is designed to have a low probability of abnormal leakage, failure, or rupture.

To meet this requirement, it is to be shown that a breach does not result in further degradation of the RCPB. Appropriate characterization and application of the blowdown and asymmetric cavity pressurization loads as design loads for affected SSC fulfills this requirement.

2.1.3 10 CFR 50 Appendix K, Section I.C.1 Compliance with this requirement is met by considering a spectrum of possible breaches in high-energy pressure boundaries. This spectrum includes instantaneous double-ended breaks ranging in cross-sectional area up to and including that of the largest pipe in the primary coolant system. Analysis also includes the effects of longitudinal splits in the largest pipes, with the split area equal to the cross-sectional area of the pipe.

For the purpose of characterizing blowdown and asymmetric cavity pressurization loads, the type of pipe break (circumferential or longitudinal) is not relevant. Also, for the NuScale design, valve actuations provide a greater flow area than is provided by postulated primary coolant pipe breaks. A spectrum of thermal hydraulic conditions and pipe break locations are considered for each postulated location.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 2.1.4 10 CFR 50 Appendix K, Section I.C.1.b To confirm the acceptability of the thermal hydraulic results, in addition to the benchmarking that is performed, the Moody model is used to confirm the thermal hydraulic results for postulated break locations and valve actuations.

2.1.5 10 CFR 50 Appendix K, Section I.C.1.d In the thermal hydraulic model used for benchmarking, nodalization is chosen to provide agreement with the experimental results. Additionally, nodalization studies are performed for each break location and valve actuation that is simulated to ensure that the nodalization is adequate and does not significantly affect the results.

2.1.6 10 CFR 50 Appendix K, Section I.C.2 Frictional losses and two phase flow multipliers are investigated in the benchmarking and are chosen to provide good agreement with the experimental results. Based on the very short time period that is relevant for calculating the blowdown and asymmetric cavity pressurization loads, these inputs do not have a significant effect on the simulated results.

2.1.7 10 CFR 50 Appendix K, Section I.C.3 The thermal hydraulic analysis of high energy line breaks and valve actuations must adequately characterize momentum changes, pressure losses and acceleration. These features are provided in the thermal hydraulic analysis code NRELAP5. Adequacy for the purpose of characterizing the high-energy pressure boundary breaches is demonstrated in the benchmarking analyses.

2.2 Regulatory Guidance Guidance in the Standard Review Plan Sections 3.6.2, 3.9.2 and 3.9.5, and Branch Technical Position 3-4 are considered in the benchmarking methodology.

2.2.1 NUREG-0609, Asymmetric Blowdown Loads on PWR Primary Systems NUREG-0609 (Reference 7.2.1) provides a historical and technical summary of blowdown load analysis for pressurized water reactors (PWRs), including criteria and guidance for conducting an evaluation.

NUREG-0609 Section 3.1.1 explains that although the effects of a sudden decompression of a PWR primary system can be determined by a straightforward calculation considering the LOCA pressure wave and its resulting interaction with internal structures of the reactor, the physical process is not as simple. The resulting motion of the core barrel to the decompression wave affects the local fluid pressure through the compressibility of the water; and the resulting hydrodynamic loads are typically reduced when this fluid-structure coupling is considered.

NUREG-0609 Section 3.1 identifies that the two steps required for calculating the hydraulic loads due to a blowdown are: (1) determine the transient pressure and velocity

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 distribution throughout the system, using a thermal hydraulic code; and (2) convert the transient pressure and velocity data into equivalent transient forces through the primary system. The transient forces are then used as input, along with other LOCA loads, in the time-history structural analysis of the primary system. The procedure can be broken down into three major categories: analytical development, application and system modeling, and computer-program verification.

Guidelines for developing loading functions for subcooled blowdown and cavity loads are summarized below.

Analytical Development:

Use of homogeneous equilibrium is acceptable since during the modeling time of interest the system fluid is primarily subcooled. However, potential nonequilibrium effects should be considered. Note that for the spectrum of breaches that are analyzed in the NuScale design, not all system fluid is subcooled. Additionally, nonequilibrium effects are simulated using NRELAP5 and the degree of nonequilibrium is investigated as a part of the benchmarking analysis.

The range of pressures and the sonic or acoustic wave speed should be described.

The convergence criteria must ensure conservation of mass, momentum, and energy.

This is important if a 1D code is used to model a multidimensional region.

Essential BCs are the break opening time and area characteristics, and must be justified.

The discharge flow model for the postulated break is the system forcing function. As such, the treatment of the subcooled critical flow and potential nonequilibrium effects must be properly accounted for in the development of a discharge flow model.

If fluid-structure coupling is considered, the method of incorporating the moving BC into the conservation equations must be justified.

Application and System Modeling:

Use of homogeneous equilibrium is acceptable since during the modeling time of interest the system fluid is primarily subcooled. However, potential nonequilibrium effects should be considered.

The ability of the model to track and transmit pressure waves must be demonstrated.

At a minimum, a two-dimensional pressure field is required to adequately evaluate the effects of the decompression waves in an annulus region.

The sensitivity of the model to spatial representation, time-step size, and to the various convergence criteria must be justified.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Code Verification:

The program and modeling procedure must be compared to selected problems and experimental data to demonstrate that the simulation provides good agreement with the phenomena.

A comparison of the code performance with test data covering a wide range of system geometries is required. A Heissdampf reactor (HDR) test facility analysis is required as part of code verification. A partial list of additional acceptable experimental data is provided in References 7.2.9 through 7.2.15.

Section 3.2 of Reference 7.2.1 provides a discussion of the analysis of asymmetric cavity pressurization loads. A subcompartment is defined as a fully or partially enclosed volume within the containment that houses or adjoins high-energy piping systems and restricts the flow of fluid to the main containment volume in the event of a postulated pipe rupture.

Following a pipe rupture, a pressure wave develops within the cavity, generating loads on the components similar to those generated on the reactor vessel internals (RVI).

Reference 7.2.1 states the subcompartment pressure analyses are to be performed to determine asymmetric pressure loading on components and subcompartment walls, to ensure that the walls and component supports can withstand the forces and the reactor can be brought to a safe shutdown condition.

For the NuScale design, there are no subcompartments within the containment vessel (CNV). However, the containment itself is a relatively small annular region in which a pressure wave could form and generate asymmetric loading on components inside the CNV. The loading on the components due to the pressure wave is bounded by other dynamic events, such as seismic and the blowdown transient, when the peak differential pressure across the vessel is reached.

Reference 7.2.1 states that codes like COMPARE and RELAP-4 MOD5 are typically used for subcompartment analyses. If other codes are used, confirmatory analysis is required.

Additionally, Reference 7.2.1 provides guidelines for subcompartment nodalization and input assumptions that must be investigated via sensitivity studies.

2.3 Modeling Approaches from Literature The simulation of loads generated during a LOCA has been the subject of increased academic research, based on the use of commercially available software to accurately simulate the loads, unlike the proprietary software solutions developed in the 1980s (e.g.,

MULTIFLEX or CRAFT-2). The following sections provide an overview of recent modeling approaches and a summary of their performance compared to the benchmark test data, where applicable.

2.3.1 R5FORCE R5FORCE is a computer program developed by the Idaho National Laboratory that uses the output of RELAP5 to generate forcing functions for structural analysis. A description of the code and the user manual are documented in Reference 7.2.2. Using the hydrodynamic outputs of RELAP5, R5FORCE solves the force equation using the

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 pressure and wall shear force terms. It is considered an improvement over legacy methods, which use pressure and fluid acceleration terms, because the use of the shear wall terms instead of the fluid acceleration terms eliminates numerical instabilities associated with computing the time derivative of the fluid acceleration term. The forcing functions generated by R5FORCE are intended for input into structural analysis codes, such as NUPIPE, SAP, or ADINA.

This code has not been used extensively in industry, and benchmarking results against experimental data are unknown. Based on advances in structural analysis codes, there does not appear to be an advantage in using an intermediate code to generate forcing functions. Modern structural analysis codes are capable of accepting thermal-hydraulic BCs and simulating FSI.

2.3.2 Computational Fluid Dynamics coupled with Finite Element Modeling References 7.2.3 and 7.2.8 use computational fluid dynamics coupled with finite element modeling codes to simulate the FSI and resulting structural responses. Both evaluations use pressure BCs from the two-phase system analysis code, APROS. In both evaluations, the pressure at a point in the nozzle exit without significant voiding is taken from the system analysis code and is applied as a BC in the computational fluid dynamics model.

References 7.2.3 and 7.2.8 show good agreement with the structural response of the core barrel region for the first 100ms. Both evaluations attribute the simulation errors after 100ms to the onset of two phase flow, which is not modeled with the computational fluid dynamics.

2.3.3 Acoustic-Structural Model with Pressure Boundary Condition Another common analytical method is using an acoustic-structural model. Reference 7.2.3 provides an acoustic-structural model with a pressure BC from APROS. Results are only in good agreement for 20ms following the transient initiation. Reference 7.2.3 attributes the deviations at larger times to the larger value of the dynamic pressure compared to the stagnation pressure. Dynamic pressure is not accounted for in the acoustic model since acoustic elements do not have a velocity.

2.3.4 Acoustic-Structural Model with Mass Flow Boundary Condition Reference 7.2.4 improves on the acoustic-structural modeling approach by considering a different pipe break BC. Reference 7.2.4 uses a natural BC based on a mass flow rate from APROS. Results indicate good agreement with the experimental results beyond 100ms.

2.3.5 Modeling Approach Selected for NuScale Analysis The acoustic-structural modeling approach provides the highest accuracy and is the simplest analytical method of the literature approaches that have been investigated. The acoustic-structural modeling and BCs discussed in References 7.2.3 and 7.2.4 are investigated for the NuScale design in Section 4.0.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 2.4 NuScale Breach Sizes, Locations, and Exclusion Zones 2.4.1 Breach Sizes and Locations The piping breaks and other breaches of high-energy pressure boundaries that are included within the mechanical design-basis are identified in Table 2-1. An overview of the postulated breach locations is provided in Table 2-2. Further detail and justification are provided in Sections 2.4.1.1 through 2.4.1.12.

As discussed in Section 1.2, the NPM design may contain flow jet diffusers at the RSV and RVV discharge outlets. The purpose of jet diffusers, which are provided for the RVVs and may be provided for the RSVs, is to prevent the valve discharge fluid jet from damaging nearby essential SSC. As a bounding engineering simplification, these protective components are not credited in the short-term transient methodology, benchmarking, or analysis and are omitted in the discussion of the breach locations in the following sections.

Table 2-1 NuScale high-energy pressure boundary breaches Plant Wide Event Applicable Acoustic Loads Service Level Inadvertent opening of an RSV Blowdown, Asymmetric Pressurization Spurious RVV actuation Blowdown, Asymmetric Pressurization Service Level C Spurious reactor recirculation valve Blowdown, Asymmetric Pressurization (RRV) actuation Chemical and volume control system Blowdown, Asymmetric Pressurization pipe break Steam generator (SG) tube failure N/A Steam piping failures Blowdown Service Level D Feedwater piping failures Blowdown Control rod assembly ejection N/A

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 2-2 Reactor pressure vessel and containment vessel nozzle schedule for breach location identification Number of Approximate Nozzle Reactor Size, Diameter Number RPV Nozzle Pressure Analyzed of CNV Description Region Vessel (Yes/No)

Nozzles RPV CNV Location (RPV)

Nozzles Reactor Upper RPV Yes, Section 2 N/A recirculation 2.25 in. N/A shell 2.4.1.3 valve (RRV) 1 - 2 Feedwater nozzle Upper RPV Yes, Section 4 2 NPS 4 NPS 5 1-4 shell 2.4.1.7 Main steam Upper RPV Yes, Section 4 4 NPS 8 NPS 12 nozzle 1-4 shell 2.4.1.6 Reactor coolant Upper RPV Yes, Section 1 1 system (RCS) NPS 2 NPS 2 shell 2.4.1.4 injection Upper RPV Yes, Section 1 1 RCS discharge NPS 2 NPS 2 shell 2.4.1.4 Pressurizer spray Yes, Section 2 1 NPS 2 NPS 2 RPV head supply 1 - 2 2.4.1.4 Yes, Section 3 N/A RVV 1 - 3 4.875 in. (1) N/A RPV head 2.4.1.2 Reactor safety Yes, Section 2 N/A NPS 3 N/A RPV head valve 1 - 2 2.4.1.1 RPV high point Yes, Section 1 1 NPS 2 NPS 2 RPV head degasification 2.4.1.4 Pressurizer No, Section 2 N/A (PZR) heater 23 in. N/A PZR shell 2.4.1.8 port 1- 2 Control rod drive mechanism No, Section 16 N/A 2.375 in. N/A RPV head (CRDM) nozzles 2.4.1.9 1-16 Instrumentation No, Section 4 4 and control - NPS 8 NPS 8 RPV head 2.4.1.10 channels A-D Feedwater Upper RPV No, Section 4 N/A plenum access 25 in. N/A shell 2.4.1.8 port 1-4

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Number of Approximate Nozzle Reactor Size, Diameter Number RPV Nozzle Pressure Analyzed of CNV Description Region Vessel (Yes/No)

Nozzles RPV CNV Location (RPV)

Nozzles Main steam Upper RPV No, Section 4 N/A plenum access 20 in. N/A shell 2.4.1.8 port 1-4 Instrumentation No, Section numerous N/A lines and 0.875 in.(2) N/A various 2.4.1.12 thermowells UT sensor nozzle Upper RPV No, Section 4 N/A 14 in. N/A 1-4 shell 2.4.1.10 decay heat No, Section N/A 2 N/A NPS 2 N/A removal 1 - 2 2.4.1.11 Reactor component No, not a high-N/A 2 N/A NPS 2 N/A cooling water energy system supply and return Notes: (1) A rounded value of 4.88 inches is used in the analysis as a conservatism.

(2) Small lines connecting to the RPV with a maximum OD of 0.5 in. exist in the containment for instrumentation and valve actuation purposes. Thermowells for temperature sensors located at RCS pressure boundary have a nozzle diameter of 2.63 in and an inner diameter of 0.875 in. These are bounded by the NPS 2 piping and therefore do not require dynamic analysis in order to generate design basis pipe break loadings.

2.4.1.1 Inadvertent Opening of a Reactor Safety Valve The RSVs are located on the reactor pressure vessel (RPV) head and discharge to the containment from the pressurizer steam space. As ASME code safety valves, they are designed for steam service. Therefore, design basis blowdowns from this location are saturated.

Asymmetric cavity pressurization analysis is required to quantify the asymmetric loading generated due to the safety valve effluent entering containment during this event.

2.4.1.2 Spurious Reactor Vent Valve Actuation The RVVs are located on the RPV head and are designed for steam service. In addition to the steam service function, they are also designed to discharge liquid for low temperature overpressure protection; however, due to the low temperature and pressure during this condition the dynamic responses are negligible. In lieu of using engineering drawings and form loss information for the valves, the break area is conservatively modeled equal to that of the nozzle. This neglects the losses and reduced flow area due to the valve internals and provides bounding blowdown loads.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Asymmetric cavity pressurization analysis is required to quantify the asymmetric loading generated due to the RVV effluent entering containment during this event.

2.4.1.3 Spurious Reactor Recirculation Valve Actuation The RRVs connect to the cold leg of the RCS and the limiting spurious actuations are expected to be subcooled. In lieu of using engineering drawings and form loss information for the valves, the break area is modeled equal to that of the nozzle. This neglects the losses and reduced flow area due to the valve internals and provides bounding blowdown loads.

Asymmetric cavity pressurization analysis is required to quantify the asymmetric loading generated based on RRV effluent entering containment during this event.

2.4.1.4 Chemical and Volume Control System Pipe Breaks Chemical and Volume Control System (CVCS) pipe breaks are assumed to occur on piping carrying fluid between the CVCS and the RCS at locations inside the CNV. While breaks in this piping could also occur outside containment, the mass flow rates for breaks inside containment bound those outside containment due to additional friction and form losses.

Per Table 2-2, there are five RCS piping connections to the RPV that could result in a design basis pipe break: the RPV high-point degasification line, the two pressurizer spray supply lines, the RCS discharge line, and the RCS injection line. These lines connect to the pressurizer steam space, the hot leg, and the cold leg and therefore represent saturated and subcooled blowdown conditions.

For analyzing blowdown loads, most of the design basis pipe break locations are bounded by the spurious RVV and RRV actuation. The RVV and RRV flow areas are larger and located in the same regions as the piping connections; therefore, a break in one of the piping connections is expected to produce a similar but smaller dynamic response. The one exception is the RCS injection line. This piping segment terminates in the hot leg riser.

The blowdown flow rate is smaller than a blowdown from the RCS discharge line, due to less subcooling in the hot leg and additional losses in the vessel internals. However, the structural response is modeled since the pressure wave travels internal to the riser, rather than external.

Asymmetric pressurization loads are similar to the loads generated during RVV and RRV actuation. However, since the pipe breaks are postulated along the length of the piping, the asymmetric pressurization loads for breaks not originating in the regions near the RRVs and RVVs are considered.

2.4.1.5 Steam Generator Tube Failure The steam generator (SG) is an ASME Section III Class 1 component, and a mechanical failure leading to a full shear of a SG tube is not considered credible for the NuScale design. In addition to the SG tube failure not being considered credible, the blowdown and

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 asymmetric pressurization loads for a SG tube failure are either bounded by loads generated due to other breaks or not applicable for this event.

