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LLC - Submittal of TR-0818-61384, Rev. 0, Pipe Rupture Hazards Analysis.
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Issue date:	10/03/2018
From:	Bergman T; NuScale
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	LO-0918-61827 TR-0818-61384, Rev. 0
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LO-0918-61827 October 3, 2018 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 "Pipe Rupture Hazards Analysis ," TR-0818-61384 , Revision 0

REFERENCE:

1. NuScale Letter to the NRC, "NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application ," dated December 31 , 2016 NuScale Power, LLC (NuScale) submitted a Design Certification Application (DCA) for its Integral Small Modular Reactor Design in Reference 1. Part 2 of Tier 2 of the DCA, NuScale Final Safety Analysis Report (FSAR) Section 3.6 , discussed the performance of a Pipe Rupture Hazard Analysis (PRHA) and provided reference to the NuScale technical report "Pipe Rupture Hazards Analysis" for details of the evaluation.

The purpose of this letter is to provide the NuScale Technical Report "Pipe Rupture Hazards Analysis ,"

TR-0818-61384 , Revision 0. This technical report provides supplementary information , data and analyses for the NuScale PRHA.

As discussed in the monthly closed teleconference to discuss FSAR Section 3.6 held on August 16, 2018 , in order to facilitate NRC review of available material , NuScale is providing Revision O of the subject technical report with this letter, and will provide the information identified in the report as "Later" in Revision 1 of this technical report. contains the proprietary version of the report entitled "Pipe Rupture Hazards 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. contains the nonproprietary version of the report entitled "Pipe Rupture Hazards Analysis. "

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, NuScale Power, LLC 1100 NE Circle Blvd , Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0918-61827 Page 2 of 2 10/03/18 Distribution: Samuel Lee, NRC, OWFN-8G94 Gregory Cranston, NRC, OWFN-8G94 Omid Tabatabai, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A Enclosure 1: Pipe Rupture Hazards Analysis, TR-0818-61384-P, Revision 0, proprietary version Enclosure 2: Pipe Rupture Hazards Analysis, TR-0818-61384-NP, Revision 0, nonproprietary version Enclosure 3: Affidavit of Thomas A. Bergman, AF-0918-61828 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0918-61827 :

Pipe Rupture Hazards Analysis, TR-0818-61384-P, Revision 0, 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-0918-61827 :

Pipe Rupture Hazards Analysis, TR-0818-61384-NP, Revision 0, 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

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report Pipe Rupture Hazards Analysis October 2018 Revision 0 Docket No.: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC, and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S.

Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

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

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008742.

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.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report CONTENTS Abstract ....................................................................................................................................... 1 Executive Summary .................................................................................................................... 2 1.0 Introduction ..................................................................................................................... 3 1.1 Purpose ................................................................................................................. 3 1.2 Scope .................................................................................................................... 4 1.3 Abbreviations and Definitions ................................................................................ 5 1.4 References ............................................................................................................ 8 1.4.1 Source Documents ................................................................................................ 8 1.4.2 Referenced Documents ......................................................................................... 9 1.4.3 Other References ................................................................................................ 10 2.0 Background ................................................................................................................... 12 2.1 NuScale Design Features Relevant to Pipe Rupture Hazards Analysis .............. 12 2.2 Regulations and Guidance .................................................................................. 15 2.2.1 History of HELB Effects Analysis Methodology ................................................... 15 2.2.2 NRC Guidance .................................................................................................... 16 2.2.3 Branch Technical Position 3-3 ............................................................................. 20 2.2.4 Standard Review Plan Section 3.6.2 ................................................................... 23 2.2.5 Standard Review Plan Section 3.6.3 ................................................................... 28 3.0 Methodology .................................................................................................................. 30 3.1 General Approach ............................................................................................... 30 3.1.1 Essential Functions ............................................................................................. 32 3.2 Description of Systems Important to Reactor Shutdown and Core Cooling ........ 32 3.2.1 Reactor Coolant System ..................................................................................... 34 3.2.2 Module Protection System .................................................................................. 34 3.2.3 Neutron Monitoring System ................................................................................. 35 3.2.4 Chemical and Volume Control System ................................................................ 35 3.2.5 Control Rod Assembly and Control Rod Drive System ....................................... 35 3.2.6 Containment System ........................................................................................... 36 3.2.7 Decay Heat Removal System.............................................................................. 37 3.2.8 Emergency Core Cooling System ....................................................................... 37 3.2.9 Ultimate Heat Sink ............................................................................................... 38 3.2.10 Post-Accident Monitoring .................................................................................... 39

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report 3.3 Systems with Potential for High- or Moderate-Energy Line Ruptures ................. 41 3.3.1 Reactor Coolant System ..................................................................................... 42 3.3.2 Containment System ........................................................................................... 42 3.3.3 Chemical Volume and Control System ................................................................ 42 3.3.4 Emergency Core Cooling System ....................................................................... 42 3.3.5 Steam Generating System .................................................................................. 42 3.3.6 Main Steam System ............................................................................................ 43 3.3.7 Feedwater System .............................................................................................. 43 3.3.8 Decay Heat Removal System.............................................................................. 43 3.3.9 Reactor Component Cooling Water System ........................................................ 43 3.3.10 Auxiliary Boiler System ........................................................................................ 44 3.3.11 Module Heatup System ....................................................................................... 44 3.4 Break Characteristics .......................................................................................... 44 3.5 Restraints, Barriers, and Shields ......................................................................... 50 3.5.1 Pipe Whip Restraints ........................................................................................... 50 3.5.2 Susceptibility of Essential Structures, Systems, and Components by Plant Location ............................................................................................................... 54 3.6 Break Exclusion ................................................................................................... 60 3.6.1 Leakage Cracks .................................................................................................. 61 3.7 Leak-Before-Break .............................................................................................. 62 3.7.1 Inside the Containment Vessel ............................................................................ 62 3.7.2 In the NuScale Power Module Bay ...................................................................... 63 3.7.3 In the Reactor Building ........................................................................................ 63 3.8 Separation ........................................................................................................... 63 3.8.1 Inside the Containment Vessel ............................................................................ 63 3.9 Analysis Methodology ......................................................................................... 64 3.9.1 Determining Break Locations .............................................................................. 65 3.9.2 Parameters Affecting Severity of High-Energy Line Break Effects ...................... 65 3.9.3 Blast Effects......................................................................................................... 66 3.9.4 Blowdown Thrust Loads ...................................................................................... 68 3.9.5 Pipe Whip Loads ................................................................................................. 70 3.9.6 Jet Zone of Influence ........................................................................................... 72 3.9.7 Jet Impingement Loads ....................................................................................... 74

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report 3.9.8 Pressurization Caused by High-Energy Line Breaks........................................... 75 3.9.9 Effects of Leakage Cracks................................................................................... 77 4.0 Results ........................................................................................................................... 78 4.1 Postulated Break Locations ................................................................................. 78 4.1.1 In the Containment Vessel .................................................................................. 78 4.1.2 In the NuScale Power Module Bay ...................................................................... 78 4.1.3 In the Reactor Building ........................................................................................ 78 4.2 Blast Effects......................................................................................................... 78 4.2.1 In the Containment Vessel .................................................................................. 79 4.2.2 In the NuScale Power Module Bay ...................................................................... 79 4.2.3 In the Reactor Building ........................................................................................ 79 4.3 Pipe Whip ............................................................................................................ 79 4.3.1 In the Containment Vessel .................................................................................. 79 4.3.2 In the NuScale Power Module Bay ...................................................................... 79 4.3.3 In the Reactor Building ........................................................................................ 79 4.4 Jet Impingement .................................................................................................. 80 4.4.1 In the Containment Vessel .................................................................................. 80 4.4.2 In the NuScale Power Module Bay ...................................................................... 80 4.4.3 In the Reactor Building ........................................................................................ 80 4.5 Subcompartment pressurization .......................................................................... 80 4.5.1 In the Containment Vessel .................................................................................. 80 4.5.2 In the NuScale Power Module Bay ...................................................................... 80 4.5.3 In the Reactor Building ........................................................................................ 81 5.0 Conclusions ................................................................................................................... 82 Appendix A. Break Exclusion - Compliance with Regulatory Acceptance Criteria.......... 84 Appendix B. Dynamic Amplification and Potential for Resonance .................................... 85 Appendix C. Pipe Whip ........................................................................................................... 93 Appendix D. Subcompartment Pressurization ................................................................... 113 Appendix E. Jet Impingement .............................................................................................. 114 Appendix F. Blast Effects ..................................................................................................... 130

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report TABLES Table 1-1. Abbreviations ............................................................................................................ 5 Table 1-2. Definitions ................................................................................................................. 7 Table 3-1. Safety-related and essential parts of structures, systems, and components vulnerable to break effects ...................................................................................... 33 Table 3-2. Separation group B and C post-accident monitoring Type B & C instruments inside containment ............................................................................................................ 40 Table 3-3. High-energy and moderate-energy system piping characteristics .......................... 46 Table 3-4. Characteristics of blowdown at postulated break locations .................................... 48 Table 3-5. Comparison of sizes of whipping pipe to potential barriers for high-energy line breaks in the containtment vessel ........................................................................... 53 Table 3-6. Comparison of main steam system and feedwater system piping in containment penetration area ...................................................................................................... 59 Table 3-7. Break exit plane parameters ................................................................................... 72 Table 5-1. Summary of approach and result for line break assessment by plant area ............ 83 Table B-2. Range of potential resonance region ...................................................................... 87 Table B-3. Wavelengths of downstream propagating waves .................................................... 88 Table C-1. Comparison of sizes of whipping pipe to potential targets for high-energy line breaks in the containment vessel ............................................................................ 94 Table C-2. Maximum hinge length Lh to avoid pipe whip ........................................................ 105 Table C-3. Example of simplified pipe whip analysis .............................................................. 110 Table C-4. Reactor building wall penetration depth (inches) for main steam system pipe whip impact ................................................................................................................... 111 Table E-1. CVCS jet impingement pressure versus. distance for limited separation in CNV . 123 Table E-2. CVCS steam jet impingement pressure versus distance ...................................... 125 Table E-3. Shape factors for jet impingement ........................................................................ 126 Table F-1. Summary of average error from validation analysis .............................................. 135 Table F-2. Overview of blast CFD modeling inside the CNV ................................................. 139 Table F-3. Maximum total forces on selected components for blasts in the containment vessel

.............................................................................................................................. 146 Table F-4. Overview of modeling scheme for blast analysis in reactor building ..................... 149 Table F-5. Key to reactor building SSC of interest for blast effects ........................................ 150 Table F-6. Peak blast wave forces on selected SSC ............................................................. 153 FIGURES Figure 3-1. Flowchart of methodology for evaluation of line breaks .......................................... 31 Figure 3-2. Reactor pressure vessel head penetrations and break locations ........................... 40 Figure 3-3. Containment vessel head penetrations and break locations (breaks on underside)

................................................................................................................................ 41 Figure 3-4. Adjacent NuScale Power Module overlap of main steam system and feedwater system piping in the Reactor Building pipe gallery ................................................. 57 Figure 3-5. Application of Nuclear Regulatory Commission break location guidance in the NuScale power module bay and the reactor building ............................................. 61 Figure 3-6. Characteristics of a blast wave (Reference 1.4.3.13) ............................................ 67 Figure C-1. Visual scale comparison of NPS 2 Sch. 160 pipe to SSC wall thickness ............... 95 Figure C-2. Separation of reactor coolant system line terminal ends from emergency core cooling system valves ............................................................................................. 98 Figure C-3. Reactor coolant system breaks on underside of containment vessel head ............ 99