Due to the low blowdown flow rate resulting from a SG tube failure, the loads on adjacent RVI are minimal compared to the loads generated due to the CVCS pipe breaks and inadvertent emergency core cooling system actuation events. Since the break does not result in coolant leaving the secondary pressure boundary, asymmetric cavity pressurization loads are not generated.

Therefore, due to both the designation of the SG as an ASME Section III Class 1 component and the minimal effect that the short-term hydraulic forces generated during the blowdown event have on the SG and surrounding components and structures, the loads do not need to be determined for this event. The long-term reactor module response to a SG tube failure event is analyzed since the pressure loads generated throughout this transient may be relevant to other components.

2.4.1.6 Steam Piping Failure The steam piping inside containment is designed for leak-before-break (LBB) requirements for pipes. Therefore, for mechanical analysis, only piping outside of containment provides postulated steam line break locations. The steam piping normally contains superheated steam.

Since there are no breaks inside containment, asymmetric pressurization is not applicable for this location.

2.4.1.7 Feedwater Piping Failure The feedwater piping inside containment is designed for LBB requirements for pipes.

Therefore, for mechanical analysis, only piping outside containment provides postulated feedwater line break locations. The feedwater piping normally contains subcooled liquid.

Since there are no breaks inside containment, asymmetric pressurization is not applicable for this location.

2.4.1.8 Pressurizer Heater Port and Main Steam and Feedwater Access Ports ASME Class 1 penetrations that do not connect to high-energy piping do not require postulating a RCPB break.

2.4.1.9 Control Rod Assembly Ejection A control rod ejection due to a rupture of the CRDM housing is not considered a design basis mechanical failure in the NuScale design.

A gross failure of a CRDM housing is not a credible event because CRDM housings are ASME Section III Class 1 components, subject to hydrostatic testing to 125 percent of system design pressure and are included in RPV hydrostatic testing. Housings are made of stainless steel with a high fracture toughness value. Stress levels due to thermal

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 transients are within the limits for ASME Section III Class 1 components, and pressure boundary welds meet the same design, procedure, examination, and inspection requirements as the welds on other ASME Section III Class 1 components.

2.4.1.10 Instrumentation & Control and Sensor Penetrations A breach in the pressure boundary of an instrumentation and control cable penetration or a sensor nozzle is not identified as a design basis event. Class 1 penetrations that do not connect to high-energy piping do not require postulating an RCPB break.

2.4.1.11 Decay Heat Removal System The decay heat removal system (DHRS) connects to the main steam piping outside of the CNV and upstream of the main steam isolation valves. The DHRS piping enters the reactor pool, and connects to the DHRS condensers. A condensate return line is provided from the bottom of each DHRS condenser, through the CNV to a tee in the feedwater line inside containment.

DHRS piping outside containment is designed for break exclusion; however, breaks are postulated at the DHRS condensate piping terminal ends inside containment. The effect of asymmetric cavity pressurization loads are not analyzed for the postulated DHRS terminal end break locations since they are a lower energy and have the same or smaller break flow area than other nearby postulated breaches such as the RRVs and the RCS discharge piping, as discussed in Section 6.1.3.

2.4.1.12 Containment Small Lines and Thermowells A breach in the pressure boundary of a thermowell is not identified as a design basis event.

Class 1 penetrations that do not connect to high-energy lines do not require postulating as an RCPB break.

Small lines are used for valve actuation and sensors. The maximum diameter of these lines is smaller than the nozzle diameter. The nozzles connect to reducers, which are similar in diameter to the RCS piping. Therefore, the blowdown and asymmetric cavity loads associated with breaks at these locations are considered bounded.

2.5 Recommended Mechanical Design Transient Analysis Methodology 2.5.1 Blowdown Analysis Based on the modeling methodologies identified in Section 2.3, it is not necessary to use or develop a proprietary code to generate forcing functions for a dynamic mechanical analysis.

Consistent with Sections 2.3.3 and 2.3.4, a modeling approach in which BCs are determined using a thermal hydraulic code and the FSI and dynamic mechanical response is analyzed in a multi-physics code is recommended. To provide consistency with other NuScale thermal hydraulic and mechanical applications, the codes NRELAP5 (Reference 7.2.19) and ANSYS (Reference 7.2.27) are used.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 For the NuScale design, the blowdown loads due to breaches in high-energy systems are not as significant as other dynamic loads (such as seismic). This allows for application of conservative analysis assumptions to ensure that the dynamic loads generated are bounding.

2.5.2 Asymmetric Cavity Pressurization Analysis As discussed in Section 2.2.1, the aspect of asymmetric cavity pressurization related to inducing loads on components, piping, and supports in the containment is relevant to the NuScale design. The phenomenon of asymmetric cavity pressurization is similar to the phenomenon internal to the vessel during the blowdown. The most significant difference is the thermal hydraulic properties through which the wave is transmitting. Asymmetric cavity pressurization inside containment results in slower wave transmission due to the lower speed of sound, and less FSI due to the lower density of the fluid.

2.6 Recommended Test Data for Benchmarking of Analysis Methodology In accordance with Reference 7.2.1, benchmarking is required for the models used to develop the loadings for blowdown and asymmetric cavity pressurization analyses.

Reference 7.2.1 requires the use of the HDR test data for benchmarking, and provides a list of other relevant test data that can be used for methodology benchmarking (see Section 2.2.1).

A discussion of the test data recommended to benchmark the NuScale methodology is described below.

2.6.1 Heissdampf Reactor Experiments Per Reference 7.2.5, the purpose of the HDR test program was to provide experimental data for use in verification of physical models, numerical methods, and computer codes for the analysis of thermal hydraulic and structural coupling during the subcooled and saturated phases of a blowdown event. The HDR experiments consist of a series of break sizes and different degrees of subcooling in the downcomer, as described in Tables 3 and 9 of Reference 7.2.5. The reported experimental data consists of the thermal hydraulic and structural time history results; therefore, this experiment provides a means to benchmark both the ability for NRELAP5 to accurately simulate the short-term thermal hydraulic phenomena, as well as the ability for ANSYS to accurately simulate the structural response using either experimental data or NRELAP5 simulated results as BCs. If differences in the experimental and simulated results exist, this allows for determining which part of the analysis is introducing the error.

Table 2-3 provides a comparison of key parameters in the HDR tests compared to the NuScale design. The HDR arrangement and test conditions are similar to the NuScale design, so this required benchmarking experiment is appropriate for the NuScale design.

Figure 2-2 provides a visual comparison of the similarities between the NuScale and HRD test facility. The figure also shows the major components and postulated break locations for the NuScale design.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 The HDR experiments were also performed to provide test data for asymmetric cavity pressurization events. The US-APWR used HDR V21.1 for benchmarking their subcompartment analysis code (Reference 7.2.7). Since the NuScale containment does not contain subcompartments, the benchmarking analysis is limited to the HDR results for the annular region surrounding the reactor vessel. Detailed modeling of the other HDR containment subcompartments is not necessary for methodology benchmarking.

Table 2-3 Comparison of Heissdampf reactor and NuScale blowdown properties HDR Parameter NuScale Reference Reference 7.2.3 Normal operating pressure (P0) 11 MPa 1595 psia 1850 psia -

Saturation pressure (Psat) - - 990 psia Assuming average temperature of 543.3°F P0-Psat 5.5 MPa 797.7 psia 860 psia -

Core outlet temperature (Tcore) - - 590.6°F Minimum design flow hot leg temperature at 102% power Downcomer temperature - - 486.8°F Minimum design (Tdowncomer) flow cold leg temperature at 102% power Tcore-Tdowncomer 0 - 50 °C 0-90°F 103.8°F -

Break diameter 7.874 in 4.875 in (1)

Section 2.4.1.2 0.2 m (RVV) and Table 2-2 Break opening time -- 1 - 5(2) ms Standard Review 1 - 2 ms Plan Section 3.6.2 Core barrel length 24.93 ft 26 ft -

7.6 m (upper riser)

Core barrel thickness 0.906 in 0.5 - 1.0 in -

23 mm (upper riser)

Core barrel diameter 8.727 ft 4.6 - 6.2 ft -

2.66 m (upper riser)

Maximum stress 100 MPa 14.504 ksi --

Maximum displacement 2 mm 0.079 in --

(1) A rounded value of 4.88 inches is used in the analysis as a conservatism.

(2) One millisecond pipe break time or five millisecond valve opening time.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 2-1 Comparison of NuScale Power Module and Heissdampf reactor 2.6.2 Marviken Experiments The Marviken critical flow tests are similar to the HDR experiments, except the pressure vessel contains a steam space and some cases involve blowdown from the steam space.

This test configuration is appropriate for benchmarking against the NuScale design because some breaches in the NuScale RCPB originate from the pressurizer steam space, such as actuation of the RVVs or some RCS line breaks. Use of this experiment for benchmarking demonstrates the ability of the NRELAP5 portion of the NuScale methodology to accurately simulate the short-term thermal hydraulic phenomena for breaks with little subcooling or at saturated conditions.

2.6.3 Bettis Hydraulic Pressure Pulse Experiment The Bettis hydraulic pressure pulse experiment is a benchmarking case documented in Reference 7.2.6, and recommended for use in Reference 7.2.1. This experiment consists of a pressure pulse test conducted with two different test sections: one solid and one flexible. A drop hammer and piston pulse were used to generate pressure pulses of up to 1150 psid over durations lasting between 6 to 47 ms. The fluid used in the experiments was room temperature water. The test series was performed with each test section at identical initial conditions to provide direct comparison of the increased FSI as a function of the test section rigidity.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 This test configuration is appropriate for benchmarking against the NuScale design because it provides a good representation of the damping that can occur when a flexible member interacts with fluid. Namely, that the structure absorbs the pressure wave energy and provides for reduced peak pressures. Use of this experiment for benchmarking demonstrates the ability of the ANSYS portion of the NuScale methodology to accurately simulate both low and high coupling between the fluid and the structure. A schematic of the test configuration is provided in Figure 2-2.

Figure 2-2 Test schematic for Bettis hydraulic pressure pulse (Fig. 4 of Reference 7.2.6)

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3.0 Validation Methods of the Short-Term Analysis Methodology The objective of the benchmarking analysis is to use NRELAP5 and ANSYS to simulate the thermal hydraulic conditions and resulting mechanical loads for various experimental configurations, and compare the simulation results to published literature testing data. The simulation results are compared to the experimental results in order to assess the error associated with the modeling methodology, and to identify important modeling parameters that must be specified to obtain appropriate results. Detailed objectives related to this benchmarking methodology that are addressed in this discussion are

define and justify a matrix of literature blowdown tests to be used for methodology benchmarking.

document the NRELAP5 and ANSYS models built to simulate the literature test configurations.

compare simulated BCs and loads to experimental results and hand calculations, where applicable.

provide modeling guidelines for calculating the design basis loads from high-energy breaches. Identify which parameters have a significant effect on the results and how they should be considered.

A discussion of the validation process and applicable validation cases for NRELAP5 and ANSYS are provided in the following sections.

3.1 NuScale Design Basis, Important Loads, and Benchmarking Test Matrix The benchmarking cases used for this analysis are integral and separate effects tests that provide experimental results of both the thermal hydraulic phenomena and the resulting mechanical loads. Research has been performed in the subject area of HELBs, and there are numerous experiments and standard problems that could be used for methodology benchmarking. Section 2.4.1 provides a summary of the break sizes, location, and phenomena that are of importance for the NuScale mechanical design basis events involving a breach in a high-energy system.

The specific mechanical loads of interest and associated break locations are summarized in Table 2-1 and Table 2-2. The parameters of highest importance for methodology benchmarking are the simulated forces and displacements determined using ANSYS. The focus of this benchmarking is the mechanical loads resulting from blowdown and asymmetric cavity pressurization, which includes thrust forces at the break location.

Section 2.6.3 recommends three test configurations for methodology benchmarking, the Bettis hydraulic pressure pulse, HDR blowdown, and Marviken jet impingement tests (JITs). The basis and justification for selecting these three tests for benchmarking is provided in Section 2.6.3.

Table 3-1 provides a discussion regarding how the selected experimental problems address the break locations and loads identified in Section 2.4.1 and comply with the guidance for dynamic analysis modeling provided in NUREG-0609 (Reference 7.2.1).

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Section 3.2 provides a discussion of the HDR tests, a description of the NRELAP5 and ANSYS models used to simulate the experiments, and an overview of the sensitivity studies performed.

Section 3.3 provides a discussion of the Bettis hydraulic pressure pulse tests, and a description of the ANSYS models used to simulate the experiments.

Section 3.4 provides a discussion of the Marviken JIT, a description of the NRELAP5 models used to simulate the experiments, and an overview of the sensitivity studies performed.

Simulation analysis and results are discussed in Sections 4.1 and 5.0.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-1 Phenomena and parameters for benchmarking Significant Mechanical Benchmarking Benchmarking Justification Phenomena Experiments Parameters The HDR experiments provide the means to compare the BCs and resulting RVI and RPV displacements and loads, HDR for two break geometries and different initial thermal hydraulic conditions.

Mass flow rate The Bettis hydraulic pressure pulse experiments are Thrust force appropriate because they provide a means to compare Fluid acceleration displacements and forces resulting from a pressure pulse, Blowdown Static pressure Bettis with both high and low degrees of FSI.

Differential pressure Strain Marviken JIT provides stagnation initial conditions that are Displacement similar to the range of conditions that will be experienced during a NuScale primary side break originating from the pressurizer steam region, or a secondary side break Marviken JIT originating from the steam piping outside containment. The results of this experiment are applicable only to the NRELAP5 simulations.

As discussed in Section 2.2.5, analysis is required to show HDR that the NPM can withstand loads generated when a Mass flow rate pressure wave develops in the annular space between the RPV and CNV. The primary difference between asymmetric Thrust force cavity pressurization (outside the RPV) and blowdown Asymmetric Fluid acceleration (inside the RPV) in the NuScale design is the fluid cavity Static Pressure Bettis properties. Since the density of the fluid in the containment pressurization Differential pressure is less than in the primary coolant, and the propagation of Strain the pressure wave in containment is slower, the degree of interaction between the pressure wave and the CNV is less.

Displacement Therefore, additional benchmarking cases to specifically Marviken JIT assess this phenomenon for the CNV region are not required.

3.2 Heissdampf Reactor Experiments The HDR test results are the preeminent reference related to blowdown loading analysis available, and are required to be used as a part of methodology benchmarking per NUREG-0609 (Reference 7.2.1). Table 2-3 provides a comparison of key parameters in the HDR tests compared to the NuScale design. The HDR arrangement and test conditions are similar to the NuScale design, so this required benchmarking case is appropriate.

Reference 7.2.5 provides the experimental results of the first three blowdown tests performed with RVI at the HDR test facility. Per Reference 7.2.5, the purpose of the HDR test program was to provide experimental data for use in verification of physical models, numerical methods, and computer codes for the analysis of thermal hydraulic and structural coupling during the subcooled and saturated phases of a blowdown event. The HDR experiments consist of a series of break nozzle sizes and different degrees of subcooling in the downcomer, as described by Tables 3 and 9 of Reference 7.2.5.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 HDR experiments were also performed to provide test data for asymmetric cavity pressurization events. Since the NuScale containment is a single compartment, the phenomena of asymmetric cavity pressurization is limited to the annular region surrounding the RPV, and the annular region inside the reactor vessel is considered appropriate for benchmarking per Table 3-1.

The following tests are recommended for benchmarking: V29.2, V31.1, and V32. HDR testing case V34 is not included in this benchmark analysis since the key feature of this test was to simulate loose core barrel supports (snubbers), which is not applicable to the NuScale design. A comparison of the HDR cases is provided in Table 3-2. With the exception of V34, each degree of sub-cooling is represented in the tests recommended for benchmarking.

Table 3-2 Comparison of Heissdampf reactor test conditions (Table 3 and 9 of Reference 7.2.5)

Upper Core Downcomer Downcomer Length of Test Number Pressure (bar) Temperature Temperature Sub-cooling Break Nozzle

(°C) (°C) (°C) (m)

V29.2 90 293 273 30 4.524 V31 V31.1 Note(1) 268 50 308 V31.2 Note(2) 110 1.369 V32 240 78 V34 300 300 18 Notes: (1) Performed to demonstrate repeatability, lack of hysteresis effects, and general quality of measurements.

(2) Performed with additional instrumentation.

3.2.1 Heissdampf Reactor Experiments NRELAP5 Models The NRELAP5 models for the HDR are comprised of fluid volumes and junctions. These components are used to represent the reactor vessel, discharge nozzle, break location, and containment.

An important factor in modeling the break location is ensuring that the model accurately predicts the subcooled blowdown and onset of choking at the break location. The NRELAP5 theory manual (Reference 7.2.19) provides a detailed discussion of the models and correlations available in NRELAP5. The general modeling approach provided below is consistent with the example subcooled critical flow model discussed in Section 2.10.1 of Reference 7.2.19.

The following sections detail the modeling geometries used to simulate the HDR experiments. Section 3.2.2 provides an overview of the variables investigated via sensitivity studies, and Section 3.2.2.5 discusses how the sensitivity studies are implemented in the model. A schematic of an example NRELAP5 model used to simulate the HDR experiments is provided in Figure 3-1. Different discharge nozzle nodalization and also component numbering conventions are used for the HDR V29.2, V31.1 and V32 simulations, and the HDR V31.1 sensitivity study.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-1 NRELAP5 models: example nodalization schematic for Heissdampf reactor experiment 3.2.1.1 Reactor Vessel The HDR reactor vessel is shown in Figure 3-2. The reactor vessel can either be modeled as a time dependent volume or a pipe. Per Reference 7.2.19, time dependent volumes are typically used to model mass sources and sinks, and pressure BCs. Alternatively, the reactor vessel can be modeled as a pipe to simulate the BC using a finite mass and energy.

Due to the short nature of these blowdown events, either component is appropriate to model the fluid in the RPV, as the fluid property changes in the RPV are small compared to the changes at the break location. The pipe component is selected for the benchmarking analyses. This RPV BC is not a significant contributor to the results and a sensitivity study is not performed.

The HDR reactor vessel geometry is summarized in Table 3-3. Control option three is used to specify the reactor vessel fluid pressure and temperature, in accordance with the values from Table 3-2 for each transient case.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-2 Schematic of the Heissdampf reactor pressure vessel and internals (Fig. 4-1 of Reference 7.2.16)

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-3 Heissdampf reactor geometric input parameters Parameter Value Reference RPV outer diameter 10.45 ft (3185 mm)(1) Figure 4-1 of Reference 7.2.16 Page 49, Enclosure 2-5 of Upper RPV thickness 4.41 in (3184-2960)/2= 112 mm)

Reference 7.2.26 Page 49, Enclosure 2-5 of Lower RPV thickness 5.59 in (135mm+7mm= 142 mm)

Reference 7.2.26 Page 74, Enclosure 2-15 of Core barrel outer diameter 8.73 ft (2660 mm)

Reference 7.2.26 Page 74, Enclosure 2-15 of Core barrel thickness 0.91 in (23 mm)

Reference 7.2.26 Mass ring mass 14,881.2 lb (6750 kg) Page 4-9 of Reference 7.2.16 Page 49, Enclosure 2-5 of RPV height 35.47 ft (10810 mm)

Reference 7.2.26 Page 74, Enclosure 2-15 of Core barrel length 24.84 ft (7571 mm)

Reference 7.2.26 Break nozzle (A1) inside Page 48, Enclosure 2-4 of 7.87 in (200 mm) diameter Reference 7.2.26 Break nozzle length for V31.1 4.49 ft (1369 mm) Table 3 of Reference 7.2.5 and V32 Break nozzle length for V29.2 14.84 ft (4524 mm) Table 3 of Reference 7.2.5 Notes: (1) The RPV outer diameter is specified as 3184 mm in certain references, such as pg. 49, enclosure 2-5 of Reference 7.2.26. The 1 mm difference does not have a detrimental effect on results.

3.2.1.2 Junction A single junction component is used to connect the reactor vessel to the discharge nozzle.

Choking is disabled at this location to avoid erroneous oscillations between choking in the nozzle and at the break plane.

3.2.1.3 Discharge Nozzle A pipe component is used to represent the piping segment leading up to the break location depicted in Figure 3-3. The cross sectional area is 0.03142 m2 (0.338 ft2), per the piping inner diameter specified. For case V29.2, the total length is 4.524 m (14.84 ft), and for the other cases a shorter pipe length of 1.3695 m (4.49 ft) is provided.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-3 Heissdampf reactor discharge nozzle dimensions and sensor locations (Fig. 6 of Reference 7.2.5)

Choking is disabled in all volumes and junctions within the nozzle, and smooth area changes are specified. Friction is modeled in the nozzle nodes assuming a smooth surface finish. The pressure and temperature initial conditions are specified in accordance with Table 3-2 for each transient case.

Sensitivity studies are performed to determine the optimal hydraulic nodalization as a function of the maximum time step. See Section 3.2.2 for additional discussion regarding the discharge nozzle nodalization.

3.2.1.4 Break Location The break location is an important parameter that must be specified to accurately simulate the blowdown event. The three components that are suited for modeling a break location in NRELAP5 are junctions, trip valves, and motor valves. The difference between a trip and motor valve is that an opening rate can be specified for a motor valve, whereas the opening rate of the trip valve is instantaneous. The difference between using a junction and valve is that the time in the transient when the break opens can be specified, whereas if a junction is used to characterize a break it must open at a restart point. Reference 7.2.5 provides an estimate of the actual break opening times that were measured during the HDR experiments. A motor valve is selected for use to provide flexibility in simulating the time and rate of the break opening.

Consistent with Reference 7.2.5, the e flag is specified and one at the break to activate the modified PV term in the energy equations. Full abrupt area change is specified at the break location. However, this selection does not affect the results since the break area is the same as the upstream pipe.

Two choking models in the NRELAP5 executable, Ransom-Trapp and Henry-Fauske are investigated in this evaluation. Sensitivity studies are performed to determine the optimal model and the most suitable model discharge coefficients during the blowdown events.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3.2.1.5 Containment Use of a pipe to model the changing conditions in the containment is not necessary since downstream conditions are not relevant after the onset of choking, which occurs rapidly.

The containment is specified as a time dependent volume at atmospheric pressure, with a static quality of one. The CNV is specified with the same length and area as the reactor vessel. This selection is acceptable since the geometric properties of the containment have minimal effect on the fluid properties, and the fluid properties are irrelevant once choking occurs provided that containment pressure remains below the critical pressure.

3.2.2 Heissdampf Reactor Experiment Sensitivity Study Parameters Sensitivity studies are performed to justify that an accurate match of the experimental data is found, and to quantify how a change in a modeling parameter affects the results. The parameters that are investigated in the sensitivity studies are the time step, spatial discretization, developmental model options, break opening time, and choking model type and coefficients. The following sections discuss the scope and implementation of the sensitivity studies. The sensitivity studies are performed using HDR V31.1, as this case represents the median amount of subcooling.

3.2.2.1 Time Step Control and Spatial Discretization The considerations for the time step and acoustic Courant limit are discussed in Section 2.1.2.2 of Reference 7.2.20. The acoustic Courant limit is a function of the minimum node length and the speed of sound in the fluid, per Eq. 3.1:

x t = Eq. 3.1 C water where, t = maximum time step (s) x = minimum node length in region of interest (ft)

Cwater = speed of sound in fluid (ft/s)

Table 3-4 provides a summary of the speed of sound, hydraulic nodalization, and the acoustic Courant limit for the V31.1 case.

A minimum time step of 10-11 seconds is specified in the input files. This value is not used in the analysis because the maximum time step remains lower than the material Courant limit; therefore, NRELAP does not perform a time step reduction during the transient. Note that a time step less than the material Courant limit is required to capture acoustic wave propagation in the nozzle upstream of the break location.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-4 Time step as a function of node length and speed of sound HDR Test Speed of Sound, Range of Node lengths Range of Time Steps Case cwater Investigated, x Investigated t=0.1x/c 0.0856m (16 nodes)

V31.1 1104.4 m/s 0.0428m (32 nodes) t=0.01x/c 0.0213m (64 nodes) (7.74 x10-7 s to 1.94x10-6 s) 3.2.2.2 Developmental Model Options Section 2.1 of Reference 7.2.21 recommends the use of Options 8 and 10 for pressure wave tracking; therefore, the effect of these options are investigated in the benchmarking to assess their effect on the NRELAP5 BC results. Option 8 decreases the time step when the there is a change in void fraction, and option 10 provides time step control based on changes in pressure within hydrodynamic volumes, such that the pressure cannot change by more than a factor of two during a time step. Note that Options 8 and 10 are not pre-verified for use in the NRELAP5 software. These model options are investigated here as a sensitivity study only. The final results of the benchmarking and guidelines for analyzing the NuScale break locations and valve actuations do not recommend use of these model options. Since detailed FSI and pressure wave tracking is handled with the ANSYS acoustic elements, these options are not important to the BC results; however, they are investigated for completeness.

3.2.2.3 Break Opening Time The use of prototypic break opening times versus the nearly instantaneous opening time of 0.001 seconds specified by NUREG-0609 is investigated to assess the effect on the NRELAP5 BCs. The sensitivity study is used to compare the experimental opening time to the NUREG-0609 recommended opening time, per Table 3-5. The break opening rate is determined by taking the reciprocal of the break opening time.

Table 3-5 Measured break opening time (Table 4 of Reference 7.2.5)

Break Opening Time Min Break Opening Simulation Case (ms) Rate (s-1)

HDR V31.1 Experiment 2.0 +/- 0.4 417 NUREG-0609 Recommended 1.0 1000 3.2.2.4 Choking Model Type and Coefficients The two choking models available in the NRELAP5 executable used in this simulation are Ransom-Trapp and Henry-Fauske. The standard choking model in NRELAP5 is Ransom-Trapp. In this model, the subcooled, two phase and superheated coefficients can be specified. The default value for each coefficient is 1.0; lower values can be used for break locations that are more similar to orifices.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 The default Henry-Fauske model discharge and thermal coefficients recommended in Section 2.3.2 of Reference 7.2.17 are 1.0 and 0.14, respectively. A discharge coefficient less than one provides for reduced flow from a break location. Reference 7.2.17 recommends using a value of nearly 1.0 for break nozzles, although a value slightly less than 1.0 may be appropriate. A value of 0.8 is used as the median discharge coefficient for the sensitivity study. This value is chosen based on the values recommended in Sections 7.2.7 and 7.2.8 of Reference 7.2.22.

For the models, a range of values above and below (0.75 and 0.85) are investigated for the discharge coefficient, in order to investigate how the coefficient can be used to provide agreement with the experimental results.

Thermal non-equilibrium is not investigated for this sensitivity case, since the case is an example of a highly subcooled blowdown.

3.2.2.5 Heissdampf Reactor NRELAP5 Sensitivity Study Methodology Table 3-6 provides a summary of the sensitivity parameters and ranges. The following section explains how the sensitivity studies are performed.

Table 3-6 Summary of sensitivity parameters for Heissdampf reactor V31.1 Parameter Range or Type Choking model type Henry-Fauske or Ransom-Trapp Choking discharge coefficients Low, Recommended, High Time step t=0.1x/c, t=0.01x/c Spatial discretization 16, 32, or 64 node Break opening time Instantaneous, experimental Developmental model options 8 & 10; None HDR V31.1 is chosen for a full sensitivity analysis because literature results for pressure and mass flow rate BCs are available for this case, and because it is the average subcooling case compared to V29.2 and V32.

The three parameters that are expected to have the largest impact on accurately representing the pressure and mass flow rate near the break location are the choking model and coefficients, the nodalization of the discharge pipe, and the time step used for the simulation. This conclusion is based on Section 7.2 of Reference 7.2.22.

To minimize the number of required runs, the effect of the acoustic Courant limit (via the spatial discretization and time step) is investigated first in cases A1 through A6. Using the optimal result from the Case A sensitivity set, the choking model and discharge coefficients are investigated in cases B1 through B6. Using the optimal result from the Case B sensitivity set, the break opening time and model option sensitivities are investigated in cases C1 through C4. The methodology assumes that there are no synergistic effects between variables not investigated at the same time.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Mass flow rate, pressure, and density are used to calculate the fluid acceleration and thrust force BCs (see Sections 3.2.3.1 and 0). For each sensitivity case, the optimal results are determined based on which set of parameters provides the best agreement with experimental data. When a parameter could not be found to provide close agreement with both the experimental mass flow rate and experimental pressure, the parameter that provides better simulation of the mass flow rate was chosen. Out of the three parameters (pressure, mass flow rate and density), mass flow rate is important in order to accurately predict the fluid acceleration and thrust force since pressure is not used to determine fluid acceleration and the mass flow rate is squared in the thrust force equation.

A discussion of the sensitivity study results and the recommended simulation parameters are provided in Section 4.1.

3.2.3 ANSYS Modeling Methodology for Heissdampf Reactor Test Simulations The dynamic loads associated with postulated piping breaks and valve discharges can be calculated by transient analysis using ANSYS. The HDR test cases V29.2, V31.1 and V32 are selected to perform the blowdown benchmark analysis. A half model with a symmetry BC is used, consistent with the HDR testing simulations in References 7.2.3 and 7.2.4.

The geometry, as shown in Figure 3-2, is overall symmetric except for several nozzles not in the symmetry plane (not modeled in ANSYS) and the bottom supports. The effect of these asymmetric details is not considered significant. Certain details of the HDR test vessel and blowdown nozzle geometry, such as the lower support skirt, are scaled from applicable references. The exact dimensions of geometric features for these regions of the test vessel do not have a significant effect on the simulation results; therefore, these simplifications are appropriate.

In the ANSYS models, the reactor vessel, core barrel and mass ring are represented by solid elements, and the fluid is represented by acoustic elements, which capture the coupling effect of the FSI at the fluid-structure interface. ANSYS provides various options for applying break BCs: (1) mass flow rate, (2) pressure, and (3) flow acceleration. As shown in Reference 7.2.3, due to the high flow velocity of water at the break location, the calculated pressure loads and structural responses caused by applying the acoustic pressure to acoustic elements at the break face, are clearly over-predicted after about 20 milliseconds. Similar results are observed in this analysis as documented in Appendix D.

In Reference 7.2.4, instead of an essential BC that applies acoustic pressure to the break face, the derivative of the mass flow rate is applied as a natural BC to the break face in ABAQUS, and good agreement with the measurements is obtained. In ANSYS, the equivalent BC is the flow acceleration for the acoustic elements. Therefore, the BC of flow acceleration is applied based on the above discussion.

At the break location, the flow acceleration is applied to the fluid area and the thrust force is applied to the nozzle cross section area. The mass flow rates are used to calculate the flow acceleration at the postulated break locations. The mass flow rate and pressure are used to calculate the thrust force. An acoustic impedance BC is applied to the fluid area.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3.2.3.1 Heissdampf Reactor ANSYS Models The HDR test cases V29.2, V31.1, and V32 are modeled using ANSYS. These testing cases have the same geometry except that the length of the nozzle for V29.2 is 4524 mm, but for V31.1 and V32 the nozzle length is only 1369 mm. The schematic of the HDR pressure vessel and internals is presented in Figure 2-2. The selected testing cases are simulated in ANSYS transient analysis using the process discussed in Section 3.2.3.

As shown in Figure 3-4, the test model is simulated by 3D structural elements in Figure 3-4(a) and acoustic elements in Figure 3-4(b), which capture the FSI effect at the interface.

The model for the V31.1 and V32 (1.369 m discharge nozzle length) configuration is shown. The bottom of the foundation is fixed. The flow acceleration is applied to the break face of the fluid as discussed in Section 3.2.3.1, and the thrust force is applied to the nozzle as a pressure at the break location, as discussed in Section 0. The transient analysis is run for 0.1 second with a time step of 0.001 second. A sensitivity study is performed with a time step of 0.0001 second to ensure that the 0.001 second time step is sufficiently small to capture the acoustic wave frequency.

(a) Structural Model (b) Acoustic Model Figure 3-4 ANSYS finite element analysis model of the Heissdampf reactor pressure vessel and internals

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3.2.3.2 Flow Acceleration at Break Locations Using the mass flow rates from experiment data or NRELAP5, fluid acceleration can be calculated using the forward difference approximation of the derivative of the fluid velocity, which uses the time-dependent fluid density.

1 m (t )

a1 (t ) = Eq. 3.2 Abreak (t )

where, 1(t) = fluid acceleration (m/s2)

(t) = mass flow rate at the break (kg/s)

Abreak = cross sectional area of break (m2)

(t) = density of the break effluent (kg/m3)

Alternatively, the flow accelerations can be calculated by taking derivative of the mass flow rate. To avoid noisy behavior of the mass flow rate derivative curves, a smoothing is first performed using a high order polynomial curve fitting with a 0 intercept. The flow acceleration is then calculated by dividing the derivative of the polynomial curve by the break area and the average fluid density, as described by the following equation.

f (t )

a2 (t ) =

Abreak Eq. 3.3 where, 2(t) = curve-fit fluid acceleration (m/s2) f(t) = derivative of best-fit mass flow rate curve (kg/s2)

Abreak = cross sectional area of break (m2)

= average density of the break effluent (kg/m3)

The two methods for calculating fluid acceleration are investigated using HDR V31.1 and HDR V32. The exact method, Eq. 3.2, is used for determining the fluid acceleration for HDR V32 (Appendix G). Eq. 3.3 is used for HDR V31.1, and provides good agreement with the experimentally determined fluid acceleration (see Appendix F).

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-5 shows an example pipe break location (which is a portion of the ANSYS model provided in Figure 3-4). Solid elements that represent the pipe wall are shown as light grey nodes. The acoustic elements representing the fluid inside the pipe are shown as dark grey nodes. The flow acceleration is applied as a body force to the acoustic element nodes on the break face.

Figure 3-5 ANSYS nozzle end nodes For example, the following ANSYS commands are used to read acceleration data from the text file acc.data and apply the acceleration time history to the acoustic element nodes on the break face BreakFace_Nodes.

dim,ACC,TABLE,2360,,,TIME

tread,ACC,acc,data !Acceleration (m/s^2)

CMSEL,S,BreakFace_Nodes BF,ALL,VELO,%ACC%,0,0 ALLS

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3.2.3.3 Nozzle Wall Thrust Force at Break Locations The thrust force generated by the fluid exiting the nozzle is equal to the thrust force exerted on the solid cross sectional area of the nozzle. The thrust force is calculated using Eq. 3.4, from page 86 of Reference 7.2.24. The mass flux is the mass flow rate per unit flow area, per Eq. 3.5. The total thrust of the fluid is equal in magnitude to the thrust on the nozzle, in accordance with Newtons third law and as described by Eq. 3.6.