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Licensing Technical Report Figure C-4. Pipe whip example ................................................................................................ 100 Figure C-5. Mass moment of inertia about centroidal axis ....................................................... 104 Figure C-6. Mass moment of inertia about hinge location. ...................................................... 104 Figure C-7. Potential high-energy line break locations in pipe gallery ..................................... 109 Figure E-1. Jet expansion and zone of influence for circumferential break with limited separation ............................................................................................................. 122 Figure E-2. Jet ZOI and expansion for circumferential break with full separation in CNV ....... 123 Figure E-3. Jet Impingement on flat plate ................................................................................ 127 Figure E-4. Expanding jet impingement on a flat plate ............................................................ 127 Figure E-5. Expanding jet impingement on a cylinder ............................................................. 128 Figure F-1 Characteristic shape of a blast wave and decay with time ................................... 131 Figure F-2 Blast wave reflection coefficient ............................................................................ 133 Figure F-3. Verification and validation case 8 results .............................................................. 138 Figure F-4. Simplified containment vessel model showing break locations and key structures, systems, and components .................................................................................... 141 Figure F-5. Cutaway view of the mesh in the center of the model (case 1) ............................ 142 Figure F-6. Detailed view of the mesh around the pipe break (case 1) ................................... 143 Figure F-7. Time history of total forces on key SSC for CNV Case 1 ...................................... 144 Figure F-8. Absolute pressure contours at four time steps for CNV blast Case 1 ................... 145 Figure F-9. Absolute pressure contours for CNV Cases 2 & 3 ................................................ 146 Figure F-10. Modeled region of reactor building........................................................................ 147 Figure F-11. Geometry of part of one pipe gallery in reactor building showing break locations 148 Figure F-12. Identification of components in reactor building .................................................... 149 Figure F-13. Cross-section view and close-up view of the mesh in case 1 ............................... 151 Figure F-14. Pressure contours for three time steps for reactor building blast Case 1 ............. 152 Figure F-15. Force time history for various SSC for reactor building blast Case 1.................... 153

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Abstract The NuScale Power, LLC (NuScale) Pipe Rupture Hazards Analysis (PRHA) describes the methodology applicable to the identification and assessment of pipe rupture hazards, and the effects of pipe ruptures and leakage cracks on the ability to achieve safe shutdown and cooldown.

Specifically, the following are addressed:

compliance with NRC regulations and guidance

identification of postulated rupture locations

characteristics of ruptures, including break types and size

determination of potential effects of high- and moderate-energy line breaks

criteria for showing the acceptability of structures, systems, and components exposed to those effects

mitigation strategies to accommodate pipe rupture hazards, where applicable The evaluation addresses external effects of high-energy line breaks, moderate-energy line breaks, and leakage cracks in piping in the NuScale Power Module (NPM) and NuScale Reactor Building (RXB). The Pipe Rupture Hazards Analysis (PRHA) evaluation of the piping beyond the NPM disconnect flange in the RXB and through the balance of plant is the responsibility of the COL applicant.

The PRHA is required to support the Design Certification per NRC Standard Review Plan Branch Technical Position 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment.

The PRHA can be summarized in the Design Certification, Part 2, Tier 2 Final Safety Analysis Report Section 3.6, Evaluation of Postulated Rupture of Piping, or submitted as a separate technical report.

This NuScale technical report provides the PRHA results for high- and moderate-energy piping ruptures in the NPM and RXB.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Executive Summary The NuScale Power, LLC (NuScale) Pipe Rupture Hazards Analysis (PRHA) methodology evaluates the postulated rupture of high- and moderate-energy piping systems and their effects on the surrounding environment. The design approach demonstrates that postulated piping ruptures in fluid systems do not cause loss of function of essential systems and that the NuScale plant is able to withstand postulated failures of fluid system piping, taking into account the direct results of such a failure and further failure of a single active component, with acceptable consequences.

The NuScale design is a compact, integral reactor that relies on passive safety features to ensure safe shutdown and cooldown for design basis events. The absence of large diameter reactor coolant system piping and active safety systems results in a minimal number of safety-related and essential structures, systems, or components (SSC) susceptible to postulated pipe rupture hazards. Examples of key NuScale design features include:

no operator actions or electrical power are required for safe shutdown and cooldown for design basis accidents.

a limited number of essential SSC outside the NPM itself.

a small-volume, metal containment operated at a vacuum and with sensitive leak-detection capability.

no potential for dislodged piping insulation blocking core cooling.

reduced energy of blast, pipe whip, and jet impingement effects due to smaller plant size and lower energy system conditions.

stainless steel primary and secondary piping within containment and areas where break exclusion is applied.

ready access for inspection.

Application of the criteria for break exclusion and leak-before-break results in a small number of locations in the containment vessel (CNV) and none in the NuScale Power Module (NPM) bay requiring evaluation of the dynamic effects (i.e., blast waves, pipe whip, jet impingement).

Consideration of nonmechanistic breaks of MSS and FWS piping in the containment penetration area involves evaluation of subcompartment pressurization and flooding. Mitigation protection is demonstrated through separation and by the robustness and qualification of safety-related and essential SSC.

For the RXB, evaluation of bounding high-energy line breaks (HELBs) and moderate-energy line breaks (MELBs) was performed to demonstrate that final piping design will be capable of meeting acceptance criteria for evaluation of line breaks.

External effects of HELBs and MELBs in the NuScale Power Plant do not adversely affect the ability to shut down and maintain core cooling of an NPM.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 1.0 Introduction 1.1 Purpose This document describes the NuScale methodology applicable to identification and assessment of pipe rupture hazards and the effects of pipe ruptures and leakage cracks on the ability to achieve safe shutdown and cooldown. Specifically, the following are addressed:

compliance with NRC regulations and guidance

identification of postulated rupture locations

characteristics of ruptures, including break types and sizes

determination of potential effects of high and moderate-energy line breaks (MELBs)

criteria for showing acceptability of structures, systems, and components (SSC) exposed to those effects This evaluation addresses the external effects of high-energy line breaks (HELBs),

MELBs, and leakage cracks in piping in the NuScale Power Module (NPM) and NuScale Reactor Building (RXB). The final Pipe Rupture Hazards Analysis (PRHA) is the responsibility of the Combined License (COL) applicant. COL Item 3.6-3 identifies the requirement to complete the PRHA for lines outside the reactor pool bay:

COL Item 3.6-3: A COL applicant that references the NuScale Power Plant design certification will perform the pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the reactor pool bay and design appropriate protection features. This includes an evaluation and disposition of multi-module impacts in common pipe galleries, and evaluations regarding subcompartment pressurization. The COL applicant will update (FSAR)

Table 3.6-2 as appropriate.

This report addresses the requirements for the as-designed PRHA Report as described in NRC Inspection Procedure 65001.21 (Reference 1.4.2.8):

The as-designed pipe rupture hazards analysis report ITAAC is a set of methodology and criteria pertaining to protection of essential systems or components inside and outside containment from the adverse effects of postulated failures in high and moderate energy piping (HELB and MELB)....

Reference 1.4.2.8 provides for three options:

1. Resolve during the design certification or amendment to the design certification

2. Resolve as part of the combined license (COL) review

3. Resolve after COL is issued For piping located inside the NuScale containment vessel (CNV) and in the NPM pool bay under the bioshield, this report satisfies option 1. For piping elsewhere in the RXB and through the balance of plant, this report establishes the methodology and criteria to be

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 applied and the bounding results to be confirmed for balance of plant arrangements as part of the COL review (option 2). The NuScale Final Safety Analysis Report (FSAR)

(Reference 1.4.2.16) Table 2.1-4 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) item 4, specifies that the as-built Pipe Break Hazard Analysis Report identifies protective features and qualification of safety-related SSC to withstand the dynamic and environmental effects of piping failures.

Detailed evaluations in some areas (e.g., stress analysis break locations, subcompartment pressurization) are currently in progress. Therefore, some report areas are denoted as LATER. Information on these evaluations will be included in the next revision.

1.2 Scope Pipe ruptures are addressed for each of the distinct regions of the NuScale plant where high- or moderate-energy piping layouts exist.

inside the CNV

outside the CNV (under the bioshield)

in the RXB (outside the bioshield)

in the Control Building (CRB)

in the Radioactive Waste Building (RWB)

onsite (outside the RXB, CRB, and RWB buildings)

Although the final pipe routing and analysis in the RXB beyond the NPM bay is the responsibility of the COL applicant, generic break postulations and mitigation evaluations are evaluated in this report. There are no high-energy systems in the CRB or RWB, or onsite (outside the buildings. There are however, moderate-energy systems.

Pipe ruptures in other areas onsite, outside the RXB, control building, and radioactive waste building (i.e., turbine buildings), are not within the scope of this report. The final PRHA is the responsibility of the COL applicant and will address postulated ruptures based on final design of high- and moderate-energy systems in the RXB and moderate-energy systems in the control building, radioactive waste building, and onsite (outside the buildings).

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 1.3 Abbreviations and Definitions Table 1-1. Abbreviations Term Definition ACRS Advisory Committee on Reactor Safeguards ANS American Nuclear Society ANSI American National Standards Institute ASME American Society of Mechanical Engineers BTP Branch Technical Position CFD computational fluid dynamics CFR Code of Federal Regulations CIV containment isolation valve CNTS containment system CNV containment vessel COL combined license CRDM control rod drive mechanism CRDS control rod drive system CVCS chemical and volume control system DAC design acceptance criteria DHRS decay heat removal system DLF dynamic load factor DSRS Design-Specific Review Standard ECCS emergency core cooling system EDSS highly reliable DC power system ESF engineered safety feature FSAR Final Safety Analysis Report FWS feedwater system GDC General Design Criteria HELB high-energy line break IAB inadvertent actuation block ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LBB leak-before-break LOCA loss-of-coolant accident LWR light water reactor

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Term Definition M&E mass and energy MELB moderate-energy line break MHS module heatup system MPS module protection system MS main steam MSS main steam system NMS neutron monitoring system NPM NuScale Power Module NPS nominal pipe size PAM post-accident monitoring 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 RXB reactor building SBO station blackout SG steam generator SRP Standard Review Plan (NUREG-0800)

SSC structures, systems, and components UHS ultimate heat sink ZOI zone of influence

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 1-2. Definitions Term Definition Active component Any component that must perform a mechanical motion or change of state during the course of accomplishing a primary safety function.

Blast wave A shock wave; viz. a high-pressure, high-density region that initiates due to the rapid opening of a pipe rupture and propagates away from the rupture location at supersonic speed.

Class 1E Safety classification of the electric equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or are otherwise essential in preventing a significant release of radioactive material to the environment. The NuScale design does not have Class 1E electrical power.

Essential systems As defined by BTP 3-4 (Reference 1.4.2.5), those systems necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without off-site power.

External effect Any consequence of a high- or moderate-energy line break or leakage crack affecting SSC outside the leaking system. External effects include both dynamic (i.e., pipe whip) and environmental (i.e., increased ambient pressure) effects.

High-energy fluid Fluid systems that, during normal plant conditions, have either or both: (a) system maximum operating temperature exceeding 200°F, (b) maximum operating pressure exceeding 275 psig.

Integral reactor A design with the entire RCS circulation path contained within a single pressure vessel (i.e., there is no loop piping).

Jet impingement force The force imparted to an object due to its intersection with the fluid issuing from a ruptured pipe. The magnitude of this force depends on such parameters as the thermodynamic conditions of the fluid in the pipe, distance of the pipe rupture from the target, area of intersection of the jet with the target surface, and the shape of the target.

Moderate-energy fluid Fluid systems that, during normal plant conditions, have: (a) maximum system operating temperature is 200°F or less, (b) maximum operating pressure is 275 psig or less, or (c) high-energy conditions that exist less than one percent of the plant life or less than two percent of the time period required for the system to accomplish its function.

NuScale Power The assembly including the reactor pressure vessel, CNV, and Module (NPM) directly-attached components out to the outboard flange connecting module systems to those in the RXB.

Outboard Identifies location of a component as farther outside the CNV boundary, regardless of flow direction inside the component.

Pipe failure hazard An area containing piping normally operating at high or moderate energies.

area Pipe whip Movement of a pipe caused by jet thrust resulting from a pipe failure. Pipe whip is assumed to occur in the plane defined by piping geometry and configuration unless limited by structural members, pipe restraints, or pipe stiffness.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Term Definition Single active failure The failure of an active component to complete its intended function upon demand. Per BTP 3-3 (Reference 1.4.2.4), Appendix A, failure of an active component of a fluid system is loss of component function as a result of mechanical, hydraulic, pneumatic, or electrical malfunction, but not the loss of structural integrity. The direct consequences of a single active failure are evaluated. A single active failure is postulated to occur simultaneously with the pipe failure; passive failures are not postulated.

Single failure criterion As defined in 10 CFR 50 Appendix A, A single-failure means an occurrence that results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single-failure. Fluid and electric systems are considered to be designed against an assumed single-failure if neither (1) a single-failure of any active component (assuming passive components function properly) nor (2) a single-failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions.

This definition is accompanied by a footnote: Single failures of passive components in electric systems should be assumed in designing against a single failure. The conditions under which a single failure of a passive component in a fluid system should be considered in designing the system against a single failure are under development.

Subcompartment A fully- or partially-enclosed volume within the NuScale plant that houses or adjoins piping systems and restricts the flow of fluid to other areas of the plant in the event of a postulated pipe rupture.

Terminal end The extremity of a piping run that connects to structures, components (e.g.,

vessels, pumps, valves), or pipe anchors that act as rigid constraints to piping motion and thermal expansion. A branch connection on a main piping run is a terminal end for the branch run, except where the branch run is classified as part of a main run in the stress analysis or is shown to have a significant effect on the main run behavior. In piping runs that are maintained pressurized during normal plant conditions for a portion of the run (i.e., up to the first normally closed valve), a terminal end of such a run is the piping connection to this closed valve.

Zone of influence (ZOI) The maximum physical range of the direct effects of pipe whip, jet impingement, and the environmental effects resulting from a pipe rupture. The size of ZOI depends on the direct effect being evaluated (e.g., within physical reach of a whipping pipe of a given length or entire compartment for pressurization).

1.4 References 1.4.1 Source Documents 1.4.1.1 American Society of Mechanical Engineers, Quality Assurance Requirements for Nuclear Facility Applications, NQA-1-2008, NQA-1a-2009 Addenda, New York, NY.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 1.4.1.2 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 Appendix B).

1.4.2 Referenced Documents 1.4.2.1 U.S. Code of Federal Regulations, General Design Criteria for Nuclear Power Plants, Appendix A, Part 50, Chapter 1, Title 10, Energy, (10 CFR 50 Appendix A).

1.4.2.2 U.S. Nuclear Regulatory Commission, Standard Review Plan, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, NUREG-0800, Chapter 3, Section 3.6.2, Rev. 2, March 2007, or Rev. 3, December 2016 (as noted).

1.4.2.3 U.S. Nuclear Regulatory Commission, Standard Review Plan, Leak-before-break Evaluation Procedures, NUREG-0800, Chapter 3, Section 3.6.3, Rev. 1, March 2007.

1.4.2.4 U.S. Nuclear Regulatory Commission, Standard Review Plan, Protection against Postulated Piping Failures in Fluid Systems outside Containment, NUREG-0800, Chapter 3, BTP 3-3, Rev. 3, March 2007.

1.4.2.5 U.S. Nuclear Regulatory Commission, Standard Review Plan, Postulated Rupture Locations in Fluid System Piping inside and outside Containment, NUREG-0800, Chapter 3, BTP 3-4, Rev. 2, March 2007.

1.4.2.6 U.S. Nuclear Regulatory Commission, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Regulatory Guide 1.97, Rev. 4, June 2006 (June 2013).

1.4.2.7 U.S. Nuclear Regulatory Commission, NRC Inspection Manual, Definition of Leak-Before-Break Analysis and its Application to Plant Piping Systems, EMCB Part 9900:

10 CFR Guidance, LBBGUIDE.CFR.

1.4.2.8 U.S. Nuclear Regulatory Commission, Inspection of Pipe Rupture Hazards Analyses (Inside and Outside Containment) Design Acceptance Criteria (DAC)-Related ITAAC, Inspection Procedure (IP) 65001.21, November 7, 2011.

1.4.2.9 U.S. Nuclear Regulatory Commission, Two Phase Jet Loads, NUREG/CR-2913, January 1983.

1.4.2.10 U.S. Nuclear Regulatory Commission, Boiling Water Reactor ECCS Suction Strainer Performance Issue No. 7 - ZOI Adjustment for Air Jet Testing, BWROG Meeting, July 20, 2011, Agencywide Document Access and Management System (ADAMS)

Accession No. ML11203A432.

1.4.2.11 U.S. Nuclear Regulatory Commission, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee - Evaluation of Potential for Pipe Breaks, NUREG-1061, Vol. 3,November 1984.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 1.4.2.12 U.S. Nuclear Regulatory Commission, GSI-191 SER, Appendix I, ANSI/ANS Jet Model.

1.4.2.13 Corradini, Michael, Advisory Committee on Reactor Safeguards, letter to Victor McCree, U.S. Nuclear Regulatory Commission, April 12, 2018, ADAMS Access No.

ML18102A074.

1.4.2.14 U.S. Nuclear Regulatory Commission, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, Rev. 1, March 2007 (R April 2005).

1.4.2.15 Advisory Committee on Reactor Safeguards, Transcript: U.S. EPR Subcommittee Meeting, February 21, 2012, ADAMS Accession No. ML120760106 1.4.2.16 NuScale Final Safety Analysis Report, NuScale Standard Plant Design Certification Application, Rev. 1.

1.4.3 Other References 1.4.3.1 American Nuclear Society, Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, ANSI/ANS-58.2-1988, LaGrange Park, IL. (Withdrawn 1998). (Note: Although withdrawn, 58.2 is still referenced by NRC documentation.)

1.4.3.2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition,Section III, Division 1, Rules for Construction of Nuclear Facility Components, New York, NY.

1.4.3.3 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition,Section XI Division 1, Rules for In-service Inspection of Nuclear Power Plant Components, New York, NY.

1.4.3.4 American Society of Mechanical Engineers, Power Piping, B31.1, 2013, New York, NY.

1.4.3.5 Westinghouse Electric Corporation, AP1000 Design Control Document, Rev. 19, June 21, 2011.

1.4.3.6 Kinney, G.F. and K.J. Graham, Explosive Shocks in Air, 2nd Edition, 1985.

1.4.3.7 Liu, J., et al., Investigation on Energetics of Ex-vessel Vapor Explosion Based on Spontaneous Nucleation Fragmentation, Journal of Nuclear Science and Technology, (2002): 39:1, pp 31-39.

1.4.3.8 Ho, C.M. and N.S. Nosseir, Dynamics of an Impinging Jet, Part I, The Feedback Phenomenon, Journal of Fluid Mechanics, 1981: 105:119-142.

1.4.3.9 Nuclear Energy Agency, Knowledge Base for Emergency Core Cooling System Recirculation Reliability, NEA Report NEA/CSNI/R (95)11, February 1996.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 1.4.3.10 Shapiro, Ascher H., The Dynamics and Thermodynamics of Compressible Fluid Flow, The Ronald Press, New York, 1953.

1.4.3.11 Indiana and Michigan Power, D.C. Cook Nuclear Plant Updated Final Safety Analysis Report, Chapter 1, Rev. 27.0, ADAMS Accession No. ML16336A246.

1.4.3.12 McBurnett, Mark, South Texas Project Units 3 & 4, letter to Document Control Desk, U.S. Nuclear Regulatory Commission, October 14, 2010, ADAMS Accession No. ML102910232.

1.4.3.13 Karlos, Vasilis. and George Solomos, Calculation of Blast Loads for Application to Structural Components, European Commission Joint Research Center, EUR 26456 EN, 2013.

1.4.3.14 U.S. Nuclear Regulatory Commission, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, NUREG/CR-6808, February 2003.

1.4.3.15 Department of Defense, Structures to Resist the Effects of Accidental Explosions, with Change 2 UFC 3-340-02, December 2008.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 2.0 Background Design requirements for piping, such as the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Reference 1.4.3.3), ensure that the probability of a pipe rupture is low. However, breaks are postulated conservatively to occur in various plant locations. To ensure protection from postulated breaks, the NRC has provided guidance and criteria for postulating break locations and assumptions to be used in assessing the break consequences.

The consequences of line breaks depend on the thermodynamic conditions in the piping system at the break location and interaction with the break surroundings. Failure of high- and moderate-energy systems have the potential for a range of external effects, as identified in SRP 3.6.2:

pipe whip impact*

blast effects*

jet reaction loads*

jet impingement*

potential feedback amplification and resonance effects of jet impingement*

subcompartment pressurization

flooding (flooding analysis is addressed in Section 3.4 of the FSAR)

Applicable to high-energy systems only. Moderate-energy line breaks are considered only for moderate-energy systems not designed to seismic category requirements.

When a pipe rupture initiates, a sudden release of energy and fluid occurs. If the system upstream conditions support continued blowdown (i.e., there is a large amount of fluid),

the discharge from the break cannot expand until exiting the pipe. In this scenario, the jet that forms is under-expanded. This means that the jet expands and accelerates once released from the confines of the pipe. Its speed, expansion, and thermodynamic conditions are a function of the fluid conditions in the pipe and the conditions of the ambient environment.

NuScale ambient pressure inside the CNV is maintained at less than 1 psia.

2.1 NuScale Design Features Relevant to Pipe Rupture Hazards Analysis The NuScale design is an integral, multi-unit, small modular reactor for which safety is provided by passive features without the need for safety-related electrical power. Because NRC regulatory guidance is premised on the existing fleet of large light water reactors (LWRs) with reactor coolant loops and active safety features, instances exist where the current NRC pipe rupture guidance is not a direct fit. In many cases, the NRC has not issued a Design-Specific Review Standard (DSRS) to address what is directly applicable for the NuScale design. Specific examples of NuScale relevant design differences are:

The response to pipe ruptures requires neither electric power nor injection of additional cooling water.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

The NPMs are mostly submerged in a large pool of water that serves as the ultimate heat sink for the NPMs and does not require replenishment.

Design basis accidents do not require operator actions or the re-establishment of electric power for long-term cooling.

Piping is small compared to the large reactors for which regulatory guidance was initially developed.

Active, safety-related components [i.e., emergency core cooling system (ECCS) valves, decay heat removal system (DHRS) actuation valves, and containment isolation valves (CIVs)] are shown to operate during refueling. As part of the start-up sequence for an NPM, each of the safety-related ECCS, DHRS, and CIVs is repositioned. These periodic system line-up activities ensure that the safety-related valves remain operable.

The NPM containment is a pressure vessel designed and fabricated to ASME Code Section III Class 1 requirements versus a building in conventional LWRs.