In ANSYS, the thrust force is applied to the nozzle wall cross-section area as an equivalent pressure, per Eq. 3.7. Note that since Eq. 3.4 uses the area and pressure at the break plane, the calculated thrust is applicable only to the break plane.

T fluid G 2n

= Pn P +

A fluid gc Eq. 3.4 m

G= Eq. 3.5 A fluid G 2n T fluid = Tnozzle = Pn P + Afluid g c Eq. 3.6 Tnozzle G 2n Afluid

= Pn P +

Anozzle gc Anozzle Eq. 3.7 where, Tfluid/Afluid = equivalent pressure, thrust per unit flow area (psi)

Tnozzle/Anozzle = equivalent pressure, thrust per unit nozzle area (psi)

Anozzle = cross-sectional metal area of break location (in2)

Afluid = cross-sectional fluid flow area of break location (in2)

Tnozzle = thrust force on the metal area of break location (lbf)

Tfluid = thrust force on the fluid flow area of break location (lbf)

= mass flow rate at the break (lbm/s)

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Pn = pressure at the nozzle discharge (psi)

Poo = pressure in the discharge reservoir (psi)

G = mass flux (lbm/s-in2) n = discharge specific volume (in3/lbm)

The thrust force is applied to the nozzle wall cross-section area nodes as an equivalent pressure, as shown in Figure 3-5.

3.2.4 ANSYS Simulation Cases For each of the HDR testing cases, three sets of pressure and mass flow rate time history results are obtained from the NRELAP5 simulations for the best estimate case, the lower bound flow and the upper bound flow. For HDR tests V31.1 and V32, the mass flow rates and pressures are also obtained from digitized experiment data. The fluid acceleration and thrust force BCs are calculated using three NRELAP5 sets of simulated pressure, mass flow rate, and density for all experiments except for HDR V29.2. For that experiment, the best-estimate fluid acceleration is applied for all three ANSYS validation cases. The two methods for calculating fluid acceleration (Eq. 3.2 and Eq. 3.3) are used for the HDR V31.1 and V32 validation cases, as identified in the second column of Table 3-7.

The ANSYS transient cases are summarized in Table 3-7. The flow acceleration and thrust force are calculated and provided to ANSYS model as inputs. They are also plotted in Appendix C.

For HDR tests V31.1 and V32, the acoustic pressure BC is also examined to confirm that the essential mass flow rate BC overestimates the dynamic response as shown in Reference 7.2.3. Since good agreement is not obtained for this BC, only the BCs from experimental data are applied, as BCs derived from NRELAP5 simulations would only introduce more error. The ANSYS transient cases for the acoustic pressure BC are summarized in Table 3-8.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-7 ANSYS validation matrix - flow acceleration boundary condition ANSYS Mass Flow Rate for Pressure for Thrust Force Validation Case Acceleration BC 1 V29.2 NRELAP5 Best Estimate 2 V29.2 NRELAP5 Best V29.2 NRELAP5 Upper Bound 3 V29.2 NRELAP5 Lower Bound 4 V31.1 Experiment Data V31.1 Experiment Data 5 V31.1 NRELAP5 Best, Equation 3.3 V31.1 NRELAP5 Best 6 V31.1 NRELAP5 Upper Bound, Equation 3.3 V31.1 NRELAP5 Upper Bound 7 V31.1 NRELAP5 Lower Bound, Equation 3.3 V31.1 NRELAP5 Lower Bound 8 V32 Experiment Data V32 Experiment Data 9 V32 NRELAP5 Best, Equation 3.2 V32 NRELAP5 Best 10 V32 NRELAP5 Upper Bound, Equation 3.2 V32 NRELAP5 Upper Bound 11 V32 NRELAP5 Lower Bound, Equation 3.2 V32 NRELAP5 Lower Bound Table 3-8 ANSYS validation matrix - acoustic pressure boundary condition ANSYS Validation Acoustic Pressure BC Pressure for Thrust Force Case 1 V31.1 Experiment Data V31.1 Experiment Data 2 V32 Experiment Data V32 Experiment Data The dynamic responses produced by the ANSYS validation cases specified in Table 3-7 and Table 3-8 are compared to the experimental data or calculation data from available reports and published papers. The dynamic responses used to benchmark the ANSYS blowdown analysis methodology include absolute pressure and differential pressure in the fluid, and displacement and strain for the reactor vessel and core barrel.

Table 3-9 and Figure 3-6 summarize the dynamic responses that are used for benchmarking, the experimental sensor type and sensor location, and the figures in the appendices that compare the ANSYS transient analysis results with existing experimental and calculation data. Notes 2 - 5 of Table 3-9 document minor differences between the sensor locations reported in literature and the response location selected in the ANSYS simulation model. These minor deviations are less than 2 percent and are typically due to differences in the design versus as-built test facility dimensions.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-6 ANSYS finite element analysis model sensor locations

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-9 Dynamic responses for benchmarking Comparison Response Type HDR Location & Sensor (1) Reference Figure Core barrel, KS1008 1 Displacement V29.2 Fig. 26 of Reference 7.2.5 Figure E-1 (1330, 90°, 8410)

RPV, BS0106 2 Displacement V29.2 Fig. 28 of Reference 7.2.5 Figure E-2 (1590, 90°, 7350)

RPV, BS0107 3 Displacement V29.2 Fig. 28 of Reference 7.2.5 Figure E-3 (1590, 180°, 7350)

RPV, BS0108 4 Displacement V29.2 Fig. 28 of Reference 7.2.5 Figure E-4 (1590, 270°, 7350)

RPV, BS0106 5 Displacement V31.1 Fig. 28 of Reference 7.2.5 Figure F-1 (1590, 90°, 7350)

RPV, BS0107 6 Displacement V31.1 Fig. 28 of Reference 7.2.5 Figure F-2 (1590, 180°, 7350)

RPV, BS0108 7 Displacement V31.1 Fig. 28 of Reference 7.2.5 Figure F-3 (1590, 270°, 7350)

Fluid, BP9109 Fig. 4-5 of Reference 8 Pressure V31.1 Figure F-4 (1330, 90°, 8850) 7.2.16 Fluid, BP9117 Fig. 4-6 of Reference 9 Pressure V31.1 Figure F-5 (1330, 270°, 8850) 7.2.16 Fluid, BP9133 Fig. 4-7 of Reference 10 Pressure V31.1 Figure F-6 (1330, 88°, 5505) (2) 7.2.16 Fluid, BP9140 Fig. 4-8 of Reference 11 Pressure V31.1 Figure F-7 (1330, 90°, 2300) 7.2.16 Fluid, BP8301 Fig. 4-10 of Reference 12 Pressure V31.1 Figure F-8 (0, 0°, 10370) 7.2.16 Fluid, KP0009 Fig. 4-12 of Reference 13 Diff. pressure V31.1 Figure F-9 (1307, 90°, 8850) 7.2.16 Core barrel, KA2009 Fig. 4-15 of Reference 14 Hoop strain V31.1 Figure F-10 (1330, 90°, 8850) (3) 7.2.16 Core barrel, KA3008 Fig. 4-16 of Reference 15 Axial strain V31.1 Figure F-11 (1330, 90°, 8850) (3) 7.2.16 RPV, BS0106 Fig. 8 of 1982 Schumann 16 Displacement V32 Figure G-1 (1590, 90°, 7350) paper (Reference 7.2.23)

RPV, BS0116 Fig. A-47 of Reference 17 Displacement V32 Figure G-2 (1590, 90°, 5550) 7.2.16 Core barrel, KS1013 Fig. A-112 of Reference 18 Displacement V32 Figure G-3 (1307, 90°, 7195) 7.2.16 Core barrel, KS1030 Fig. A-118 of Reference 19 Displacement V32 Figure G-4 (1307, 90°, 2265) (4) 7.2.16 Core barrel, KS1032 Fig. A-120 of Reference 20 Displacement V32 Figure G-5 (1307, 270°, 2265) (5) 7.2.16

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Comparison Response Type HDR Location & Sensor (1) Reference Figure Core barrel, KA2008 Fig. A-66 of Reference 21 Hoop strain V32 Figure G-6 (1330, 90°, 8845) (3) 7.2.16 Core barrel, KA3009 Fig. A-71 of Reference 22 Axial strain V32 Figure G-7 (1330, 90°, 8825) (3) 7.2.16 Core barrel, KS1030 Fig. A-118 of Reference 23 (6) Displacement V32 Figure G-8 (1307, 90°, 2265) (4) 7.2.16 Core barrel, KS1032 Fig. A-120 of Reference 24 (6) Displacement V32 Figure G-9 (1307, 270°, 2265) (5) 7.2.16 Core barrel, KA2008 Fig. A-66 of Reference 25 (6) Hoop strain V32 Figure G-10 (1330, 90°, 8845) (3) 7.2.16 Core barrel, KA3009 Fig. A-71 of Reference 26 (6) Axial strain V32 Figure G-11 (1330, 90°, 8825) (3) 7.2.16 Notes: (1) Cylindrical coordinate system (R, , Z) is used with R and Z in mm.

(2) Used (1330, 90°, 5505) for approximation in ANSYS.

(3) Used (1330, 90°, 8847) for approximation in ANSYS.

(4) Used (1307, 90°, 2300) for approximation in ANSYS.

(5) Used (1307, 270°, 2300) for approximation in ANSYS.

(6) Responses are presented for sensitivity study of time step.

3.3 Bettis Hydraulic Pressure Pulse Experiment The Bettis hydraulic pressure pulse experiment was performed as part of the light water breeder reactor development program. The experiment is a separate effects benchmarking case documented in Reference 7.2.6. Reference 7.2.6 provides experimental results of flexible member tests that were performed to provide benchmarking of a computer code used to calculate pressure variations during a LOCA.

This experiment is recommended for use in HELB benchmarking applications in Reference 7.2.1.

A schematic of the test configuration is provided in Figure 2-2 and the test parameters defined in Table 3-10. The pressure pulse test apparatus consists of a pressure vessel with a piston on the top, and a test section that is mounted to the bottom of the vessel.

This experiment consists of a pressure pulse test conducted with two different test sections: one solid and one flexible. A drop hammer and piston pulse were used to generate pressure pulses of up to 1150 psid over durations lasting between 6 to 47 ms.

The fluid used in the experiments was room temperature water at 37.7 psia. The test series is performed with each test section at identical conditions to provide direct comparison of the increased FSI as a function of the test section rigidity. Cases 10 and 20 are selected for benchmarking because Reference 7.2.6 provides the most complete results for these particular cases. These cases provide a comparison of two different piston diameter sizes, with a flexible and solid wall.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 3-10 Bettis hydraulic pressure pulse test parameters and geometry (Table I of Reference 7.2.6)

Test Section Run Piston Piston Drop Hammer Drop Type Number Diameter Weight Weight Height 10F 1.0 inch 3.11 lb Flexible wall 20F 2.0 inch 10.4 lb 42.3 inch 6.0 inch 10S 1.0 inch 3.11 lb Rigid wall 20S 2.0 inch 10.4 lb 3.3.1 ANSYS Modeling Methodology for Bettis Hydraulic Pressure Pulse Test Simulations The high pressure pulse for the Bettis hydraulic pressure pulse test can be simulated with a transient analysis in ANSYS. The test configuration as shown in Figure 2-2 contains a cylindrical test vessel and a squared test section that can be represented by a 1/8th symmetric model. The vessel flange is not considered in the analysis models because the effect of this asymmetric detail is not significant for simulating the FSI, based on the best engineering practice. The piston, test vessel and test section are modeled using 3D structural elements and the water is modeled using 3D acoustic elements, as illustrated in Figure 3-7 for Run 10S/10F (with a 1-inch diameter piston).

A few parameters and BCs require calculation to define the model. These include the velocity of the piston just after impact, adjusted piston density (to account for the mass of the hammer), and the elastic modulus at the experimental pressure condition.

The initial velocity of the piston after the impact is calculated based on the momentum conservation principle with zero restitution. Since the hammer is not included in the ANSYS models, the density of the piston is adjusted to account for the mass of the hammer and the piston. The velocity of the piston after impact is applied as a BC for the transient mechanical analysis.

The velocity of the hammer, and adjusted density and velocity of the pistons are as follows:

2h t piston =

ac Eq. 3.8 vhammer = act piston Eq. 3.9 mhammer vhammer vi = Eq. 3.10 mhammer + mi

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 mhammer + mi i = Eq. 3.11 Vi where, tpiston = piston drop time to reach hammer (s) h = drop height between piston and hammer (in) ac = acceleration due to gravity (in/s2) vi = velocity of the piston just after impact (in/s) vhammer = velocity of the hammer (in/s) mi = mass of the piston (lb) mhammer = mass of the hammer (lb)

Vi = volume of the piston (in3) i = adjusted density of the piston (lb/in3)

The solid test section in runs 10S and 20S is simulated by the same geometry as the flexible test section in runs 10F and 20F, but with very high elastic modulus. An elastic modulus of 1E10 psi is 1000 times higher than the flexible test section and no deflection is expected. Therefore, it is equivalent to that of a solid test section.

A piston damping coefficient that accounts for the friction between the piston and the sleeve is applied to the models based on the damping coefficients. The value is divided by eight since a 1/8th symmetric model is used. For Run 10S/10F, a total time of 30 ms is simulated, while for Run 20S/20F, a total time of 10ms is simulated. A time step of 10-6 seconds is used as suggested on Page 15 of Reference 7.2.6. A stiffness damping constant of 5E-6 is used for the cases which provides a critical system damping ratio of about 1.5 percent for 1000 Hz. This is deemed reasonable since the test results are recorded with a frequency response of about 900 Hz, as discussed on Page 3 of Reference 7.2.6.

The total absolute pressure is obtained by adding the acoustic pressure simulated by ANSYS at the top and bottom transducer locations and the static pressure which is 37.7 psia (Page 13 of Reference 7.2.6).

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 (a) Structural Model (b) Acoustic Model Figure 3-7 ANSYS finite element analysis model of Bettis hydraulic pressure pulse tests Solid model components are shown on the left of Figure 3-7 and fluid elements are shown on the right of Figure 3-7. The acoustic elements are joined to the solid elements that represent the vessel and test section using a conformal mesh. Conformal mesh is recommended for fluid-structure interaction analysis to avoid convergence issues. Due to

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 the velocity boundary condition applied at the piston, the piston and acoustic body are joined using a bonded contact at the piston bottom surface.

3.4 Marviken Jet Impingement Test Experiment The Marviken JITs were performed to assess the ability for NRELAP5 to match the required BCs when the initial fluid conditions upstream of the break location are saturated.

This condition is applicable since the NuScale design has various high-energy lines and valves that contain steam. A summary of the testing initial conditions for JIT 1-12 are provided in Table 3-11.

Table 3-11 Comparison of Marviken test conditions (Tables 2-2 and 2-4 of Reference 7.2.25)

Nozzle Discharge Test Pressure Sub- Water Diameter Entrance Number (MPa) cooling (°C) Level (m)

(mm) Level (m) 1 509 4.96 32 16.7 0.74 2 299 5.24 33 9.1 0.74 3 509 4.97 52 18.6 0.74 4 200 5.24 33 7.5 0.74 5 299 5.12 Less than 3 8.8 18.33 6 509 5.04 32 18.2 4.0 7 509 5.01 35 16.0 0.74 8 509 5.00 34 16.4 0.74 9 200 5.20 32 7.5 0.74 10 509 5.00 34 16.4 0.74 11 299 5.00 Less than 3 10.2 18.33 12 509 5.00 34 15.4 0.74 A schematic of the experimental setup is shown in Figure 3-8. A detailed discussion of the Marviken JITs and a portion of the testing results are provided in Reference 7.2.17. A detailed RELAP5-3D developmental assessment of Marviken JIT-11 is documented in Section 4.5 of Reference 7.2.18.

The cross-sectional area at the break location in the JITs is significantly larger than the piping and valves inside containment in the NuScale design, and the breaks modeled in the HDR test series. Therefore, this experiment bounds the thrust forces and fluid acceleration that could be experienced for breaks originating from saturated locations in the NuScale design.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-8 Marviken jet impingement test schematic of test configuration (Figures 2-3 and 2-4 of Reference 7.2.25) 3.4.1 NRELAP5 Models -Marviken Jet Impingement The NRELAP5 model for the Marviken JIT-11 experiment is comprised of fluid volumes and junctions representing and connecting the pressure vessel, discharge nozzle, break location, and containment, similar to the HDR simulation models. The following sections detail the modeling geometries used to simulate the Marviken JIT-11 experiment. An example NRELAP5 model schematic of the Marviken JIT experiment is provided in Figure 3-9.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 3-9 NRELAP5 models: example schematics for Marviken experiment 3.4.1.1 Pressure Vessel and Standpipe The pressure vessel BC is modeled as a pipe component, which allows NRELAP5 to calculate the change in fluid conditions in the reactor vessel over time. Since the JIT vessel contains saturated steam and liquid, the pipe is divided into nodes to specify a water level.

Modeling of the pressure vessel in this simulation is different than in Section 4.5 of Reference 7.2.18. Specifically, the JIT model in Reference 7.2.18 simulates the pressure vessel using a time dependent volume, and applies the experimentally-measured pressure time history to the volume. This is not an appropriate method for the purpose of this evaluation since it is not possible to implement this simulation methodology for the NuScale plant break location (i.e., experimental data for the pressure time history in the reactor vessel for any NuScale break is not available).