Piping of the NPM, including secondary system piping, is made of corrosion-resistant stainless steel.

Main steam system (MSS) and feedwater system (FWS) piping inside the containment boundary is designed to RCS design pressure and temperature.

MSS and FWS piping inside the CNV meets LBB criteria.

HELBs inside the CNV are limited to NPS 2 piping.

The length of piping in which breaks must be postulated is minimal and the size of high-energy piping is small compared to current design reactors.

The NPM containment is operated at a vacuum.

Equipment and piping inside the NPM containment are not covered with insulation, which is important for multiple reasons:

o Jet impingement does not dislodge piping insulation that could lead to blockage of long-term cooling recirculation.

o Detection of small leakage cracks is not impeded by retention of moisture in insulation.

o The bare piping is readily inspectable during refueling, because insulation does not need to be removed to note deposits, discoloration, or other signs of degradation.

o Corrosive substances (e.g., chlorides) cannot be trapped and held in contact with the piping surface.

Safety-related and essential components inside the NPM containment are qualified to be functional after exposure to saturated steam at containment design pressure up to 1000 psia, requiring designs that are robust.

The small containment results in congestion that makes difficult the addition of traditional pipe whip restraints and the physical separation of essential components

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 from break locations, but whipping pipes in turn, have a limited range of motion before encountering an obstacle.

The CIVs are installed outside of containment. Where two valves in series are required (i.e., for containment penetrations governed by GDC 55 and 56), both are in a single-piece valve body (i.e., no piping or welds between the CIVs, which precludes breaks in between). Also, the lines directly connected to the primary system or the containment have only a single weld in the area between the containment wall and the CIV.

Containment pressure suppression is not required and there are no sprays that introduce chemical additives.

During refueling, the NPM is disconnected from supporting systems by removal of piping spools, transported by crane to a refueling location, and disassembled. This provides access for inspection to portions of the plant not normally accessible.

Up to 12 NPMs are operating at the same time and in proximity, so the potential for a rupture in a system of one module to affect others is considered.

The plant main control room is in a separate building that does not contain high-energy piping systems.

Effects of postulated ruptures on multiple modules are evaluated, and protection for post-accident monitoring (PAM) capability and reliable DC power is provided by separating them in different compartments within the RXB.

In the NuScale design, postulated HELBs are evaluated in three discrete areas of the plant because of differences of both the potential piping hazard and the surrounding environment:1

1. inside the containment of the NPM

2. in the pool bay area above each NPM and under the bioshield

3. in the RXB The NuScale methodology for evaluating pipe ruptures across the three plant areas accounts for the break hazard and the surrounding environment for that break area. The design approach demonstrates that postulated ruptures in fluid systems do not cause loss of function of essential systems and that the NuScale plant is able to withstand postulated failures of any fluid system piping, taking into account the direct results of such a failure and subsequent failure of a single active component, with acceptable consequences.

1 Moderate-energy systems are not in use in areas 1 or 2, with the exception of the reactor component cooling water system (RCCWS) lines for the control rod drive mechanisms (CRDMs), the rupture effects of which are bounded by HELBs. The containment flooding and drain system is moderate energy when in use, but is isolated whenever the reactor is operating. Evaluation of line breaks in other plant areas (i.e., turbine buildings) is outside the scope of this report.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Long-term core cooling for the NuScale plant following an HELB (or MELB) is, therefore, dependent only upon the following:

The safe state for safety-related actuation devices is defined as de-energized. With no power, the reactor trip and engineered safety feature (ESF) components go to their safe state. The module protection system (MPS) trip signal equipment interface module outputs de-energize to actuate reactor trip and ESF components. Reactor trip occurs upon an automatic or manual MPS trip signal, loss of two or more MPS channels, or loss of electric power.

Isolation of the CNV by shutting CIVs, for which no AC or DC electric power is required.

Opening ECCS valves, for which no AC or DC electric power is required.

Opening DHRS actuation valves, for which no AC or DC electric power is required.

As such, pipe ruptures have limited potential to adversely affect essential functions.

Although the mitigation objective is consistent with that of other large LWRs, as described in the NRC guidance, the unique features and passive safety attributes of the NuScale design justify other considerations as part of this PRHA and mitigation, as described in following sections.

2.2 Regulations and Guidance In this section, regulatory criteria relevant to line breaks are summarized, followed by a brief discussion of the NuScale approach. In subsequent sections, each of the aspects of assessing line breaks is discussed, along with their likelihood of occurring, their applicability to the NuScale design, the detailed methodology for addressing them, and the results of evaluations.

Environmental and Dynamic Effects Design Bases, General Design Criterion (GDC) 4, of Appendix A to 10 CFR 50, (Reference 1.4.2.1) requires SSC important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

2.2.1 History of HELB Effects Analysis Methodology The nuclear industry has traditionally used ANSI/ANS Standard 58.2-1988 (Reference 1.4.3.1) for estimating jet plume geometries and impingement loads, based on fluid conditions internal and external to the piping, even though it was officially withdrawn in 1988. In 2004, following interactions with the Advisory Committee on Reactor Safeguards (ACRS) on the jet models described in ANSI/ANS 58.2, the staff determined that there were potential nonconservatisms in these models with respect to the strength, Zone of Influence (ZOI), and space and time-varying nature of the loading effects of postulated pipe ruptures on neighboring SSC.

In the time since, the NRC has not developed detailed acceptance criteria for these complex phenomena. The NRC concerns are based on experimental data and are intended to ensure that modeling of HELB effects includes margin to compensate for

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 possible deficiencies in the available data and its interpretation. The lack of acceptance criteria founded on physical phenomena has led to iterative interactions between the industry and the NRC regarding acceptability not just of results, but also of methodologies.

The approach discussed herein considers NRC guidance, past precedent, and NuScale specific design features.

2.2.2 NRC Guidance The following is a description of the relevant guidance, but is not intended to be all-inclusive or to imply that the NuScale implementation of the guidance does not address criteria that are not explicitly discussed. Also, note that this guidance was developed on the premise of application to LWRs with large containments and active safety systems dependent on electrical power availability.

2.2.2.1 Branch Technical Position 3-4 Branch Technical Position 3-42 notes that the NRC staffs observation of actual piping failures has indicated that they generally occur at high stress and fatigue locations, such as at the terminal ends of a piping system and at its connection to the nozzles of a component. The BTP 3-4 criteria use the available piping design information by postulating pipe ruptures at locations having higher potential for failure, such that an adequate and practical level of protection is achieved. Branch Technical Position 3-4 also points out that, subject to certain limitations, GDC 4 allows that dynamic effects associated with postulated pipe ruptures be excluded from the design basis when analyses reviewed and approved by the NRC demonstrate that the probability of fluid system piping rupture is extremely low under design-basis conditions. An example of these analyses is LBB analyses, which are reviewed in accordance with SRP Section 3.6.3.

2.2.2.1.1 High-Energy Breaks For the purpose of assessing separation from high-energy piping, the effects of postulated piping breaks should be physically remote from essential systems and components. A footnote defines essential systems and components as those necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without off-site power.

Containment penetration areas For piping in containment penetration areas, breaks and cracks need not be postulated (break exclusion) in sections between the containment wall to and including the inboard or outboard CIVs, provided:

They have at least a specified margin to design criteria of the ASME Boiler and Pressure Vessel Code (Code),Section III, and do not exceed fatigue limits specific to the piping code class. For example, at a high level, Class 1 piping should not exceed 2 SRP 3.6.2 states that BTP 3-4 should be used to determine rupture locations both inside and outside containment.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 a maximum stress range between any two load sets (including the zero load set) of 2.4 Sm and should be calculated using Equation (10) of ASME Code Section III Subarticle NB-3653, a cumulative usage factor of 0.1, and a maximum stress of 2.25 Sm and 1.8 Sy. Class 2 provisions are similar.

The length of such piping should be as short as practical.

The number of fittings and welds are minimized.

Welded restraints are avoided.

Welds in the zone are 100 percent volumetrically inspected every interval.

Other than containment penetration areas For Class 1 piping not in containment penetration areas, breaks should be postulated at:

terminal ends

intermediate locations where the maximum stress range exceeds 2.4 Sm

intermediate locations where the cumulative usage factor exceeds 0.1 (0.4 if environmental fatigue considered)

Branch Technical Position 3-4 acknowledges that reanalysis may cause the highest stress locations to shift, but allows the initially determined intermediate break locations be retained, provided the mitigation by original pipe whip restraints and jet shields is still satisfactory and the pipe size, wall thickness, and routing remain similar.

For Class 2 and 3 piping, postulated breaks should be assumed at:

terminal ends

intermediate locations selected by one of the following criteria:

o Each fitting or at the extreme ends of a piping run if there are no fittings o Locations where stresses are calculated to exceed 0.8 times the sum of the limits in Subarticles NC/ND-3653 (/ indicates the applicable article should be used).

Breaks in seismically analyzed non-ASME class piping are postulated according to the same criteria as for ASME Class 2 and 3 piping.

If a structure separates a high-energy line from an essential component, that separating structure should be designed to withstand the consequences of a pipe rupture in the high-energy line that produces the greatest effect at the structure, irrespective of the fact that the above criteria might not need such a break location to be postulated.

Safety-related equipment is environmentally qualified in accordance with SRP Section 3.11. Appropriate pipe ruptures and leakage cracks (whichever controls) should be included in the design bases for environmental qualification of electrical and mechanical equipment both inside and outside containment.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 With the exception of locations meeting the break exclusion criteria for piping in containment penetration areas, leakage cracks in piping should be postulated for

Class 1 piping where the calculated stress range per Subarticle NB-3653 exceeds 1.2 Sm at axial locations.

Class 2, 3, and nonsafety piping where calculated stress exceeds 0.4 times the sum of the stress limits of Subarticles NC/ND-3653.

Nonsafety-class piping not evaluated to obtain stress information at axial locations that produce the most severe environmental effects.

NuScale follows the guidance of BTP 3-4 with exceptions as described below.

Inside containment -

o The CIVs are outside the containment. A break inside the CNV does not lead to containment bypass. Therefore, there is no containment penetration area inside the CNV, and BTP 3-4 Rev. 2 Paragraph B.A.(ii) does not apply.

o The congestion associated with an integrated small modular reactor design limits the ability to separate fluid systems from essential SSC. Those SSC (e.g., ECCS valves, instrumentation) are designed to function in the severe environment resulting from ECCS initiation, which bounds the HELB effects.

o The RCS-connected and DHRS piping inside containment, all of which are NPS 2, are assessed for compliance with BTP 3-4 Rev. 2 B.A.(iii):

Breaks are postulated at terminal ends, where the piping attaches to RPV or CNV nozzles.

Breaks at intermediate locations are precluded by designing for compliance with BTP 3-4 Rev. 2 Paragraph B.A.(iii)(1)(b) and (c). If those criteria are not met at certain locations, then breaks are considered.

o Large-diameter secondary piping (i.e., MSS and FWS) is analyzed for LBB and shown to meet the criteria, as described below in the Section 2.2.5 discussion of SRP 3.6.3.

o Postulated break locations are assessed for the effects of a break.

NuScale Power Module bay under the bioshield -

o The containment penetration area is defined as the segment from the CNV nozzle to and including the weld connecting pipe to the outboard nozzle of the CIV or check valve in a line. The only physical piping in the containment penetration area is the DHRS. The design of piping and valves, nozzles, and fittings within the containment penetration area complies with break exclusion criteria of BTP 3-4 Rev. 2 Paragraph B.A.(ii):

Stress and cumulative usage factor criteria are met.

Welded attachments and restraints are not used.

The number of welds and length of piping is minimized.

Guard pipes are not used.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 A 100-percent volumetric and surface examination is performed for the NPS 2 piping in addition to larger piping.

o Remaining piping, including the refueling pipe spools, comply with BTP 3-4 Rev. 2 Paragraph B.A.(iii).

o Based on the above BTP 3-4, no breaks in the NPM bay outside the CNV (under the bioshield) are postulated. However, leakage cracks are considered.

Reactor Building - The effects on the RXB of a rupture of high-energy piping in any location are bounded and shown to be acceptable for HELB effects.

2.2.2.1.2 Moderate-Energy Piping Separation adequate to isolate the effects of through-wall cracks should be provided from essential systems and components. However, leakage cracks need not be considered in the containment penetration area provided 1) they meet the criteria of the ASME Code,Section III, NE-1120, and 2) the stresses calculated by the sum of Eqs. (9) and (10) in ASME Code,Section III, NC-3653 do not exceed 0.4 times the sum of the stress limits given in NC-3653.

Outside the containment penetration areas, leakage cracks should be considered where stress criteria are not met, except where exempted. Leakage cracks in moderate-energy piping do not need to be evaluated if environmental conditions resulting from an HELB in the vicinity are more limiting.

NuScale Approach: Consequences of leakage cracks in piping in the moderate-energy piping systems are either bounded by HELB effects or are evaluated and shown to be acceptable.

2.2.2.2 Types of Breaks and Cracks Circumferential breaks should be postulated in high-energy piping as follows:

For piping over NPS 1, except where specific stress criteria are met.

For unanalyzed piping, at each weld to a fitting, valve, or attachment.

Pipe separation of at least one-diameter laterally occurs unless physically constrained.

Dynamic force of jet discharge should be based on effective flow area of pipe and fluid pressure modified by a thrust coefficient. Limited pipe displacement at the break location, line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge.

Pipe whip should be assumed to occur in the plane defined by the piping geometry and configuration and to initiate pipe movement in the direction of the jet reaction.

Longitudinal breaks should be postulated in high-energy piping as follows:

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

For piping NPS 4 and larger where circumferential breaks are considered, except where specific stress criteria are met and excluding terminal ends.

The opening is an axial split without severance, oriented at either of two diametrically opposed points that result in out of plane bending (or in the highest tensile stress location).

Dynamic force is based on a circular (or elliptical 2D by 1/2D) area equal to the effective pipe cross-sectional area, modified by a thrust coefficient and considering line restrictions, flow limiters, and etc.

Piping movement occurs in opposite direction of the jet reaction unless physically constrained.

Leakage cracks should be postulated, as follows:

In piping larger than NPS 1.

At circumferential (and axial for moderate-energy systems) locations resulting in the most severe environmental consequences.

Need not be postulated in moderate-energy piping in an area where a HELB is postulated, provided such leakage cracks would not result in more limiting environmental conditions than the HELB.

Flow is based on circular opening of area equal to that of a rectangle of one half of the diameter by one half of the wall thickness.

Flow wets unprotected components within the compartment and communicating compartments.

NuScale Approach: NuScale follows this guidance, including as appropriate for piping shown to meet break exclusion (i.e., no breaks or cracks) or LBB (no breaks assumed for purposes of assessing dynamic effects), where essential systems and components could be affected.

2.2.3 Branch Technical Position 3-3 Branch Technical Position 3-3, Rev. 3, describes the approaches acceptable for the design, including the arrangement, of fluid systems located outside of containment to ensure that the plant can be safely shut down in the event of piping failures outside containment. The intent is to show that postulated piping failures combined with the failure of any single active component do not cause the loss of function of essential systems. The BTP is also intended to provide clear guidance on acceptable means of protecting against MSS and FWS breaks.

The BTP identifies that protection of essential systems and components against postulated piping failures in high- or moderate-energy fluid systems that operate during normal plant conditions and that are located outside of containment should be provided by (in order of preference):

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Plant arrangements that separate fluid system piping from essential systems and components. Separation should be achieved by plant physical layouts that provide sufficient distances between essential systems and components and fluid system piping, such that the environmental effects of any postulated piping failure therein cannot impair the integrity or operability of essential systems and components.

Separation of MSS and FWS lines should be implemented even if the criteria of BTP 3-4 are met for break exclusion. A nonmechanistic longitudinal break should have a cross sectional area of at least one square foot 3 and be postulated to occur at a location that has the greatest effect on essential equipment.

Fluid system piping or portions thereof not satisfying the separation provisions above should be enclosed within structures or compartments designed to protect nearby essential systems and components. Alternatively, essential systems and components may be enclosed within structures or compartments designed to withstand the effects of postulated piping failures in nearby fluid systems.

If the above cannot be satisfied, then redundant design features that are separated or otherwise protected from postulated piping failures, or additional protection, should be provided so that the effects of postulated piping failures are shown to be acceptable.

Additional protection may be provided by designing or testing essential systems and components to withstand the environmental effects associated with postulated piping failures.

The BTP states that protective structures should be designed to withstand the effects of a postulated piping failure (i.e., pipe whip, jet impingement, pressurization of compartments, water spray, and flooding, as appropriate) in combination with loadings associated with the design basis earthquake, within the respective design load limits for structures. Fluid system piping in containment penetration areas should be designed to meet the break exclusion provisions of BTP 3-4.

Piping failures should be postulated in accordance with BTP 3-4 and include full circumferential ruptures of non-seismic moderate-energy piping (because BTP 3-4 only applies during normal conditions, not seismic events). Each longitudinal or circumferential break or leakage crack should be considered separately as a single postulated initial event occurring during normal plant conditions. An analysis should be made of the effects of each such event, taking into account the provisions BTP 3-4 and of the system and component operability considerations.

In analyzing the effects of postulated piping failures, the following assumptions should be made with regard to the operability of systems and components:

3 The area of 1.0 ft2 is based on the assumption that the piping in which the longitudinal rupture occurs has a diameter such that the flow area out of a complete circumferential break is at least as large. For the largest high-energy piping in the NuScale plant, which is NPS 12 in the MSS, the circumferential break area is only 0.63 ft2 (91 in.2). See Section 3.5.2.5.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

1. Off-site power should be assumed to be unavailable if a trip of the turbine-generator system or reactor protection system is a direct consequence of the postulated piping failure. Also, off-site power should be assumed unavailable following seismic events.

2. A single active component failure should be assumed in systems used to mitigate consequences of the postulated piping failure and to shut down the reactor, except as noted in item 3 below. The single active component failure is in addition to the postulated piping failure and any direct consequences of the piping failure.

3. Where the postulated piping failure is assumed to occur in one of two or more redundant trains of a dual-purpose moderate-energy essential system (i.e., one required to operate during normal plant conditions as well as to shut down the reactor and mitigate consequences of postulated piping failure), single active failures of components in the other train or trains of that system or other systems necessary to mitigate the consequences of the piping failure and shut down the reactor, need not be assumed provided the systems are designed to seismic Category I standards, are powered from both off-site and on-site sources, and are constructed, operated, and inspected to quality assurance, testing, and in-service inspection standards appropriate for nuclear safety systems. Examples of systems that may, in some plant designs, qualify as dual-purpose essential systems are service water systems, component cooling systems, and residual heat removal systems.

4. Available systems, including those actuated by operator actions, may be employed to mitigate the consequences of a postulated piping failure. In judging the availability of systems, account should be taken of the postulated failure and its direct consequences such as unit trip and loss of off-site power, and of the assumed single active component failure and its direct consequences. The feasibility of carrying out operator actions should be judged on the basis of ample time and adequate access to equipment being available for the proposed actions. For breaks in non-seismic piping systems, only seismically-qualified systems should be assumed to be available to mitigate the consequences of the failure because a seismic event may have caused the pipe break.

Environmental effects of a postulated piping failure should not preclude habitability of the control room or access to areas important to safe control of reactor operation needed to cope with the consequences of the piping failure. The functional capability of essential systems and components should be maintained after a failure of piping not designed to seismic Category I standards, assuming a concurrent single active failure.

The considerations related to the LBB approach should conform with the provisions of SRP Section 3.6.3.

NuScale Approach: LBB is applied inside the CNV to steam generator system large-bore (i.e., MSS and FWS) piping. The containment penetration area extends from the CNV nozzle weld to the outermost nozzle weld of a CIV or check valve in a line.

Moderate-energy piping leakage cracks are bounded by HELBs in the CNV, are non-limiting in the NPM bay outside the CNV (under the bioshield), and are bounded by HELBs in other parts of the RXB. Where breaks can occur, the impact of external effects such as pipe whip is evaluated. NuScale conforms to the provisions of regulatory guidance, but has slightly extended applicability of the containment penetration area.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 2.2.4 Standard Review Plan Section 3.6.2 NuScale is working to SRP 3.6.2, Rev. 2, based upon approved guidance existing six months before submittal of the design certification application (DCA). Paragraph III.3 identifies a concern with the guidance in ANSI/ANS Standard 58.2-1988 and notes that reviews of the technical adequacy of jet modeling are being done on a case-by-case basis while the NRC assesses the issue. Revision 3 of SRP 3.6.2, dated December 2016, still does not contain definitive guidance to resolve the concerns with ANSI/ANS Standard 58.2-1988, and states that alternate standards are not yet available to address the problems identified with ANSI/ANS 58.2. It also states that each new reactor design certification applications dynamic jet load modeling is assessed on a case-by-case basis.

The following sections outline the NRCs position on the three classes of HELB jet impingement effects. In order to address the NRC concerns regarding HELB effects, this report considers the information in SRP 3.6.2 Rev. 3, despite being not applicable to the NuScale DCA.

Standard Review Plan 3.6.2 states that BTP 3-4 should be used to determine rupture locations both inside and outside containment.

2.2.4.1 Jet Thrust Loads Static Analysis Model: The jet thrust force is represented by an amplified static loading, and the ruptured system is analyzed statically. An amplification factor can be used to establish the magnitude of the forcing function. However, the factor should be based on a conservative value obtained by comparison with factors derived from detailed dynamic analyses performed on comparable systems.

Dynamic Analysis Models:

1. The time-dependent function representing the thrust force caused by jet flow from a postulated pipe break or crack should include the combined effects of the following:

the thrust pulse resulting from the sudden pressure drop at the initial moment of pipe rupture; the thrust transient resulting from wave propagation and reflection; and the blowdown thrust resulting from buildup of the discharge flow rate, which may reach steady state if there is a fluid energy reservoir having sufficient capacity to develop a steady jet for a significant interval. Alternatively, a steady state jet thrust function may be used.

2. A rise time not exceeding one millisecond should be used for the initial pulse, unless a combined crack propagation time and break opening time greater than one millisecond can be substantiated by experimental data or analytical theory based on dynamic structural response.

3. The time variation of the jet thrust forcing function should be related to the pressure, enthalpy, and volume of fluid in the upstream reservoir and the capability of the reservoir to supply a high-energy flow stream to the break area for a significant interval.

The shape of the transient function may be modified by considering the break area and the system flow conditions, piping friction losses, flow directional changes, and application of flow-limiting devices.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

4. The jet thrust force may be represented by a steady state function if the energy balance model or the static model is used in the subsequent pipe motion analysis. In either case, a step function amplified as indicated above, is acceptable. The function should have a magnitude not less than T = KPA where P = system pressure before pipe break, A = pipe break area, and K = thrust coefficient4. To be acceptable, K values should not be less than 1.26 for steam, saturated water, or steam-water mixtures, or 2.0 for subcooled, nonflashing water.