3.4.1.2 Discharge Pipe and Nozzle The discharge pipe and nozzle characteristics are consistent with the modeling approach in Section 4.5.3 of Reference 7.2.18.

3.4.1.3 Break Plane The break location is the most important parameter that must be specified to accurately simulate the blowdown event. A motor valve is selected for use to provide flexibility in simulating the time and rate of the break opening, consistent with the HDR simulations.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Consistent with Reference 7.2.16, the e flag is specified and one at the break to activate the modified PV term in the energy equations. Full abrupt area change is specified at the break location. However, this selection does not affect the results since the break area is the same as the upstream pipe.

3.4.1.4 Containment Use of a pipe to model the changing conditions in the containment is not necessary since downstream conditions are not relevant after the onset of choking, which occurs rapidly in this event. The containment is specified as a time dependent volume at atmospheric pressure, with a static quality of one. The containment vessel is specified with the same length and area as the reactor vessel. This selection is arbitrary since the geometric properties of the containment have minimal effect on the fluid properties, and the fluid properties are irrelevant once choking occurs provided that containment pressure remains below the critical pressure.

3.4.2 Jet Impingement Test Experiment Sensitivity Studies Based on agreement with experimental results using the optimal parameters identified via the HDR V31.1 sensitivity study, a sensitivity study for JIT 11 is not performed. Agreement with the experimental data is obtained with the modeling parameters used for the HDR simulations.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 4.0 Validation Analysis 4.1 Thermal Hydraulic Analyses The time history results of the test cases are provided in Appendix A. Table 4-1 provides a summary of the simulation parameters that are used to achieve the benchmarking results for the HDR tests. The parameters summarized in Table 4-1 are determined using the sensitivity study for HDR V31.1 (per Section 3.2.2), and the results are discussed below. Experimental mass flow rates and pressures are plotted for sensor locations identified in Figure 3-4. Select plots are provided in Appendix A and are discussed below.

Case A investigates time steps of 1 percent and 10 percent of the acoustic Courant limit and spatial discretization of the discharge nozzle of 16, 32, or 64 nodes. Figure A-1 and Figure A-2 provide the sensitivity results for a time step of 10 percent of the acoustic Courant limit for three nodalization options (Case A2), and Figure A-3 and Figure A-4 provide results for the 1 percent acoustic Courant limit time step case (Case A5). The figures show that the degree of nodalization in the nozzle is more important for accurately matching the experimental pressure than for matching the mass flow rate, and that there is not a significant difference between a time step based on a 1 percent or 10 percent acoustic Courant limit. A nodalization of 32 nodes in the discharge nozzle provides the best results in the sensitivity study for HDR V31.1. A time step of 10 percent of the acoustic Courant limit is selected since a smaller time step does not improve simulation accuracy.

This nodalization and time step are used as the basis for the set of Case B models.

Case B investigates two choking models and three discharge coefficients. As shown in Figure A-5 and Figure A-6, the Henry-Fauske choking model provides better agreement with the short-term pressure time history than the Ransom-Trap model. With manipulation of the choking model discharge coefficients, either model can provide good agreement with the mass flow rate experimental results, as shown in Figure A-7. For continuity, the Henry-Fauske model with a discharge coefficient of 0.8 provides good agreement with the mass flow rate results (Figure A-8), and is recommended. This choking model and discharge coefficient are used as the basis for the set of Case C models.

Case C investigates developmental options and break opening time. As shown in Figure A-9 and Figure A-10, these parameters do not have a significant effect on the simulation results. Therefore, the NUREG-0609 recommended opening time and no developmental options are recommended for the HDR simulation models.

For the HDR simulations, Appendix C provides thrust force and fluid acceleration BCs using the experimental and NRELAP5 simulation data. Figure C-3 is calculated using Eq.

3.3, and Figure C-5 is calculated using Eq. 3.2. Figure C-3 uses the mass flow rate curve-fit and provides superior agreement with the experimental fluid acceleration BC. This is discussed further in Sections 4.2.2.2 and 4.2.2.3.

The simulation parameters for the HDR sub-cooled tests and the Marviken saturated steam are provided in Table 4-1 and Table 4-2, respectively. Figure A-1 and Figure A-2 provide the NRELAP5 results for the HDR mass flow rate simulations. Figure A-3 through Figure A-5 provide the experimental and simulated density, mass flow rate, and thrust force for JIT-11 simulation. Good agreement is shown with the flow rate and density

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 results. Differences in the simulated verses experimental forces are observed; however, these differences are within approximately 10 percent.

Table 4-1 Summary of optimal modeling parameters for Heissdampf reactor benchmarking cases Simulation Acoustic Options, Data Set Case and Time Courant Choking Discharge valve Name Used Nodes Boundary Step (s) Limit Model Coefficients opening For Post-Condition Ct/x time Processing HDR V29.2 0.8 96 3.87e-6 Best pressure 100.0 HDR V31.1 and 1.75 V32 Best 100.0 pressure 0.7 None, 0.1 HF 1000/s Lower bound 32 3.905e-6 0.14 HDR V31.1 and 0.8 V32 Best 0.14 flow rate 0.85 Upper bound 0.14 As discussed in Section 3.4.2, the simulation parameters that showed good agreement for the HDR experiments also show good agreement for the JIT-11 simulation. The parameters used are summarized in Table 4-2.

From the NRELAP5 results, the mass flow rate is read from the valve that represents the break plane. Pressure and density are read from the last node in the discharge nozzle (i.e., directly upstream of the valve). For the NuScale HELB NRELAP5 simulations, the results are read from consistent locations to conform with the benchmarking results, unless additional sensitivity studies, benchmarking, or alternate calculations are used to justify an alternate method.

Table 4-2 Summary of optimal modeling parameters for Marviken jet impingement test 11 benchmarking cases Simulation Case Time Acoustic Options and Data Set Name Choking Discharge and Boundary Step Courant Solution Used For Post-Model Coefficients Condition (s) Limit Ct/x Scheme Processing 0.75 Lower bound 0.14 Marviken 0.8 1.18e-3 0.1 HF None, 1000/s Best JIT 11 0.14 0.85 Upper bound 0.14

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 4.2 Mechanical Dynamic Analyses 4.2.1 Bettis Hydraulic Pressure Pulse Tests The simulated pressures by ANSYS are compared with the measured pressures. Figure B-1 and Figure B-2 compare the simulated pressures at both top and bottom transducer locations for Run 10S and 10F, respectively, which show good agreement with the experimental data. Similarly, Figure B-3 and Figure B-4 compare the simulated pressures at both top and bottom transducer locations for Run 20S and 20F, respectively, which also show good agreement with the experimental data. The peak pressures are summarized below in Table 4-3. The data for experiment and FLASH-34 calculation results (Bettis reference analysis tool) are obtained from Table II of Reference 7.2.6. As shown in Table 4-3, the dynamic analysis in ANSYS is able to calculate the peak pressure reasonably accurately. The time-history pressure results in Appendix B also show good agreement.

Table 4-3 Summary of peak pressures (psia) for Bettis hydraulic pressure pulse test simulations Test Run 10S 10F 20S 20F Experiment data 1130 850 900 800 Top transducer 950 1130 880 FLASH-34 1190 (Figure 2-1)

ANSYS 1072 909 959 821 Experiment data 1150 840 975 790 Bottom transducer 935 1330 980 FLASH-34 1260 (Figure 2-1)

ANSYS 1064 898 1030 831 4.2.2 Heissdampf Reactor Results Overview For HDR tests V29.2, V31.1 and V32, the fluid acceleration BC along with the thrust force results are presented in Appendix C.

Appendix D shows the structural model results with the pressure BC applied. The pressure between the nozzle and the core barrel decrease faster than that observed from experiments. This is because the ANSYS acoustic model is not able to capture the effect of high flow velocity of water at the break location when the acoustic pressure BC (i.e., the essential BC) is applied. The faster pressure decrease shown in Figure D-1 creates a higher differential pressure on the core barrel as shown in Figure D-2, which results in an over-estimated structural response on the core barrel. Figure D-3 and Figure D-4 show that the radial displacements of the core barrel bottom at both the nozzle side and the opposite side are over-estimated, compared to the experimental data. Similar results have been presented in Figure 15 of Reference 7.2.3. Therefore, the acoustic pressure BC is not recommended for blowdown dynamic analysis.

Appendix E, Appendix F, and Appendix G present the comparison of dynamic responses for the two different flow acceleration BCs along with the thrust force, for HDR V29.2, V31.1 and V32, respectively. The depressurization propagation at 0.1 second is shown in

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 4-1 for HDR V32. The core barrel deformations are presented at selected time points in Figure 4-2, which agree with the core barrel deformed shapes in Fig. 7 of Reference 7.2.4. The dynamic response comparisons are discussed in the following sections.

Figure 4-1 Heissdampf reactor V32 depressurization propagation from the break location (at 100 ms)

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 4-2 Heissdampf reactor V32 core barrel deformations (displacement scale factor of 200) 4.2.2.1 Heissdampf Reactor V29.2 Appendix E provides the results for the HDR V29.2 simulations. For this set of simulations, the fluid acceleration calculated using the NRELAP5 result is determined using Eq. 3.2, which provides an un-smoothed fluid acceleration. These responses match the experimental data reasonably well; however, the agreement is not as good using the smoothed fluid acceleration BC as was found for the HDR V31.1 simulations. The displacements on the core barrel and RPV are selected for comparison for this test case.

The displacement of the core barrel relative to the RPV at the sensor location KS1008 is compared to the experimental data in Figure E-1, which shows reasonable agreement with the experiment data. The displacements of the RPV at the sensor location BS0106, BS0107 and BS0108 are compared with the experimental data in Figure E-2, Figure E-3, and Figure E-4, respectively. Reasonable agreement has been achieved, although the results show that some displacements are over-predicted. It is also observed that the dynamic responses are not very sensitive to the mass flow curves from the NRELAP5 model. The local discrepancy presented in these figures might be introduced from ANSYS modeling and simulation, digitizing of the data in Fig. 26 of Reference 7.2.5 (which is plotted for 1 second), and mass flow rate and pressure from the NRELAP5 prediction.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 4.2.2.2 Heissdampf Reactor V31.1 The dynamic responses for HDR V31.1 are presented in Appendix F. For this set of simulations, the fluid acceleration calculated using the NRELAP5 result is determined using Eq. 3.3, which provides a smoothed fluid acceleration. Overall, these responses match the experimental data well. The displacements of the RPV at the sensor location BS0106, BS0107 and BS0108 are compared with the experimental data in Figure F-1, Figure F-2, and Figure F-3, respectively. The peak displacements for BS0106 and BS0108 are slightly over-estimated. Considering that the RPV displacement is dependent on the RPV stiffness, the over-estimation might be due to the model simplification that a symmetric model is0 used and the nozzle openings except for the break nozzle are not included in the ANSYS models. Overestimation of the displacement is conservative from the perspective of determining component dynamic loads; therefore, these results are acceptable. The simulated pressures are compared to the experimental data at the sensor locations in Figure F-4, Figure F-5, Figure F-6, Figure F-7, and Figure F-8, for BP9109, BP9117, BP9133, BP9140 and BP8301, respectively, which suggest that the propagation of the depressurization is slightly over-predicted. For example, at the end of 0.1 second, the pressure calculated for the experimental mass flow (denoted by green triangles in these figures) is about 0.5 MPa lower than the experimental data. This discrepancy could result from the assumed fluid element properties; however, the discrepancies are considered small and acceptable. It is also shown that the lower bound mass flow from NRELAP5 produces pressure match to the experimental data.

Figure F-9 compares the differential pressure at the core barrel inside and outside surfaces close to the break nozzle, which demonstrates agreement with the experimental data although the peak differential pressure is slightly over-estimated.

The hoop and axial strains of the core barrel outer diameter close to the break nozzle (sensors KA2009 and KA3008) are presented in Figure F-10 and Figure F-11, respectively, which demonstrate agreement with the experimental data. It is also shown that the ANSYS dynamic analysis produces a similar level of accuracy or better compared to the WHAMSE code calculation reported in Reference 7.2.16.

4.2.2.3 Heissdampf Reactor V32 The dynamic responses for HDR V32 are presented in Appendix G. For this set of simulations, the fluid acceleration calculated using the NRELAP5 result is determined using Eq. 3.2, which provides an un-smoothed fluid acceleration. These responses match the experimental data reasonably well; however, the agreement is not as good using the smoothed fluid acceleration BC as was found for the HDR V31.1 simulations. The displacements of the RPV at the sensor location BS0106 and BS0116 are compared with the experimental data in Figure G-1 and Figure G-2, respectively. The local discrepancy might be due to the model simplification that a symmetric model is used, and the nozzle openings except for the break nozzle are not modeled in the ANSYS models. Additionally, the reduced agreement with experimental results could be attributed to using the un-smoothed fluid acceleration BC.

The core barrel radial displacements relative to RPV are compared to the experimental data at the sensor locations KS1013, KS1030 and KS1032 in Figure G-3, Figure G-4, and

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-5, respectively. Larger discrepancies compared to the experimental data exist for the V32 simulations as compared to the V31.1 simulations. This is attributed to the use of an un-smoothed fluid acceleration BC.

The hoop and axial strains of the core barrel outer diameter close to the break nozzle (sensors KA2008 and KA3009) are presented in Figure G-6 and Figure G-7, respectively, which demonstrate good agreement with the experimental data.

As discussed in Section 3.2.3.1, the ANSYS transient analysis is run for 0.1 second with a time step of 0.001 second. A sensitivity study has been performed using the HDR V32 model with a time step of 0.0001 second to confirm that the 0.001 second time step is sufficient. As shown in Figure G-8 through Figure G-11, the results using the time step of 0.0001 second are only slightly different from the results using the time step of 0.001 second, which confirms that 0.001 second is a proper time step for use in the short-term transient analysis.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 5.0 Validation Conclusions The results of the thermal hydraulic and mechanical dynamic analyses documented in Section 4.0 provide a high level of confidence that the dynamic loads associated with a HELB can be acceptably modeled for the NuScale design using the methodologies described in Section 4.1. Section 4.2.2 shows that the dynamic analysis results are not sensitive to the thermal hydraulic BCs, and that an accurate structural response can be generated with known modeling simplifications. It is concluded that a similar fidelity in the NuScale structural response can be obtained.

Appendix A through Appendix G demonstrate that the parameters important to dynamic analysis compare favorably with the experimental results. The fluid acceleration boundary condition of Eq. 3.3 applied to the acoustic elements and the thrust force applied to the solid elements at the break plane (as shown in Appendix F) provide the best agreement with the experimental results.

Based on overall favorable agreement, in addition to the bounding simplification of neglecting the valve flow diffusers, application of a biasing margin on the thermal hydraulic or dynamic analysis results for the NuScale design is not necessary.

Section 5.1 provides the recommended NRELAP5 and ANSYS modeling guidelines to perform the NuScale HELB analyses consistent with the methodology and results of the benchmarking analyses for HDR Blowdown, Marviken Jet Impingement and Bettis Hydraulic Pressure Pulse.

5.1 NuScale Power Module Modeling Guidelines 5.1.1 NRELAP5 Modeling Guidelines Hand calculations are performed to ensure the thermal hydraulic conditions calculated for the NuScale HELBs are reasonable. This is an important step in modeling the NuScale HELBs because some breaks include extrapolation outside of the range of pressures, temperatures, and cross sectional areas over which benchmarking was performed. The hand calculations are performed using Figure 2.20(a) of Reference 7.2.24 and a steam table. The maximum mass flow rate is estimated based on the initial enthalpy and pressure of the fluid at the break location.

For blowdowns that are initially subcooled liquid or superheated steam, the mass flow rate at the time the system reaches saturation is also determined. At this time, the system pressure is roughly equal to the initial saturation pressure. The enthalpy is estimated by assuming the system has come to equilibrium at a saturated steam or saturated liquid condition. The energy contribution due to flow is not considered when specifying the stagnation pressure and enthalpy, since the assumed pressure and temperature are relevant to the entire RCS, not only the break location.

Table 5-1 provides the modeling guidelines for simulating the thermal hydraulic BCs for the NuScale break locations. The modeling parameters are consistent with the discussion in Section 4.1. Engineering judgment is used in determining the acoustic Courant limit and the discharge coefficients. As described in Section 4.2.2, the variations in structural

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 response using the natural BC derived primarily from the curve-fit mass flow rate time history results are not significant.

For completeness, sensitivity studies are performed to confirm that the BCs determined in the NuScale analysis, using the recommended acoustic Courant limit and discharge coefficients, are acceptable. The sensitivity ranges are provided in Table 5-1.

Because pressure, mass flow rate, and density time histories are necessary for generating the ANSYS BCs, the time step and plot frequencies must be specified such that for each case the plot frequencies are the same. Providing a common plot frequency for each event simplifies the data handling associated with calculating the ANSYS BCs.

Table 5-1 Modeling parameters for the NuScale break locations Acoustic Containment Reactor Courant Discharge Options and Volume Choking Model Vessel Limit Coefficients Opening Rate Volume Ct/x 0.8 (2) None Time

0. 1 (1) Henry-Fauske pipe dependent 0.14 (3) 1000/s volume Notes: (1) Sensitivity studies one order of magnitude above and below are performed (1.0, 0.01).

(2) Sensitivity studies on the discharge coefficients of +/- 20% are performed.

(3) Sensitivity studies on the thermal non-equilibrium values are performed (0.05, 100.0).