NuScale Approach: NuScale has applied the approach noted in item 4 above and uses a steady state function of the form jet thrust force T = KPA. For other analyses, such as subcompartment pressurization, a non-steady discharge based on the characteristics of the upstream reservoir is applied.

2.2.4.2 Jet Plume Expansion and Zone of Influence Although ANSI/ANS Standard 58.2-1998 has been accepted by the NRC, the ACRS (Reference 1.4.2.14) noted the potential for nonconservative assessments of jet impingement loads of postulated pipe breaks on neighboring SSC. The NRC staff has been assessing the technical adequacy of information pertaining to dynamic analyses models for jet thrust force and jet impingement load. Pending completion of this effort, the NRC staff reviews analyses of jet impingement forces on a case by case basis. These analyses should show that jet impingement loadings on nearby safety-related SSC do not impair or preclude their essential functions. The assumptions are as follows:

1. The jet area expands uniformly at a half angle, not exceeding 10 degrees.

2. The impinging jet proceeds along a straight path.

3. The total impingement force acting on any cross-sectional area of the jet is time and distance invariant, with a total magnitude equivalent to the jet thrust force as defined above.

4. The impingement force is uniformly distributed across the cross-sectional area of the jet, and only the portion intercepted by the target is considered.

5. The break opening may be assumed to be a circular orifice of cross-sectional flow area equal to the effective flow area of the break.

6. Jet expansion within a zone of five pipe diameters from the break location is acceptable if substantiated by a valid analysis or testing (i.e., Moody's expansion model). However, jet expansion is applicable to steam or water-steam mixtures only and should not be applied to cases of saturated water or subcooled water blowdown.

As a result of the ACRS questions, the NRC concluded that some physically incorrect assumptions form the basis for portions of the ANSI/ANS 58.2 methodology.

4 Elsewhere in this report, the thrust coefficient symbol is C , consistent with more common usage.

T

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 The standard assumes that it is conservative to model a jet issuing from an HELB as expanding at a constant 45-degree half-angle out to the asymptotic plane and then at a 10-degree half-angle. The asymptotic plane is described as the point at which the jet begins to interact with the surrounding environment, or the point where the jet conditions (e.g., pressure, temperature, flow rate) at the break mix with the surrounding environment.

In particular, expansion angle at given distance downstream depends on the relative conditions at the periphery of the jet in relation to the ambient conditions.

Supersonic jet behavior can persist over distances from the break that are longer than those estimated by the standard, extending the ZOI of the jet and the number of SSC that could be impacted by a supersonic jet.

NuScale approach: NuScale has applied assumptions 2 through 5 above. Blowdown from HELBs inside the NuScale CNV differs because of the initially lower surrounding air pressure. The NuScale approach for jet expansion from postulated breaks inside the CNV therefore differs, as described in Section 3.9.5.2 and Appendix E. Jet expansion for postulated HELBs beyond the CNV in the RXB bounds the methodology accepted by the NRC, as also as described in Section 3.9.5.2 and Appendix E.

2.2.4.3 Distribution of Pressure within the Jet Plume Appendices C and D of ANSI/ANS Standard 58.2-1988 discuss how to determine spatial pressure distribution across different types of jets. For an expanding jet, the Standard assumes variable (not uniform) pressure with a maximum at the jet centerline. The NRC considers that, while this is reasonable in the vicinity of the break, the pressure profile can vary farther away, with peaking near the jets outer envelope. To ensure impingement loading is conservatively calculated, applicants must justify the pressure distribution as a function of downstream and radial distance.

NuScale approach: Inside the CNV, where jet expansion is constrained only by momentum, impingement effects are ((2(a),(c) See Section 3.9.7 and Appendix E for detailed discussion. 2.2.4.4 Dynamic Loading and Potential Amplification due to Fluid-Structure Interaction The NRC has identified that unsteadiness in free jets, especially supersonic jets, tends to propagate in the shear layer (i.e., the region with a large velocity gradient near the boundary of the jet) and induce time-varying oscillatory loads on obstacles in the flow path. The NRC concern is that pressures and densities vary nonmonotonically with distance along the axis of a typical supersonic jet, feeding and interacting with shear layer unsteadiness. In addition, for a typical supersonic jet, interaction with obstructions could © Copyright 2018 by NuScale Power, LLC 25

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 lead to backward-propagating transient shock and expansion waves that cause further unsteadiness in downstream shear layers. Synchronization of the transient waves with the shear layer vortices emanating from the jet break can lead to significant amplification of the jet pressures and forces (a form of resonance) that is not considered in ANSI/ANS 58.2. Should the dynamic response of the neighboring structure also synchronize with the jet loading time scales, further amplification of the loading can occur, including that at the source of the jet. General observations by Ho and Nosseir are that strong discrete frequency loads occur when the impingement surface is within 10 diameters of the jet opening, and that when resonance within the jet does occur, amplification of impingement loads can result (Reference 1.4.3.8). NuScale approach: This phenomenon is not applicable to pipe breaks in the NuScale plant, as discussed in Section 3.9.7.1 and Appendix B. 2.2.4.5 Blast Waves In the event of a high-pressure pipe rupture, the first significant fluid load on nearby SSC is induced by a blast wave in the surrounding air. A spherically expanding blast wave is approximated to be a short duration transient and analyzed independently of any subsequent jet formation. However, the expansion of blast waves in an enclosed space is not purely spherical, and reflections and amplifications need to be addressed. Blast waves are not considered in ANSI/ANS 58.2 for evaluating the dynamic effects associated with the postulated pipe rupture. NuScale approach: ((

                                                            }}2(a),(c) 2.2.4.6 Pipe Whip Standard Review Plan 3.6.2 establishes that pipe whip analyses should show that pipe motions do not result in unacceptable impact upon, or overstress of, any SSC:

to the extent that essential functions would be impaired or precluded . . . The analysis methods used should be adequate to determine the resulting loadings in terms of the kinetic energy or momentum induced by the impact of the whipping © Copyright 2018 by NuScale Power, LLC 26

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 pipe, if unrestrained, upon a protective barrier or a component important to safety and to determine the dynamic response of the restraints induced by the impact and rebound, if any, of the ruptured pipe. The SRP acknowledges that a determination can be made for pipe-on-pipe impact: An unrestrained whipping pipe should be considered capable of causing circumferential and longitudinal breaks, individually, in impacted pipes of smaller nominal pipe size, and of developing through-wall cracks in equal or larger nominal pipe sizes with thinner wall thickness, except where analytical or experimental, or both, data for the expected range of impact energies demonstrate the capability to withstand the impact without rupture. In case of a circumferential rupture, the need for a pipe-whip dynamic analysis may be governed by considerations of the available driving energy. Dynamic analysis methods used for calculating piping and restraint system responses to the jet thrust developed after the postulated rupture should adequately account for the following effects: (a) mass inertia and stiffness properties of the system, (b) impact and rebound, (c) elastic and inelastic deformation of piping and restraints, and (d) support boundary conditions. The SRP states that acceptable models for high-energy piping systems include:

Lumped parameter analysis that accounts for inertia and stiffness properties of the system and maximum initial clearances at restraints.
Energy balance in which kinetic energy from the first quarter cycle of movement of the ruptured pipe is converted to equivalent strain energy. Deformations should be compatible with the level of absorbed energy, and energy absorbed by pipe deformation may be subtracted from that available. Where rebound may occur, an amplification factor of 1.1 should be used unless another value is justified by detailed analysis.
Static analysis where the jet thrust force is represented by a conservatively amplified static loading. An amplification factor can be used to establish a forcing function.
Other attributes if justified.

NuScale approach: Inside the CNV, each of the postulated locations is shown to be acceptable on at least one of the following bases, in order of preference:

1. The piping has insufficient energy to whip (see Appendix C.4.1).

2. The length of whipping pipe is insufficient to reach essential SSC.

3. The whipping pipe is blocked by a barrier, which is a robust component (e.g., RPV, CNV) or wall, from reaching an essential SSC.

4. The essential structure, system or conponent is justified as being sufficiently robust to withstand the impact without loss of function.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

5. The number of redundant components is such that the damage caused by pipe whip combined with a single active failure does not cause a loss of an essential function.

6. Dynamic structural analysis of the impact force on a given essential structure, system, or component.

Unless clear (e.g., breaking a power cable de-energizes a component), damage caused by an HELB is assumed to fail a component in a way that does not allow the essential function to occur. For the NPM bay area, because break exclusion eliminates the need to postulate HELBs, pipe whip is not relevant. For integrity of the RXB concrete structure, an estimate is made of the kinetic energy of the whipping pipe, to be applied over a limited contact area using the methodology described for missile impact in Appendix C. Regarding multimodule effects, only the impact of ((

                     }}2(a),(c) 2.2.5    Standard Review Plan Section 3.6.3 Standard Review Plan 3.6.3, Rev. 1 (Reference 1.4.2.3) provides guidance on performing an LBB analysis acceptable to the NRC staff. If approved, the LBB analysis precludes the need to postulate HELBs, and the consequent pipe whip restraints and jet impingement barriers (and analysis of dynamic effects), but LBB still requires consideration of cracking.

The adequacy of detection of RCS leakage must be shown, in addition to consideration of the specifications of selecting reactor coolant leakage detection systems in Regulatory Guide 1.45, Rev. 1. In order to demonstrate reliability, redundancy, and sensitivity of the detection system, a margin of 10 on detection of a precursor leak rate is required (i.e., must be able to withstand a leak rate through a precursor crack that is 10 times that detectable). Leak-before-break may only be applied to high-energy, ASME Code Class 1 or 2 piping, although other applications can be considered. Leak-before-break is applied on a system or system segment basis (i.e., not to individual locations). A deterministic fracture mechanics and leak rate evaluation must be performed to demonstrate that the probability of pipe rupture is extremely low under conditions consistent with the design basis for the piping. Using fracture mechanics stability analysis or limit load analysis, the critical crack size for the postulated through-wall crack using loads from the normal plus safe shutdown earthquake (SSE) must be shown to have a margin of two between the critical crack size © Copyright 2018 by NuScale Power, LLC 28

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 and the leakage crack size (after application of the factor of 10 on leak rate). Also, a crack stability analysis is performed to demonstrate that leakage cracks do not become unstable if 1.4 times the normal plus SSE loads are applied or if the sum of the absolute values of the deadweight, thermal expansion, pressure, SSE (inertial), and seismic anchor motion loads is satisfactory. Limit load analysis requirements are identified for specific weld types. Confirmation of LBB acceptability is per ITAAC because as-built, not as-designed, configuration of pipe, supports, gaps, and etc. must be addressed. Pipe integrity degradation must consider water hammer, creep damage, erosion, corrosion, fatigue, and environmental conditions. Standard Review Plan 3.6.3 identifies material specifications and property testing. NuScale approach: NuScale has applied LBB only for MSS and FWS piping inside the CNV. Leak-before-break is not applied to RCS-connected piping due to its small size (i.e., NPS 2). Small piping LBB would involve such low leak rates that, after applying the factor of 10, make discerning leakage cracks subject to a threshold so low that it would likely be masked by normal, non-RCPB leakage. Section 3.6.3 in the FSAR provides a detailed description of the LBB analysis. © Copyright 2018 by NuScale Power, LLC 29

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.0 Methodology In Section 2.2.2, the NuScale approach for addressing regulatory expectations is briefly identified in relation to specific NRC regulations and guidance. This section discusses the methodology for demonstrating compliance with GDC 4 and addressing the related regulatory guidance. Details of how the analyses are performed and a summary of results are provided in Appendices A through F. 3.1 General Approach This PRHA report is prepared to supplement the information contained in the NuScale FSAR relative to the occurrence of postulated pipe ruptures inside the NPM and in the NuScale RXB. Figure 3-1 is a flowchart depicting the process for evaluation of potential line breaks. The NuScale methodology applicable to identification and assessment of pipe rupture hazards addresses:

design features of the NuScale plant.
compliance with NRC regulations and guidance.
identification of postulated rupture locations
characteristics of ruptures including break types and size.
determination of potential effects of HELBs and MELBs.
criteria for showing acceptability of essential SSC exposed to those effects.

External effects of HELBs and MELBs in the NPM and NuScale RXB do not adversely affect the ability to shut down and maintain core cooling. An integral, passive, multi-module, small modular reactor has different line break risks than the larger-scale, single-unit LWR for which regulatory guidance was developed and refined. In particular, the arrangements, number, and operating mechanism of essential systems differs from that assumed by regulatory guidance. NuScale specific differences are discussed in Section 2.1. © Copyright 2018 by NuScale Power, LLC 30

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.1.1 Essential Functions NuScale shutdown and core cooling functions rely on natural forces (i.e., buoyancy driven natural circulation) or local energy storage (i.e., nitrogen accumulator to close CIVs). The functions initiate without human interaction. In accordance with the footnote in BTP 3-3 that defines essential systems and components as those necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without off-site power, the essential systems in the NuScale design are identified below and discussed in the following sections:

RCS
module protection system (MPS)
neutron monitoring system (NMS)
chemical and volume control system (CVCS)
control rod assembly (CRA) and the control rod drive system (CRDS)
containment system (CNTS)
DHRS
ECCS
ultimate heat sink (UHS) / reactor pool In addition, the NuScale design is evaluated for the capability to maintain long-term PAM and plant DC electrical power in order to limit simultaneous multimodule effects due to a loss of AC power. The NuScale design does not rely on any Class 1E power or safety-related and essential systems or components external to the modules to perform any active safety function.

Where a line break is postulated to occur, the appropriate dynamic and environmental effects are considered, with the additional assumptions of a loss of off-site power and a single active failure. Because of the passive safety design, loss of off-site power is not a threat to the core cooling of any module, and highly reliable DC power, backed up by the ECCS valve inadvertent actuation block (IAB), ensures that multiple modules do not blow down simultaneously from de-energization of the ECCS valve solenoids. The IAB holds its ECCS valve closed until RCS pressure has decreased to the range of 1200 to 1000 psia, by preventing venting of the control chamber above the main valve disc even if the trip valve opens or the trip/reset actuator line is breached. 3.2 Description of Systems Important to Reactor Shutdown and Core Cooling Essential SSC are those needed for reactor shutdown and core cooling. The simplicity and passive safety features of the NuScale design result in a small number of SSC being required for reactor shutdown and core cooling. In some cases, only portions of an SSC are essential. Table 3-1 summarizes the safety-related and essential SSC and, in particular, which portions are necessary for shutdown and core cooling. © Copyright 2018 by NuScale Power, LLC 32

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 3-1. Safety-related and essential parts of structures, systems, and components vulnerable to break effects System Location Component RCS CNV All ECCS CNV RRVs & RVVs NPM bay Trip and reset valves* DHRS CNV Piping NPM bay Actuation valves Valve position indicators Passive condenser* NPM pool UHS RXB Spent fuel pool SGS CNV Piping MPS Various Separation Groups CNV CIVs CNTS NPM bay MSS tee fittings between CNV and CIVs Electrical penetration assemblies (EPAs) Hydraulic power unit skids CRDS CNV CRDM**

Submerged in pool

        ** Pressure boundary only Structures, systems, and components are classified as A1, A2, B1, and B2 in accordance with their safety and risk categories:

A1 both safety-related and risk-significant
A2 safety-related but not risk-significant
B1 risk-significant but not safety-related
B2 neither safety-related nor risk-significant

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.2.1 Reactor Coolant System Chemical and volume control system lines inside the CNV are part of the RCS. Reactor coolant system circulation is entirely contained within the RPV (i.e., there is no RCS loop piping). Hot coolant exiting the top of the core moves up through a riser by buoyancy-driven natural circulation, turns at the pressurizer baffle plate, passes downward around the tubes of the steam generators (SGs) where it is cooled, and then passes outside the core to reach the bottom of the RPV to rise again through the core. Pilot-actuated reactor safety valves (RSVs) are mounted on the RPV. 3.2.2 Module Protection System The MPS monitors conditions in the NPM and connected systems, and automatically executes safety-related functions when required. Each NPM has its own dedicated MPS that is not shared with other NPMs. The MPS maintains components in an energized, or nonactuated, state during normal operation. Plant design criteria require that Class 1E circuits in the CNV and in the NPM bay outside the CNV (under the bioshield) are qualified for environmental conditions, and also be evaluated for pipe whip and jet impingement. Class 1E circuits in the RXB are separated from areas containing high-energy piping. Class 1E circuits are routed or protected so that failure of the mechanical equipment of one division cannot disable Class 1E circuits or equipment essential to the performance of the safety-related function by the systems of the redundant division(s). The effects of failure or misoperation of a mechanical system on its own division is considered when the Class 1E circuits or equipment are required to mitigate the consequences of such failure or misoperation. The effects of pipe whip, jet impingement, water spray, flooding, radiation, pressurization, elevated temperature, or humidity on redundant electrical systems caused by failure, misoperation, or operation of mechanical systems are considered. The potential hazard of missiles resulting from failure of rotating equipment or high-energy systems are considered. Protection of nonhazard and limited hazard areas from pipe failure hazard areas is accomplished by the use of barriers, restraints, separation distance, or an appropriate combination thereof. The routing of Class 1E and associated circuits in pipe failure hazard areas conforms to the following requirements, unless it can be demonstrated that pipe failure cannot prevent the Class 1E circuits and equipment from performing their safety-related function.

Where the piping involved is qualified for design-basis events, is not assignable to a single division, and the pipe failure requires no protective action, Class 1E equipment, associated circuits, or raceways routed through the area are limited to a single division.
Where the pipe failure requires protective action, Class 1E equipment, associated circuits, or raceways are not routed through the area, except those cables that must terminate at devices or loads within the area.
Where the piping involved is qualified for design-basis events, it is assignable to a single division, and the pipe failure requires no protective action, Class 1E equipment, associated circuits, or raceways routed through the area are limited to the same division as the piping.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Where the piping involved is not qualified for design-basis events, Class 1E equipment, associated circuits, or raceways are not located in the area, except for those cables that must terminate at devices or loads within the area.

The location of RPV head penetrations used for instrumentation is shown in Figure 3-2. Class 1E circuits are routed or protected so that failure of the mechanical equipment of one division cannot disable Class 1E circuits or equipment essential to the performance of the safety-related function by the systems of the redundant division(s). The effects of failure or misoperation of a mechanical system on its own division are considered when the Class 1E circuits or equipment are required to mitigate the consequences of such failure or misoperation. The effects of pipe whip, jet impingement, water spray, flooding, radiation, pressurization, elevated temperature, or humidity on redundant electrical systems caused by failure, misoperation, or operation of mechanical systems are considered. The potential hazard of missiles resulting from failure of rotating equipment or high-energy systems is considered. No electric power is required to accomplish a reactor trip or initiate engineered safety features, but power is needed to maintain PAM indication. 3.2.3 Neutron Monitoring System The NMS provides neutron flux data for reactor trip, operating bypasses, and actuation of the MPS and information signals for PAM. A failure (e. g., broken signal cable) that causes an off-scale indication is registered as a fault by the MPS and does not adversely affect trip capability. Such a failure could remove one channel of PAM indication, therefore cables in areas of postulated breaks are routed out of range of HELB effects. 3.2.4 Chemical and Volume Control System The CVCS maintains RCS inventory during normal operation, provides purification and chemical injection to the RCS, provides pressurizer spray, and supplies heated water to warm up the RCS during start-up. None of these functions are essential or safety-related for HELB scenarios. 3.2.5 Control Rod Assembly and Control Rod Drive System The CRA includes neutron absorber control rods that are mechanically raised and lowered by control rod drive mechanisms (CRDMs) located on top of the RPV. The principal safety function is to achieve an immediate shutdown of the fission process when required by plant conditions. During normal plant operation, an electric coil in the CRDM is maintained energized to hold the control rods withdrawn from the core in a static position. Interruption of electric power de-energizes the CRDM electric coils causing the control rods to be unlatched and fall into the core (i.e., trip) via gravity. This passive insertion of negative reactivity results in the core becoming subcritical, and remaining subcritical during RCS cooldown to 70-degrees Fahrenheit with all rods inserted. The CRAs are located inside the RPV and, therefore, protected from the external effects of an HELB. Internal effects such a differential pressure forces induced by blowdown are not within the scope of this report. © Copyright 2018 by NuScale Power, LLC 35

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 No electric power is required to accomplish the trip. 3.2.6 Containment System The CNV is an ASME Class MC (steel) containment that is designed, analyzed, fabricated, inspected, tested, and stamped as an ASME Code Class 1 pressure vessel with a design pressure of 1000 psia. The CNV wholly contains the RPV and is mostly immersed in the reactor pool. The containment system (CNTS) is comprised of valves, fittings, and piping connecting systems inside and outside of the CNV (i.e., connecting the RCS lines inside containment to the CVCS piping outside). The CNTS boundary is from the CNV inner pipe nozzles out to the flange, where a pipe spool can be removed so that the NPM can be moved. The containment evacuation system line is open to the CNV to allow the operating vacuum pump to maintain the low internal pressure, and the flooding and drain line is normally isolated and exposed to CNV vacuum. These lines are not discussed in this report because they are not high- or moderate-energy. Each line connected to the RCS or open-ended to containment has two series CIVs in accordance with 10 CFR 50 Appendix A, GDC 55 and 56, except that both CIVs are located outside the containment within a single piece valve body that is welded directly to the nozzle. Each CIV ((

                                                                                   }}2(a),(c)

The MSS, FWS, DHRS, and reactor component cooling water system (RCCWS) lines are closed-loop systems inside containment (i.e., GDC 57). The MSS and FWS lines each have a single CIV of the same design as described above. Because of the need for DHRS hot leg connections, the main steam system CIVs are separated from the CNV nozzles by two tee fittings. A bypass line around each main steam system CIV is used to introduce steam for secondary system start-up before opening the MSIVs. The bypass valve is closed whenever the plant is operating. With the MSS and FWS systems isolated, secondary side water inventory is maintained for decay heat removal. The RCCWS lines are small diameter and use the same dual valve design as used for open-ended lines. Piping lines except those for the MSS and DHRS have only one weld between the CNV head nozzle and the CIVs. In addition to the CIVs, each normally open line directly connected to the RCS has a check or excess flow check valve directly welded to the outboard nozzle of the CIV body. Each feedwater system CIV body also contains an integral check valve that shuts upon flow reversal caused by a FWS line break outside the NPM. No electric power is required to close the CIVs. © Copyright 2018 by NuScale Power, LLC 36