5.1.2 ANSYS Modeling Guidelines The structural responses to the blowdown event are simulated using ANSYS FSI transient analysis with the recommended BC, which is the flow acceleration corresponding to the derivative of the mass flow rate along with the thrust force at the break location. The acoustic pressure BC is not used because it is not able to capture the effect of the high flow velocity of water at the break location. The flow acceleration at the break location is calculated from the NRELAP5 mass flow rate. A high-order order polynomial curve fitting is used to smooth the mass flow rate data to avoid any noisy behavior of the mass flow rate derivative curves, based on the results in Sections 4.2.2.2 and 4.2.2.3. A uniform temperature and initial pressure may be assumed for the acoustic body. The flow acceleration and thrust force are calculated based on the constant density of the acoustic body.

A conformal mesh between the structure and the fluid is used for the NuScale modeling to avoid potential convergence issues with dissimilar meshes and contact definitions. The overall geometry and loading conditions should be reviewed to determine whether a half model can be used. The ability to use a half model is dependent on the break location and whether the expected deformation has symmetry planes.

A time step of 0.001 second is used for the short-term blowdown transient analysis based on the results of sensitivity studies (Section 4.2.2.3). The speed of sound associated with the acoustic elements used in this benchmarking analysis is greater than the average speed of sound throughout the fluid for the NuScale design. The time step of 0.001

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 seconds provides adequate resolution for tracking the pressure wave in the acoustic elements for the NuScale simulations.

If BCs other than thrust force and fluid acceleration are desired for dynamic analysis, sensitivity studies in ANSYS are performed to demonstrate that the BCs used are appropriate.

5.1.3 Discussion of Extrapolation As identified in Table 5-2, the HDR and Marviken experiments provide similar break locations that are applicable to the NuScale design. However, not all parameters important for characterizing a break in the NuScale design are within the range of the parameters in the HDR and Marviken experiments. The parameters judged to be most important for accurately characterizing the break location are pressure, temperature, degree of subcooling, break area, and nozzle length.

The thermal hydraulic initial conditions are important because they determine the time at which choking at the break plane occurs. Once choking has occurred, the velocity of the flow is equal to the local speed of sound, and further decompression disturbances cannot propagate upstream. The break area is important because the larger the break area, the faster choking occurs. The nozzle length is important in that it contributes to the degree of homogeneous equilibrium in the discharge fluid. A larger subcooling increases the magnitude of the decompression wave and the loads on internals. The estimated subcooling temperatures in the NuScale design (during normal operating conditions) are within the ranges of the Marviken and HDR experiments.

The NuScale normal operating downcomer temperature is within the range of the Marviken and HDR experiments. The NuScale pressure and hot leg temperature exceed the conditions modeled in Marviken and HDR; however, the break areas in the NuScale design are smaller than HDR and Marviken. Therefore, despite the higher energy of the NuScale break locations, the critical mass flow rates are expected to be lower than the flow rates observed in the HDR and Marviken experiments.

Based on this discussion, it is concluded that significant extrapolation errors are not present based on the similarities between the benchmarking tests and the postulated NuScale breach locations. Further, any errors that may be present are bounded by the simplification of not modeling the ISRs and flow diffusers, which would significantly reduce the dynamic loads associated with blowdown and asymmetric cavity pressurization.

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 5-2 Heissdampf reactor, Marviken, and NuScale high-energy line break comparison Upper Break Test Saturation Downcome Downcomer Break Pressure Core Nozzle Number Temperature r Temp. Sub-cooling Area Temp. Length HDR 90 bar 303.3°C 293°C 273°C 30°C 4.524 m V29.2(1)

HDR 0.03142 m2 268°C 50°C V31.1(1) 110 bar 318.1°C 308°C 1.369 m HDR V32 240°C 78°C Marviken saturated 50 bar 263.9°C 262.4°C 262.4°C 1.18 m 0.07022 m2 JIT 11(2) steam NuScale operating

((

conditions(3)2(a),(c),ECI Notes: (1) Values for the HDR tests taken from Table 3-2. (2) Values for the Marviken tests taken from Table 3-11. (3) Nominal, best-estimate NuScale temperatures, dimensions and pressure are listed in this table to provide a comparison with the testing configurations. Each NuScale HELB calculation provides the NuScale HELB conditions and an evaluation of whether additional modeling margin is required due to extrapolation outside of the benchmarking range. © Copyright 2019 by NuScale Power, LLC 59

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 6.0 NuScale Power Module Asymmetric Cavity Pressurization and Blowdown This section summarizes how the dynamic loads associated with the blowdown inside the RPV and asymmetric pressurization of the cavity between the CNV and the RPV are generated. 6.1 NRELAP5 Boundary Conditions for Asymmetric Cavity Pressurization and Blowdown NRELAP5 analysis is performed to characterize the breaches in high energy lines and generate appropriate BCs for dynamic analysis. The break locations are analyzed in three groups:

Subcooled primary coolant breaks
Saturated primary coolant breaks
Secondary side pipe breaks

             - Steam and feedwater breaks postulated to occur outside containment (transient load applied directly in piping structural analyses) 6.1.1    NRELAP5 Subcooled Breaches 6.1.1.1 NRELAP5 Subcooled Modeling Considerations Subcooled breaches in the primary coolant consist of breaks in the chemical and volume control system (CVCS) injection and discharge lines, and inadvertent operation of the RRV.

A simplified model of the RCS for simulating the subcooled breaches is provided, as shown in Figure 6-1. Pipe components are used to allow NRELAP5 to determine the time-history changes in the RCS pressure and temperature due to the blowdown event. Five pipes are used to simulate the different sections of the RCS: the hot leg, cold leg, pressurizer, SG region and core, in order to implement the temperatures and pressure in the RCS loop. RCS volumes are prototypic; however, arbitrary pipe lengths are selected to comply with loop closure requirements. Junctions with no form losses and no choking models are used to connect the pipes. Containment is modeled using a time-dependent volume. The RCS provides a BC for determining the conditions at the break location. Per Section 5.1.1, it is not necessary to provide a detailed model of the RCS, since the changes in the RCS are small relative to the changes at the break location for the timescales of interest. Dynamic loads can be accurately calculated in ANSYS using BCs from a simplified NRELAP5 model. The CVCS reactor vessel internals are modeled to represent the connection of the injection line to the RCS hot leg. The RRV and discharge line connects directly to the RPV nozzle safe ends. To bound the possible breaks inside containment, breaks are simulated at the RPV and CNV terminal ends. © Copyright 2019 by NuScale Power, LLC 60

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 6-1 Schematic for the reactor coolant system subcooled blowdown model To simulate the transients, the modeling time step is set at 10 percent of the acoustic Courant limit based on the minimum node length. The time step is increased as the event progresses because the acoustic Courant limit is no longer a consideration once choking occurs; however, the time step is not increased above the material Courant limit. The valve opening time and pipe rupture times are set at 0.005 and 0.001 seconds, respectively. Sensitivity studies are performed for the acoustic Courant limit and variations in the choking model thermal non-equilibrium constant and discharge coefficient. The objective of the sensitivity studies is to qualitatively demonstrate the effect of the modeling parameters on the calculation results, in order to confirm that the parameters recommended via the benchmarking evaluations are appropriate for the NuScale design. Sensitivity studies are performed for the inadvertent RRV opening and the CNV terminal end break on the CVCS injection line piping. These two cases are selected for full sensitivity (i.e., acoustic Courant and choking model checks) because they provide the highest and lowest mass flow rates out of the cases modeled. Additionally, the CVCS discharge line pipe break at the RPV is evaluated with a larger thermal non-equilibrium constant. Hand calculations are performed to ensure that the thermal hydraulic conditions calculated for the NuScale HELBs are reasonable. This is an important step in modeling the NuScale HELBs because some NuScale breaks include extrapolation outside of the range of © Copyright 2019 by NuScale Power, LLC 61

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 pressures, temperatures, and cross-sectional areas over which HELB methodology benchmarking was performed. This hand calculation provides verification of the accuracy of the simulated results. The mass flow rate is estimated based on the enthalpy and pressure of the fluid at the break location using Figure 2.20(a) of Reference 7.2.24. The theoretical initial mass flow rate is calculated from the mass flux corresponding to the RCS pressure and enthalpy at the break and the break flow cross-sectional area. The mass flow rate at the time that the system reaches saturation is also determined. At that time, the system pressure is roughly equal to the initial saturation pressure. The enthalpy can be estimated by assuming the system has come to equilibrium with no change in the system temperature. Note that the energy contribution due to flow is not considered when specifying the stagnation pressure and enthalpy, since the assumed pressure and temperature are relevant to the entire RCS, not only the break location. The Reference 7.2.24 method for estimating the blowdown mass flow rate assumes an isentropic nozzle, and therefore over-predicts the mass flow rate simulated using NRELAP5 with the discharge coefficients recommended in Section 5.1.1. The expected mass flow rate is estimated applying the recommended discharge coefficient reduction. 6.1.1.2 NRELAP5 Subcooled Results The mass flow rates predicted by NRELAP5 are generally in good agreement with Figure 2.20(a) of Reference 7.2.24. For the discharge line break at the RPV and the inadvertent RRV opening, there is less than 1 percent variation in mass flow rate from the expected values. There is a greater under-prediction for the injection line break at the RPV but this can be attributed to the greater friction and form losses in the CVCS reactor vessel internals piping between the riser and the RPV wall. Friction and form losses in the lengths of piping between the RPV nozzle and CNV wall have an even larger flow-limiting effect for the injection and discharge line breaks at the CNV. Differences may also result in part from errors in estimating the theoretical mass flow rate using Figure 2.20(a) of Reference 7.2.24. In general, Table 6-1 demonstrates that the NRELAP5 results are in good agreement with hand calculations and are appropriate for use. The sensitivity studies demonstrate that the results are reasonable based on the acoustic Courant limit and specified choking model parameters. © Copyright 2019 by NuScale Power, LLC 62

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-1 Critical mass flow rate theoretical, expected, and NRELAP5 results Expected Peak Mass Flow Break Theoretical Mass Rate from Location/ Case Mass Flow Flow Rate NRELAP5 Valve Rate (lb/s) (lb/s) Modeling (lb/s) Operation Break at CVCS reactor coolant system injection RPV (( line Break at CNV Break at CVCS reactor coolant system RPV discharge line Break at CNV RRV RRV }}2(a),(c),ECI opens(1) Notes: (1) Inadvertent single RRV opening is a design basis event. The time history pressure, mass flow rate and density results for each break and each set of initial conditions are used to generate the flow acceleration and thrust force ANSYS BCs. Plots of the CVCS injection line pipe break and the RRV opening are provided in Figure 6-2 through Figure 6-5. Figure 6-2 shows fluctuations in the fluid acceleration, which are more pronounced for the CNV terminal end break. These fluctuations are due to small but abrupt changes in the mass flow rate at the break location. The changes in mass flow rate and the calculated fluid acceleration are likely because choked flow conditions have not yet been reached. The fluid acceleration results are repeatable in the sensitivity studies and are judged to be acceptable. © Copyright 2019 by NuScale Power, LLC 63

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                          }}2(a),(c),ECI Figure 6-2      Flow acceleration boundary condition - chemical and volume control system injection pipe break
  © Copyright 2019 by NuScale Power, LLC 64

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                             }}2(a),(c),ECI Figure 6-3      Thrust force boundary condition - chemical and volume control system injection pipe break
  © Copyright 2019 by NuScale Power, LLC 65

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                          }}2(a),(c),ECI Figure 6-4      Flow acceleration boundary condition - reactor recirculation valve inadvertent opening
  © Copyright 2019 by NuScale Power, LLC 66

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                              }}2(a),(c),ECI Figure 6-5      Thrust force boundary condition - reactor recirculation valve inadvertent opening 6.1.2    NRELAP5 Saturated Breaches Saturated breaches in the primary coolant consist of breaks in the pressurizer spray and high point vent lines, operation of the RSVs and inadvertent opening of a RVV. A simplified model of the RCS is provided, as shown in Figure 6-6.
  © Copyright 2019 by NuScale Power, LLC 67

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 6-6 Schematic for the reactor coolant system saturated blowdown model The model development, sensitivity studies, and analysis process is consistent with Section 6.1.1. The mass flow rates predicted by NRELAP5 are in good agreement with Figure 2.20(a) of Reference 7.2.24. There is an over-prediction of mass flow rate for the RVV and RSV operation cases. The higher mass flow rate is attributed to the static quality of the discharge fluid. The theoretical and expected mass flow rates were determined using Figure 2.20(a) of Reference 7.2.24 and assuming a static quality of 1.0; however, the static quality decreases shortly after blowdown as liquid swells in the pressurizer. This results in an increase in the mass flow rate for a given pressure, as shown in Figure 2.20(a). A larger pressurizer level swell results in a higher fluid density at the breach. In general, Table 6-2 demonstrates that the NRELAP5 results are in good agreement with hand calculations and are appropriate for use in analyses. The sensitivity studies demonstrate the results are reasonable based on the recommended time step and choking model parameters. Plots of the degasification and spray line pipe breaks and the RVV opening are provided in Figure 6-7 through Figure 6-10 (the RSV result is bounded by the RVV case and therefore, plots for the RSV are not shown). © Copyright 2019 by NuScale Power, LLC 68

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-2 Critical mass flow rate theoretical, expected and NRELAP5 results ((

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

((

                                                                                                               }}2(a),(c),ECI Figure 6-7      Flow acceleration boundary condition- spray and degasification line pipe breaks
  © Copyright 2019 by NuScale Power, LLC 69

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                            }}2(a),(c),ECI Figure 6-8      Thrust force boundary condition- spray and degasification line pipe breaks

((

                                                                                                                 }}2(a),(c),ECI Figure 6-9      Flow acceleration boundary condition- inadvertent reactor vent valve opening
  © Copyright 2019 by NuScale Power, LLC 70

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 ((

                                                                                                             }}2(a),(c),ECI Figure 6-10 Thrust force boundary condition- inadvertent reactor vent valve opening 6.1.3    NRELAP5 Secondary Side Breaches Unlike the primary side breaches, the steam generator system secondary side piping inside containment is qualified to LBB. Therefore, pipe breaks in the steam or feedwater piping do not generate asymmetric cavity pressurization loads. Postulated double ended breaks in the feedwater and steam piping are outside of the NPM, in the reactor building (RXB) gallery area.

Pipe breaks are postulated at the DHRS condensate piping terminal ends inside containment. However, the effect of asymmetric cavity pressurization loads are not analyzed for the postulated DHRS terminal end break locations since they are a lower energy and have smaller break flow area than other nearby postulated breaches such as the RRVs and the RCS discharge pipe. NRELAP5 is used to determine thrust forces and loading due to the pressure wave traveling through the piping based on a momentum balance of fluid. The transient loads calculated in NRELAP5 are applied directly to the affected piping systems in transient structural analyses. The structural analyses quantify the pipe stresses and provide the reaction forces at the nozzle and supports. The results of these analyses confirm the postulated pipe break locations defined in the NuScale Pipe Rupture Hazards Analysis.

  © Copyright 2019 by NuScale Power, LLC 71

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 6.2 ANSYS Analysis for Asymmetric Cavity Pressurization and Blowdown The purpose of the ANSYS analysis is to determine structural loads that result from HELBs in the primary coolant system. Structural loads associated with asymmetric cavity pressurization of the CNV acting simultaneous to the blowdown of the RPV are evaluated. Loads specified in this section contribute to the basis for the mechanical design of the NPM. Because of the negligible blowdown flow rate resulting from a single SG tube failure, the loads on adjacent RVI are minimal compared to the loads generated due to the design basis pipe breaks, inadvertent RSV opening and inadvertent emergency core cooling system valve opening events. As discussed in Section 2.4.1.5, loads due to SG tube rupture are not considered. The dynamic loads associated with CVCS injection line break, reactor vent valve (RVV), or reactor recirculation valve (RRV) opening are calculated using transient structural analyses in ANSYS. A single model including two acoustic bodies of CNV and RPV used to simulate the loads. Blowdown and asymmetric cavity pressurization loads are not separated because blowdown and asymmetric cavity pressurization are considered to act simultaneously in the dynamic finite element model of the NPM. In addition to thrust loads acting at the break location, HELBs result in pressure transients associated with propagation of the acoustic wave, acting upon surfaces of the fluid-structure interfaces of the NPM. These pressure transients create dynamic responses of the CNV, RPV, and internal structures, in the horizontal and vertical directions. For each break event, the in-structure acceleration and displacement time history responses are determined at key locations of the CNV, RPV, and internal structures. Also, at key structural cross sections and structural interfaces, the maximum forces and moments due to the most limiting break are provided. The propagation of the pressure wave results in differential pressure loads across internals structures or in transient hoop stresses within axisymmetric structures. Loads specified in this section apply to the design of the CNV, containment supports, RPV, and reactor internals. Core plate displacement time histories are provided for use in design of the fuel assemblies. Table 6-3 lists the valve openings and pipe break modeled in the ANSYS modeling cases. Due to the small line size, the pipe breaks are bounded by the loads from the valve openings. For example, a RCS discharge line break is bounded by an RRV valve opening. The RCS injection line break is included in the model. The RCS injection line terminates in the riser, resulting in a pressure wave internal to the riser. © Copyright 2019 by NuScale Power, LLC 72

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-3 Cases for valve opening and break locations Case No. Valve Opening / Break Location 1 RVV - RPV16 (or RVV16) 2 RVV - RPV17 (or RVV17) 3 RVV - RPV83 (or RVV83) 4 RRV - RPV1 (or RRV1) 5 RRV - RPV2 (or RRV2) 6 CVCS Injection Line Break 6.2.1 Geometry A full geometry model with two acoustic bodies is used for this analysis as shown in Figure 6-11. The minor features do not affect the gross structural behavior of the model and removing them allows for simplified meshing techniques to be used. The mass of existing structures in the model are adjusted to account for these model simplifications. The model mass is compared to the NPM mass calculation and the difference is applied to the model using a combination of point masses, distributed mass elements, and adjusted densities. Additionally, components that are not included in the model, such as the mass of the fuel, are accounted for in the mass of adjacent components. For fuel, the mass is accounted for in the mass specified for the reflector. © Copyright 2019 by NuScale Power, LLC 73

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 6-11 Full Model with Two Acoustic Bodies 6.2.2 Mesh The CNV, RPV, lower RVI, upper RVI, RVV, RRV, RSV, and CVCS Injection line are represented by solid elements. The fluid is represented by acoustic elements. A conformal mesh is created between the fluid and the solid elements to avoid potential convergence issues with dissimilar meshes and contact definitions. An example of this mesh is shown in Figure 6-12. Acoustic elements capture the coupling effect of the FSI at the fluid-structure interface. © Copyright 2019 by NuScale Power, LLC 74

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure 6-12 Meshes for two acoustic bodies 6.2.3 Boundary Conditions The full blowdown model is supported at the CNV lugs and at the CNV skirt through remote points. The circumferential directions of the remote points for the CNV lugs are constrained and all three translational degrees of freedom of the remote point for the CNV skirt are fixed. The acoustic bodies (fluid volumes) and acoustic interfaces (interface between the fluid mesh and the solid mesh) are set using ANSYS parametric design language code. As discussed in Section 5.1.2, a flow acceleration BC is applied to the acoustic body at the break location along with the thrust force on the solid face. The thrust force is divided by the break area and applied as a pressure. A time step of 0.001 seconds is applied as is recommended by the benchmark analysis (Section 5.1.2) to capture the acoustic wave frequency. Since the input data from the thermal-hydraulic analysis has time steps less than 0.001 seconds, the time points for the input data are used when the time increment is less than 0.001 second. If the time increment is greater than 0.001 second, then multiple substeps are used to ensure the simulation time step is within 0.001 seconds. The transient analyses are run for 0.2 © Copyright 2019 by NuScale Power, LLC 75

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 seconds. In the case for the CVCS injection line break analysis, only a subset of time points are used in order to reduce the computational time. These time points are selected to include all of the major peaks and valleys in the flow acceleration and pressure. The time steps are ensured to be 0.001 second or less. 6.2.4 Time Histories and Maximum Results The force and moment time histories of the blowdown model are extracted for each of the 6 cases identified in Table 6-3. The absolute maximum of time histories is calculated for each case. At each location the maximum forces and moments are compared. Only the largest of the forces and moments are reported. Maximum forces and moments acting on 83 of the key structural interfaces or supports are listed in Table 6-4. The forces and moments in Table 6-4 are bounding values for all break and valve opening conditions. The component interface locations for the force and momentum generation are listed in Table 6-5. © Copyright 2019 by NuScale Power, LLC 76

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-4 Maximum forces and moments at component interfaces Component Coord. FX (lbf) FY (lbf) FZ (lbf) MX (in*lbf) MY (in*lbf) MZ (in*lbf) Interface ID System 1 (( 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42

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

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Component Coord. FX (lbf) FY (lbf) FZ (lbf) MX (in*lbf) MY (in*lbf) MZ (in*lbf) Interface ID System 43 (( 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83

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

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-5 List of component interfaces for force and moment generation Component Name X (in) Y (in) Z (in) Interface ID 1 CNV Skirt - Top 0.0 22.5 0 2 CNV Flange 0.0 328.3 0 3 RPV Lower Support 0.0 11.9 0 4 RPV Upper Support, -54.6 500.9 55 Segment -X+Z 5 RPV Upper Support, 54.6 500.9 55 Segment +X+Z 6 RPV Upper Support, 54.6 500.9 -55 Segment +X-Z 7 RPV Upper Support, -54.6 500.9 -55 Segment -X-Z 8 Lower Core Support -32.2 42.3 22 Block -X+Z 9 Lower Core Support 21.5 42.3 32 Block +X+Z 10 Lower Core Support 32.2 42.3 -22 Block +X-Z 11 Lower Core Support -21.5 42.3 -32 Block -X-Z 12 Upper Core Support -47.3 128.7 0 Block -X 13 Upper Core Support 0.0 128.7 47 Block +Z 14 Upper Core Support 47.3 128.7 0 Block +X 15 Upper Core Support 0.0 128.7 -47 Block -Z 16 RPV Flange 0.0 175.3 0 17 Reflector, Lower Core 0.0 47.3 0 Plate 18 Fuel Assemblies, Lower 0.0 51.3 0 Core Plate 19 Fuel Assemblies, Upper 0.0 137.8 0 Core Plate 20 Lower Riser/Upper Core 0.0 147.3 0 Plate 21 Top of Riser Transition 0.0 255.4 0 22 Hanger Brace 1 22.8 565.1 -6 23 Hanger Brace 2 6.1 565.1 -23 24 Hanger Brace 3 -6.1 565.1 -23 25 Hanger Brace 4 -22.8 565.1 -6 26 Hanger Brace 5 -22.8 565.1 6 27 Hanger Brace 6 -6.1 565.1 23 © Copyright 2019 by NuScale Power, LLC 79

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Component Name X (in) Y (in) Z (in) Interface ID 28 Hanger Brace 7 6.1 565.1 23 29 Hanger Brace 8 22.8 565.1 6 30 Upper Riser Hanger 1 -21.3 567.1 6 31 Upper Riser Hanger 2 -5.7 567.1 21 32 Upper Riser Hanger 3 5.7 567.1 21 33 Upper Riser Hanger 4 21.3 567.1 6 34 Upper Riser Hanger 5 21.3 567.1 -6 35 Upper Riser Hanger 6 5.7 567.1 -21 36 Upper Riser Hanger 7 -5.7 567.1 -21 37 Upper Riser Hanger 8 -21.3 567.1 -6 38 NPM Platform Legs, -67.6 855.5 68 Group -X+Z 39 NPM Platform Legs, 67.6 855.5 68 Group +X+Z 40 NPM Platform Legs, 67.6 855.5 -68 Group +X-Z 41 NPM Platform Legs, -67.6 855.5 -68 Group -X-Z 42 CRDM Support Leg, -42.4 692 2 Segment -X 43 CRDM Support Leg, 0 685.8 51 Segment +Z 44 CRDM Support Leg, 45.5 690.5 0 Segment +X 45 CRDM Support Leg, 1.4 692.4 -42 Segment -Z 46 CRDM Nozzle Safe-end 0 709.9 -25 1 47 CRDM Nozzle Safe-end -8.5 709.9 -17 2 48 CRDM Nozzle Safe-end 8.5 709.9 -17 3 49 CRDM Nozzle Safe-end -16.9 709.9 -9 4 50 CRDM Nozzle Safe-end 0 709.9 -9 5 51 CRDM Nozzle Safe-end 16.9 709.9 -9 6 52 CRDM Nozzle Safe-end -25.4 709.9 0 7 53 CRDM Nozzle Safe-end -8.5 709.9 0 8 54 CRDM Nozzle Safe-end 8.5 709.9 0 9 © Copyright 2019 by NuScale Power, LLC 80

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Component Name X (in) Y (in) Z (in) Interface ID 55 CRDM Nozzle Safe-end 25.4 709.9 0 10 56 CRDM Nozzle Safe-end -16.9 709.9 9 11 57 CRDM Nozzle Safe-end 0 709.9 9 12 58 CRDM Nozzle Safe-end 16.9 709.9 9 13 59 CRDM Nozzle Safe-end -8.5 709.9 17 14 60 CRDM Nozzle Safe-end 8.5 709.9 17 15 61 CRDM Nozzle Safe-end 0 709.9 25 16 62 CRDM Rod Travel 0 787.2 -25 Housing Bot 1 63 CRDM Rod Travel -8.5 787.2 -17 Housing Bot 2 64 CRDM Rod Travel 8.5 787.2 -17 Housing Bot 3 65 CRDM Rod Travel -16.9 787.2 -9 Housing Bot 4 66 CRDM Rod Travel 0 787.2 -9 Housing Bot 5 67 CRDM Rod Travel 16.9 787.2 -9 Housing Bot 6 68 CRDM Rod Travel -25.4 787.2 0 Housing Bot 7 69 CRDM Rod Travel -8.5 787.2 0 Housing Bot 8 70 CRDM Rod Travel 8.5 787.2 0 Housing Bot 9 71 CRDM Rod Travel 25.4 787.2 0 Housing Bot 10 72 CRDM Rod Travel -16.9 787.2 9 Housing Bot 11 73 CRDM Rod Travel 0 787.2 9 Housing Bot 12 74 CRDM Rod Travel 16.9 787.2 9 Housing Bot 13 75 CRDM Rod Travel -8.5 787.2 17 Housing Bot 14 76 CRDM Rod Travel 8.5 787.2 17 Housing Bot 15 77 CRDM Rod Travel 0 787.2 25 Housing Bot 16 © Copyright 2019 by NuScale Power, LLC 81

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Component Name X (in) Y (in) Z (in) Interface ID 78 UCP Support Block Bolts 0 142.3 0 79 Top of Reflector Block 1 0 62.4 0 80 Top of Reflector Block 2 0 77.6 0 81 Top of Reflector Block 3 0 92.7 0 82 Top of Reflector Block 4 0 107.8 0 83 Top of Reflector Block 5 0 123 0 Forces and moments are generated for 22 internal sections of NPM components. Time histories are extracted for each node. Maximum forces and moments on the CNV, RPV, riser assemblies, and the core barrel assembly are summarized in Table 6-6. The forces and moments in Table 6-6 are bounding values for all break and valve opening conditions. Locations of the component section for force and momentum are listed in Table 6-7. © Copyright 2019 by NuScale Power, LLC 82

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-6 Maximum forces and moments on containment vessel, reactor pressure vessel, riser, and core barrel assembly Component Coord. FX (lbf) FY (lbf) FZ (lbf) MX (in*lbf) MY (in*lbf) MZ (in*lbf) Section ID System 1 (( 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22

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

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-7 List of component sections for force and moment generation Component Name Elevation, Y (in) Section ID 1 CNV - Elevation 37.8 37.8 2 CNV - Elevation 138.5 138.5 3 CNV - Elevation 179.9 179.9 4 CNV - Elevation 328.3 328.3 5 CNV - Elevation 512.9 512.9 6 CNV - Elevation 673.0 673.0 7 CNV - Elevation 816.0 816.0 8 CNV - Elevation 834.8 834.8 9 RPV - Elevation 20.6 20.6 10 RPV - Elevation 43.9 43.9 11 RPV - Elevation 91.9 91.9 12 RPV - Elevation 160.6 160.6 13 RPV - Elevation 270.8 270.8 14 RPV - Elevation 352.3 352.3 15 RPV - Elevation 410.0 410.0 16 RPV - Elevation 543.1 543.1 17 RPV - Elevation 625.5 625.5 18 RPV - Elevation 671.0 671.0 19 Lower Riser - Elevation 158.3 158.3 20 Lower Riser - Elevation 236.3 236.3 21 Core Barrel - Elevation 47.3 47.3 22 Core Barrel - Elevation 142.3 142.3 Table 6-8 identifies the largest five forces acting on RPV, CNV, and RVI in any direction, and Table 6-9 identifies the five largest moments on RPV, CNV and RVI in any direction. The corresponding event of valve opening or CVCS injection line break associated with each of the forces and moments is identified. The highest forces and moments result from one RVV opening cases. The magnitude differences of the forces and moments for three RVV locations of RVV16, RVV17, and RVV83 are small and have no significance. The forces on the RPV are one order of magnitude higher than those on the CNV and two orders of magnitude higher than those on the RVI. All of the largest forces act in the Y direction and generally, the forces in the Y direction are greater than in the X and Z directions at any location for a particular case. The exceptions for this are at the lower core plate, top of reflector blocks 1-5, upper riser hangers 1-8 (see Table 6-4), the lower RPV head (or RPV at Elevation 20.6) and core barrel at Elevation 142.3 (see Table 6-6). These five regions do not bear the load from the pressure wave traveling vertically inside the RPV and therefore experience comparable or higher lateral forces than vertical for the cases analyzed. The five largest forces result from the one RVV opening (RVV16, RVV17, and RVV83) and occur within the RPV, from the region just below the feedwater plenum up to just above the PZR heater nozzle. One RVV opening provides the largest forces due to the high mass flow rates and correspondingly high fluid accelerations generated by this event. © Copyright 2019 by NuScale Power, LLC 84

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 The five largest moments generated by the scenarios analyzed come from the one RVV opening cases (RVV83 and RVV16). The largest forces and moments act on the RPV, from the region 50 inches above feedwater plenum up to just below main steam nozzle. The region of the RPV with the highest forces bounds the locations with the highest moments for these events. It is noted that one RRV opening cases (RRV1 or RRV2) has highest moments in the lower core plate and core barrel. Table 6-8 Summary of largest five forces for RPV, CNV, and RVI Interface/Support Location Force Fx, Fy or Fz (lbf) Corresponding Event Largest Forces on RPV RPV-Elevation 625.5 (( RVV17 RPV-Elevation 671.0 RVV83 RPV-Elevation 270.8 RVV16 RPV-Elevation 352.3 RVV16 RPV-Elevation 543.1 }}2(a),(c) RVV17 Largest Forces on CNV CNV-Elevation 512.9 (( RVV83 CNV-Elevation 673.0 RVV83 CNV-Elevation 179.9 RVV83 CNV-Elevation 816.0 RVV83 CNV Flange /CNV-Elevation 328.3 }}2(a),(c) RVV83 Largest Forces on RVI UCP Support Block Bolts (( RVV83 Lower Riser-Elevation 158.3 RVV17 Top of riser Transition RVV17 Lower Riser-Elevation 236.3 RVV17 Reflector-Lower Core Plate }}2(a),(c) RRV2 © Copyright 2019 by NuScale Power, LLC 85

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-9 Summary of largest five moments for RPV, CNV, and RVI Interface/Support Location Moment Mx, My or Mz (in*lbf) Corresponding Event Largest Moments on RPV RPV-Elevation 543.1 (( RVV83 RPV-Elevation 543.1 RVV16 RPV-Elevation 410.0 RVV83 RPV-Elevation 410.0 RVV16 RPV-Elevation 352.3 }}2(a),(c) RVV83 Largest Moments on CNV CNV-Elevation 512.9 (( RVV83 CNV-Elevation 512.0 RVV16 CNV-Elevation 328.3 RVV83 CNV-Elevation 328.3 RVV16 CNV-Elevation 328.3 }}2(a),(c) RVV17 Largest Moments on RVI Reflector-Lower Core Plate (( RRV2 Reflector-Lower Core Plate RRV1 Reflector-Lower Core Plate RRV1 Core barrel-Elevation 47.3 RRV1 Core barrel-Elevation 47.3 }}2(a),(c) RRV2 Table 6-10 provides the largest five forces and moments on RPV, CNV, and RVI due to the CVCS injection line break, including riser locations near to where the pressure wave originates. The forces near the riser location for this event are smaller than are experienced for the case of the one RVV opening per Table 6-8. For the CVCS injection line break event, the forces and moments on many parts of the RPV and reactor vessel internals are close in magnitude; implying that the pressure wave imparts similar forces on the structures. The case of the pressure wave originating from the upper riser assembly is not bounding for mechanical design. © Copyright 2019 by NuScale Power, LLC 86

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Table 6-10 Largest forces and moments for CVCS injection line break for RPV, CNV and RVI Interface/Support Force Fx, Fy or Fz Interface/Support Moment Mx, My or Location (lbf) Location Mz (in-lbf) Largest Forces and Moments on RPV RPV-Elevation 270.8 (( RPV-Elevation 43.9 (( RPV-Elevation 625.5 RPV-Elevation 352.3 RPV-Elevation 671.0 RPV-Elevation 543.1 RPV-Elevation 352.3 RPV-Elevation 91.9 2(a),(c) RPV-Elevation 410.1 }} RPV-Elevation 410.1 }}2(a),(c) Largest Forces and Moments on CNV CNV Skirt - Top (( CNV Flange (( CNV - Elevation 37.8 CNV - Elevation 328.3 CNV - Elevation 138.5 CNV - Elevation 512.9 CNV - Elevation 179.9 CNV - Elevation 179.9 CNV Flange/CNV -

                                                   }}2(a),(c)   CNV - Elevation 138.5                      }}2(a),(c)

Elevation 328.3 Largest Forces and Moments on RVI Reflector, Lower Core UCP Support Block Bolts (( (( Plate Fuel Assemblies, Lower Core Barrel -Elevation Core Plate 47.3 Core Barrel - Elevation 47.3 Top of Reflector Block 1 Core Barrel -Elevation Top of Riser Transition 142.3 Fuel Assemblies, Upper 2(a),(c) UCP Support Block

                                                  }}                                                       }}2(a),(c)

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Acoustic pressure time histories are reported for 38 regions of the RCS and containment. Figure 6-13 provides a time history plot of the maximum acoustic pressure difference between any point on the top of the baffle plate and any point on the bottom of the baffle plate for the RRV1 opening case. The maximum differential pressure across the baffle plate remains below 15 psi. ((

                                                                                                        }}2(a),(c),ECI Figure 6-13 Differential pressure time history across baffle plate for one reactor recirculation valve (RRV1) opening
  © Copyright 2019 by NuScale Power, LLC 88

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 7.0 References 7.1 Source Documents 7.1.1 American Society of Mechanical Engineers, Quality Assurance Program Requirements for Nuclear Facility Applications, NQA-1-2008, NQA-1a-2009 Addenda, as endorsed by Regulatory Guide 1.28, Rev. 4. 7.1.2 U.S. Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Facilities, Appendix B, Part 50, Chapter 1, Title 10, Energy, (10 CFR 50). 7.2 Referenced Documents 7.2.1 U.S. Nuclear Regulatory Commission, Asymmetric Blowdown Loads on PWR Primary Systems, Resolution of Generic Task Action Plan A-2, NUREG-0609, January 1981. 7.2.2 Watkins, J.C., R5FORCE/MOD3s: A Program to Compute Fluid-Induced Forces Using Hydrodynamic Output from the RELAP5/MOD3 Code, Idaho National Engineering Laboratory, September 1990. 7.2.3 Timperi, A., et al., Validation of Fluid-Structure Interaction Calculations in a Large Break Loss of Coolant Accident, ICONE16-48206, 16th International Conference on Nuclear Engineering, Orlando, FL, May 2008. 7.2.4 Khknen, J., P. Varpasuo, M. Vuorinen, Analyzing HDR Fluid-Structure-Interaction Pipe Break Case with Acoustic Finite Element Method, ICONE19-43205, 19th International Conference on Nuclear Engineering, Chiba, Japan, May 2011. 7.2.5 Wolf, L., Experimental Results of Coupled Fluid-Structure Interactions during Blowdown of the HDR-vessel and Comparisons with Pre- and Post-test Predictions, Nuclear Engineering and Design, Vol. 70, 269-308, 1982. 7.2.6 Prelewicz, D.A., Hydraulic Pressure Pulses with Structural Flexibility: Test and Analysis, WAPD-TM-1227, Bettis Atomic Power Laboratory, West Mifflin, PA, April 1976. 7.2.7 Mitsubishi Heavy Industries, Ltd, Subcompartment Analyses for US-APWR Design Confirmation, US-APWR, MUAP-07031-NP, Rev. 1, Nonproprietary, October 2009. 7.2.8 Brandt, T., et al., Validation of fluid-structure interaction simulation environment in analysis of large-break loss of coolant accident, System Simulation and Scientific Computing, pp. 520-527, October 2008. © Copyright 2019 by NuScale Power, LLC 89

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 7.2.9 Allemann, R.T., et al., Battelle Pacific Northwest Laboratories, AEC Research and Development Report: Coolant Blowdown Studies of a Reactor Simulator Vessel containing a Simulated Reactor Core, BNWL-1524, June 1971. 7.2.10 Christensen, D.D., EG&G Idaho, Inc., United States Standard Problem 5, Final Report, PG-R-77-44, December 1977. 7.2.11 Delaney, H.M., EG&G Idaho, Inc., United States Standard Problem 6 and International Standard Problem 4, Final Report, CVAP-TR-4-78, March 1978. 7.2.12 Brockett, G.F., et al., EG&G Idaho, Inc., Experimental Investigations of Reactor System Blowdown, IN-1348, September 1970. 7.2.13 Robinson, H.C., EG&G Idaho, Inc., Experimental Data Report for LOFT Nonnuclear Test LI-2, TREE-NUREG-1026, January 1977. 7.2.14 Fritz, R.J., and E. Kiss, General Electric Company, The Vibration Response of a Cantilevered Cylinder Surrounded by an Annular Fluid, KAPL-M-6539 (RJF-10), February 24, 1966. 7.2.15 Prelewicz, D.A., Hydraulic Pressure Pulses with Structural Flexibility: Test and Analysis, WAPD-TM-1227, Bettis Atomic Power Laboratory, West Mifflin, PA, April 1976. 7.2.16 Electric Power Research Institute, Comparison of the Calculation of HDR RPV-1 Blowdown Loads for Test V32 with the Experimental Data, NP-3677, EPRI, Palo Alto, CA, August 1984. 7.2.17 INEEL-EXT-98-00834-V2, RELAP5-3D© Code Manual Volume II: Users Guide and Input Requirements, Rev. 4.1, September 2013. 7.2.18 INEEL-EXT-98-00834-V3, RELAP5-3D© Code Manual Volume III: Developmental Assessment, Rev. 4.1, September 2013. 7.2.19 SwUM-0304-17023, NRELAP5 Code Manual Theory Manual: Models, correlations and Solution Methods, Rev. 3. 7.2.20 INEEL-EXT-98-00834-V5, RELAP5-3D© Code Manual Volume V: Users Guidelines, Rev. 4.1, September 2013. 7.2.21 INEEL-EXT-98-00834-V2, RELAP5-3D© Code Manual Volume II Appendix A: RELAP5-3D Input Data Requirements, Rev. 4.1, September 2013. 7.2.22 INEEL-EXT-98-00834-V4, RELAP5-3D© Code Manual Volume IV: Models and Correlations, Rev. 4.1, September 2013. 7.2.23 Schumann, U., Experimental and Computed Results for Fluid-Structure Interactions with Impacts in the HDR Blowdown Experiment, Nuclear Engineering and Design 73 (1982): 303-317. © Copyright 2019 by NuScale Power, LLC 90

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 7.2.24 Moody, F. J., Introduction to Unsteady Thermofluid Mechanics, John Wiley & Sons, New York, NY, 1990. 7.2.25 Electric Power Research Institute, Two-Phase Jet Modeling and Data Comparison, NP-4362, EPRI, Palo Alto, CA, March 1986. 7.2.26 Wolf, L., Design Report for the HDR-RPV-I Blowdown Experiments V31.2, V32, V33, and V34 with Specifications for the Pretest Computation, HDR Safety Program, Report No. 3.243/81, Battelle Institute, Frankfurt, 1981. 7.2.27 ANSYS Mechanical, Release 16.0, ANSYS, Inc., Canonsburg, PA 2015. © Copyright 2019 by NuScale Power, LLC 91

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix A. NRELAP5 Heissdampf Reactor and Jet Impingement Test Results Figure A-1 Heissdampf reactor Test V31.1 Sensitivity Case A2, mass flow rate © Copyright 2019 by NuScale Power, LLC 92

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 2500 2000 Flow Rate (kg/s) 1500 RM3204 RM3103 1000 RM3000 RELAP Upper Bound RELAP 500 RELAP Lower Bound 0 0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 Time (sec) Figure A-11 Mass flow rate for Heissdampf reactor Test V31.1 © Copyright 2019 by NuScale Power, LLC 102

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 3000 2500 2000 Flow Rate (kg/s) 1500 RM 3093 1000 NRELAP5 Upper Bound NRELAP5 500 NRELAP5 Lower Bound 0 0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 Time (sec) Figure A-12 Mass flow rate for Heissdampf reactor Test V32 © Copyright 2019 by NuScale Power, LLC 103

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 600.00 500.00 Test 11 RELAP Upper Bound RELAP 400.00 RELAP Lower Bound Flow Rate (kg/s) 300.00 200.00 100.00 0.00 0 10 20 30 40 50 60 Time (seconds) Figure A-13 Mass flow rate for Marviken jet impingement test-11 © Copyright 2019 by NuScale Power, LLC 104

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 30.00 25.00 Test 11 RELAP Upper Bound RELAP 20.00 RELAP Lower Bound Density (kg/m3) 15.00 10.00 5.00 0.00 0 10 20 30 40 50 60 Time (seconds) Figure A-14 Density for Marviken jet impingement test-11 © Copyright 2019 by NuScale Power, LLC 105

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 600.00 500.00 Test 11 RELAP Upper Bound RELAP 400.00 RELAP Lower Bound Force (kN) 300.00 200.00 100.00 0.00 0 10 20 30 40 50 60 Time (seconds) Figure A-15 Thrust force for Marviken jet impingement test-11 © Copyright 2019 by NuScale Power, LLC 106

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix B. Pressure Comparison for Bettis Hydraulic Pressure Pulse 1200 1000 Pressure, psia 800 600 400 200 Experiment - Rigid Test Section ANSYS - Rigid Test Section 0 0 5 10 15 20 25 30 Time, msec 1200 1000 Pressure, psia 800 600 400 200 Experiment - Rigid Test Section ANSYS - Rigid Test Section 0 0 5 10 15 20 25 30 Time, msec Figure B-1 Pressure at top and bottom transducers for Run 10S © Copyright 2019 by NuScale Power, LLC 107

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 1200 1000 Pressure, psia 800 600 400 Experiment - Flexible Test Section 200 ANSYS - Flexible Test Section 0 0 5 10 15 20 25 30 35 Time, msec 1200 1000 Pressure, psia 800 600 400 Experiment - Flexible Test Section 200 ANSYS - Flexible Test Section 0 0 5 10 15 20 25 30 35 Time, msec Figure B-2 Pressure at top and bottom transducers for Run 10F © Copyright 2019 by NuScale Power, LLC 108

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 1300 Experiment - Rigid 1200 Test Section FLASH Rigid Test Section 1100 ANSYS - Rigid Test Section 1000 900 Top transducer Pressure, psia 800 700 600 500 400 300 200 100 0 0 1 2 3 4 5 6 7 8 9 10 Time, msec 1400 1300 Experiment - Rigid Test Section 1200 ANSYS - Rigid Test 1100 Section 1000 Pressure, psia 900 Bottom transducer 800 700 600 500 400 300 200 100 0 0 1 2 3 4 5 6 7 8 9 10 Time, msec Figure B-3 Pressure at top and bottom transducers for Run 20S © Copyright 2019 by NuScale Power, LLC 109

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 1000 Experiment - Flexible Test Section 900 FLASH Flexible Test Section ANSYS - Flexible Test 800 Section 700 Pressure, psia Top transducer 600 500 400 300 200 100 0 0 1 2 3 4 5 6 7 8 9 10 11 12 Time, msec 1100 Experiment - Flexible Test 1000 Section ANSYS - Flexible Test Section 900 800 Pressure, psia 700 Bottom transducer 600 500 400 300 200 100 0 0 1 2 3 4 5 6 7 8 9 10 11 12 Time, msec Figure B-4 Pressure at top and bottom transducers for Run 20F © Copyright 2019 by NuScale Power, LLC 110

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix C. Thrust Force and Fluid Acceleration Boundary Conditions 5,000 4,000 Acceleration, m/s2 3,000 RELAP 2,000 1,000 0

                       -1,000 0.00                  0.02          0.04              0.06            0.08               0.10 Time, s Figure C-1 Flow acceleration at the break location for Heissdampf reactor Test V29.2 450,000 400,000 350,000 Thrust Force, N 300,000 250,000 200,000 150,000                                                                   RELAP - Lower Bound 100,000                                                                   RELAP 50,000 RELAP - Upper Bound 0

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure C-3 Flow acceleration at the break location for Heissdampf reactor Test V31.1 450,000 400,000 350,000 Thrust Force, N 300,000 250,000 200,000 150,000 Experimental RELAP - Lower Bound 100,000 RELAP 50,000 RELAP - Upper Bound 0 0.00 0.02 0.04 0.06 0.08 0.10 Time, s Figure C-4 Thrust force at the break location for Heissdampf reactor Test V31.1 © Copyright 2019 by NuScale Power, LLC 112

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 25,000 20,000 Acceleration, m/s2 15,000 Experimental 10,000 RELAP - Lower Bound RELAP RELAP - Upper Bound 5,000 0

                       -5,000 0.00          0.02    0.04              0.06            0.08               0.10 Time, s Figure C-5 Flow acceleration at the break location for Heissdampf reactor Test V32 500,000 450,000 400,000 350,000 Thrust Force, N 300,000 250,000 200,000 150,000                                                    Experimental RELAP - Lower Bound 100,000                                                    RELAP 50,000                                              RELAP - Upper Bound 0

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix D. ANSYS Heissdampf Reactor Results with Pressure Boundary Condition Figure D-1 Heissdampf reactor Test V31.1 pressure, BP9109 (1330, 90°, 8850) (Fig. 4-5 of Reference 7.2.16) Figure D-2 Heissdampf reactor Test V31.1 pressure, KP0009 (1307, 90°, 8850) (Fig. 4-12 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 114

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure D-3 Displacement comparison for Heissdampf reactor Test V32, KS1030 (1307, 90°, 2265) (Fig. A-118 of Reference 7.2.16) Figure D-4 Displacement comparison for Heissdampf reactor Test V32, KS1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 115

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix E. Heissdampf Reactor V29.2 ANSYS Results, Flow Acceleration Boundary Condition Figure E-1 Displacement at upper part of the core barrel for V29.2, KS1008 (1330, 90°, 8410) (Fig. 26 of Reference 7.2.5) Figure E-2 Outside reactor pressure vessel displacement for V29.2, BS0106 (1590, 90°, 7350) (Fig. 28 of Reference 7.2.5) © Copyright 2019 by NuScale Power, LLC 116

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure E-3 Outside reactor pressure vessel displacement for V29.2, BS0107 (1590, 180°, 7350) (Fig. 28 of Reference 7.2.5) Figure E-4 Outside reactor pressure vessel displacement for V29.2, BS0108 (1590, 270°, 7350) (Fig. 28 of Reference 7.2.5) © Copyright 2019 by NuScale Power, LLC 117

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix F. Heissdampf Reactor V31.1 ANSYS Results, Flow Acceleration Boundary Condition Figure F-1 Outside reactor pressure vessel for V31.1, BS0106 (1590, 90°, 7350) (Fig. 28 of Reference 7.2.5) Figure F-2 Outside reactor pressure vessel for V31.1, BS0107 (1590, 180°, 7350) (Fig. 28 of Reference 7.2.5) © Copyright 2019 by NuScale Power, LLC 118

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure F-3 Outside reactor pressure vessel displacement for V31.1, BS0108 (1590, 270°, 7350) (Fig. 28 of Reference 7.2.5) Figure F-4 Pressure for V31.1, BP9109 (1330, 90°, 8850) (Fig. 4-5 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 119

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure F-5 Pressure for V31.1, BP9117 (1330, 270°, 8850) (Fig. 4-6 of Reference 7.2.16) Figure F-6 Pressure for V31.1, BP9133 (1330, 88°, 5505) (Fig. 4-7 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 120

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure F-7 Pressure for V31.1, BP9140 (1330, 90°, 2300) (Fig. 4-8 of Reference 7.2.16) Figure F-8 Pressure for V31.1, BP8301 (0, 0°, 10370) (Fig. 4-10 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 121

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure F-9 Differential pressure for V31.1, KP0009 (1307, 90°, 8850) (Fig. 4-12 of Reference 7.2.16) Figure F-10 Hoop strain for V31.1, at core barrel outside diameter KA2009 (1330, 90°, 8850) (Fig. 4-15 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 122

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure F-11 Axial strain for V31.1, at core barrel outside diameter KA3008 (1330, 90°, 8850) (Fig. 4-16 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 123

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Appendix G. Heissdampf Reactor V32 ANSYS Results, Flow Acceleration Boundary Condition Figure G-1 Outside reactor pressure vessel displacement for V32, BS0106 (1590, 90°, 7350) (Fig. 8 of Reference 7.2.23) Figure G-2 Outside reactor pressure vessel displacement for V32, BS0116 (1590, 90°, 5550) (Fig. A-47 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 124

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-3 Core barrel displacement for V32, KS1013 (1307, 90°, 7195) (Fig. A-112 of Reference 7.2.16) Figure G-4 Core barrel displacement for V32, KS1030 (1307, 90°, 2265) (Fig. A-118 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 125

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-5 Core barrel displacement for V32, KS1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) Figure G-6 Core barrel hoop strain for V32, KA2008 (1330, 90°, 8845) (Fig. A-66 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 126

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-7 Core barrel axial strain for V32, KA3009 (1330, 90°, 8825) (Fig. A-71 of Reference 7.2.16) Figure G-8 Sensitivity study: core barrel displacement for V32, KA1030 (1307, 90°, 2265) (Fig. A-118 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 127

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-9 Sensitivity study: core barrel displacement for V32, KA1032 (1307, 270°, 2265) (Fig. A-120 of Reference 7.2.16) Figure G-10 Sensitivity study: core barrel hoop strain for V32, KA2008 (1330, 90°, 8845) (Fig. A-66 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 128

NuScale Power Module Short-Term Transient Analysis TR-1016-51669-NP Rev. 1 Figure G-11 Sensitivity study: core barrel axial strain for V32, KA3009 (1330, 90°, 8825) (Fig. A-71 of Reference 7.2.16) © Copyright 2019 by NuScale Power, LLC 129

LO-0719-66357 : Affidavit of Zackary W. Rad, AF-0719-66358 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 technical report reveals distinguishing aspects about the process and analytical results by which NuScale performed its NPM Short-Term Transient Analysis. NuScale has performed significant research and evaluation to develop a basis for this analysis 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 technical report titled NuScale Power Module Short-Term Transient Analysis. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. AF-0719-66358 Page 1 of 2

(5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on July 30, 2019. Zackary W. Rad AF-0719-66358 Page 2 of 2}}

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