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.2.7 Decay Heat Removal System The DHRS passively removes decay heat by natural circulation, to establish safe shutdown conditions following a reactor trip, without need for operator action or on-site or off-site power. The DHRS consists of two independent and redundant trains, and each train alone has 100 percent capacity to provide heat removal. During normal plant operation, the DHRS is in standby mode with flow blocked by closed DHRS actuation valves in the inlet line. These valves are maintained closed by energized actuator solenoids. When power to the solenoids is interrupted, the valves (which are the same design as the CIVs, except that they fail open) open through the force of the nitrogen pressurized accumulator. Opening either of two DHRS actuation valves on one of the two redundant trains provides sufficient cooling. The DHRS is comprised of two closed loop flow paths, each consisting of an SG, a DHRS passive condenser, and associated piping that provides natural circulation of secondary water flow from the SGs to the passive condensers, where heat is rejected to the UHS, the reactor pool. No electric power is required to open the DHRS actuation valves. 3.2.8 Emergency Core Cooling System The ECCS ensures core cooling by maintaining the core covered with water during design-basis events in which the system is actuated. Unlike other larger LWR designs, the NuScale ECCS does not require a source of water for injection or the availability of electric power. Decay heat removal occurs by releasing coolant to the CNV, which is cooled by condensation and conduction through the CNV wall to the reactor pool. Water in the CNV flows back into the RPV by natural circulation. Similar to the DHRS, the ECCS is started by interrupting power to the ECCS valves. The system has five main valves and associated hydraulic lines and actuator assemblies, including control solenoids. The main valves are bolted to nozzles on the RPV. Three reactor vent valves (RVVs) and two reactor recirculation valves (RRVs) provide sufficient flow area even if one valve fails to open. Each main valve has two associated pilot valves: the trip valve and the reset valve. The pilot valves have solenoids that are used to reposition the pilot valves and subsequently reposition the associated main valves. The trip valve solenoids are kept energized to prevent the ECCS main valves from opening. The reset valve solenoids are only energized to allow the ECCS main valves to close while the plant is being started. The trip and reset valves are part of a single manifold located on the exterior of the CNV, submerged in the pool. With the trip valve solenoid energized, a vent path for the control chamber above the main valve disc is blocked, and the control chamber is maintained pressurized. With the control chamber pressurized, the main valve is held shut. De-energizing the trip valve solenoid repositions the trip valve removing it from the control chamber vent path. The ECCS contains an inadvertent actuation block (IAB) feature to prevent opening the main valves at normal operating pressure. The IAB valve, bolted to the main valve body, is installed in the hydraulic line between the control chamber and the trip valve. When the © Copyright 2018 by NuScale Power, LLC 37

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 IAB valve is closed and RCS pressure is near normal, the control chamber cannot be vented, regardless of trip valve position or integrity of the trip/reset line. During normal power operation, the five ECCS main valves are closed, their IAB valves are open, and their trip valves are energized and closed. If electric power to them is interrupted, the trip valves de-energize and open, depressurizing the trip line. The IAB valves are forced closed by the differential pressure between the RCS and CNV. As a result, the ECCS valves remain closed. Should the RCS be completely depressurized before main valve control chamber pressure is vented, springs open the valve. The ECCS valves open in the following scenarios:

Because the decay heat removal capacity of the DHRS exceeds decay heat levels, it results in lowering RCS temperature and pressure. If the trip valve is de-energized, then reduction in RCS pressure and the resultant reduction in differential pressure across the actuation valve clears the IAB feature, so that RCS pressure (even though declining) opens each ECCS valve.
For a reactor coolant system HELB inside the CNV, the RCS depressurization and CNV pressurization clear the IAB, permitting the ECCS main valve disc to open as soon as the trip valve power in interrupted.

No electric power is required to open the ECCS valves because they are opened by RCS pressure (or springs). 3.2.9 Ultimate Heat Sink The UHS consists of a large pool complex where the NPMs and spent fuel are housed. The combined volume of water is in the associated water-retaining structures and components of the reactor pool, refueling pool, and spent fuel pool (SFP). The water volume in the reactor pool and refueling pool portions of the UHS is connected with the water volume in the SFP through the space above the top of the UHS weir wall. Water level is maintained during normal operation via interface with the spent fuel pool cooling system, and temperature is controlled using the reactor pool cooling system and the spent fuel pool cooling system. If AC power is lost, these nonsafety-related systems are unavailable. The UHS has capacity for decay heat from up to 12 NPMs operating at full power and a full SFP. Lowering of UHS level due to evaporation and pool boil-off during a loss of AC power event is gradual enough to ensure the DHRS passive condensers remain submerged for greater than 30 days without operator action, electric power, or addition of water. By the time the condensers uncover, decay heat is low enough that heat loss to ambient air is sufficient. Passive venting from the area of the RXB above the pools transfers the energy to the environment. No electric power is required to fulfill the UHS function of the NuScale plant. © Copyright 2018 by NuScale Power, LLC 38

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.2.10 Post-Accident Monitoring Post-accident monitoring is a nonsafety-related function that uses other systems components. Although no operator action is required for a design basis event to ensure core cooling for an unlimited duration, monitoring of the status of the NPMs is desirable and is necessary to meet regulatory guidance. Post-accident monitoring information is displayed on the safety display and indication system. Post-accident monitoring does not have a capability to control any equipment. Post-accident monitoring uses available instrumentation to monitor Type A, B, C, D, and E variables, as defined in Regulatory Guide 1.97, Rev. 4 (Reference 1.4.2.6). NuScale has no Type A variables. Type B, C, and D variables inside containment are listed in Table 3-2. The location of RPV head penetrations for instrumentation is shown in Figure 3-2. Separation groups B and C are preferred for PAM purposes because they are provided with highly reliable DC power via the EDSS for 72 hours, whereas groups A and D have DC power available for only 24 hours, assuming no operator action during a loss of off-site power or station blackout (SBO). For the purpose of satisfying PAM requirements for an HELB, only separation groups B and C are considered.5 A single failure of one of these is also assumed, although single failures for an SBO are not considered. Post-accident monitoring indication by at least one channel must still be available after an HELB, which requires both separation groups B and C to be protected for HELB effects. For protection against pipe whip and jet impingement, cables are routed at least 6.75 inches from the path of a whipping pipe from an RCS line terminal end break in the CNV. The cable is qualified for CNV design pressure and temperature, and for jet impingement effects, which bounds RCS line break conditions. Figure 3-3 shows the exterior, topside view of the break locations at the interior, underside of the CNV head. The functionality of PAM is neither safety-related nor essential, but is addressed in this report consistent with regulatory guidance. Power from the highly reliable DC power system (EDSS) is required for PAM indication. 5 The definition of station blackout in 10 CFR 50 states, in part: Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident. NuScale does not rely on off-site or alternate source AC power for any safety-related or essential function. The ECCS initiation occurs within two hours for reactor coolant system HELBs inside the CNV, removing the ECCS trip valve load. Removal of the ECCS solenoid load within two hours results in separation group A & D battery power being available for at least 48 hours. If a loss of all AC power occurred, operator action to provide alternate power to battery chargers within 48 hours sustains PAM but is not assumed. © Copyright 2018 by NuScale Power, LLC 39

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 3-2. Separation group B and C post-accident monitoring Type B & C instruments inside containment Variable Number of Sensors RPV Nozzle CNV Nozzle Numbers Numbers Wide Range RCS Hot 2 60, 63 18, 19 Core Inlet Temperature 12 40, 41 17, 18, 19, 20 Core Exit Temperature 12 40, 41 17, 18, 19, 20 Narrow Range CNT Pressure 2 N/A 18, 19 Wide Range RCS Pressure 2 40, 41 18, 19 Wide Range CNT Pressure 2** N/A 18, 19 RPV Riser Water Level 2 40, 41 18, 19 CNT Water Level 2 N/A 18, 19 ((

                                                                                               }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                }}2(a),(c)

Figure 3-3. Containment vessel head penetrations and break locations (breaks on underside) 3.3 Systems with Potential for High- or Moderate-Energy Line Ruptures The following sections provide a description of the high- and moderate-energy systems that could experience a pipe rupture. These systems are summarized in Table 3-3. The table identifies the line operating and design conditions, size, piping design code, and HELB status (i.e., the approach taken to demonstrate essential SSC are protected). The final design for piping systems beyond the NPM bay is the responsibility of the COL applicant, as stated in the NuScale FSAR 1.4.2.16 COL Items 3.6-1, 3.6-2, and 3.6-3. This includes final equipment location, pipe routing, support placement and design, piping stress evaluation, pipe break mitigation, and evaluation of subcompartment pressurization and multimodule effects. This report documents analyses of bounding scenarios that were performed to ensure the design, when finalized, can comply with NRC regulations and guidance. © Copyright 2018 by NuScale Power, LLC 41

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.3.1 Reactor Coolant System The RCS is wholly contained within the CNV and has no loop or other large piping to rupture. The RCS lines run between the RPV nozzles and the CNV nozzles for pressurizer spray (two lines), RPV high point degasification (hereafter, just degasification or degas), discharge, and injection. The RCS lines are NPS 2 Schedule 160. The degasification line is isolated with its CIVs closed during normal operation. Welds and fittings are minimized through use of pipe bends. 3.3.2 Containment System With the exception of the two reducing tees in each of the two MS lines discussed in Section 3.2.6 and below, the CIVs are welded directly to the CNV nozzles. The CNTS consists of a CNV nozzle on the inside of the CNV, another nozzle on the outside of the CNV, a single (MSS or FWS) or dual CIV in a single body, and two tees in each of the MSS lines. Also, a check valve is incorporated into the body of the feedwater system CIVs. 3.3.3 Chemical Volume and Control System The CVCS includes the RCS-connected lines off the CNV. The CVCS lines in the RXB are NPS 2, 21/2, and 3. The lines to the NPM consist of the following major segments: pressurizer spray, injection, degas, and letdown (each of which has a check or excess flow check valve welded to the CIV outboard nozzle). The piping is stainless steel. In the NPM bay, lines run between the inboard disconnect flanges and the pool wall. Dual, single-body CIVs are directly welded to the CNV nozzles and are part of CNTS. Outboard of the valves and check or excess flow check valves, NPS 2 flanged piping spool pieces provide the capability to disconnect the NPM from system piping in preparation for movement for refueling. In the Reactor Building, CVCS lines connect to the balance of the system (i.e., nonregenerative heat exchanger, ion exchangers). The RXB portions of the CVCS have many different state points, including some with high pressure/low temperature. 3.3.4 Emergency Core Cooling System As described in Section 3.2.8, the ECCS has no physical piping, other than the trip and reset lines, which are small diameter (i.e., less than one-inch diameter). The design basis blowdown for the RCS and CNV is inadvertent opening of an ECCS valve. As described in Section 3.2.8, the IAB is provided to avoid an inadvertent actuation while an NPM is near normal operating pressure. Discharge from the RRVs is directed downward away from essential SSC and discharge from the RVVs is directed toward the CNV walls through diffusers. The ECCS main valves are bolted directly onto the RPV. 3.3.5 Steam Generating System The steam generator system (SGS) is the in-containment feed water piping, SG, and main steam piping. In the CNV, four NPS 8 Schedule 120 SGS steam lines from the SG outlet © Copyright 2018 by NuScale Power, LLC 42

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 plena connections on the RPV merge into two NPS 12 Schedule 120 steam lines to the CNV nozzles. The piping has a design pressure equal to that of the RCS. The two NPS 5 feed lines from the CNV nozzles split into two NPS 4 lines (total of four) that supply feedwater to the SG plena. Just upstream of the split, the DHRS return line tees in. The SGS lines are Schedule 120 with a design pressure and temperature equivalent to that of the RCS. 3.3.6 Main Steam System In the NPM bay outside the CNV (under the bioshield), two main steam lines consist of NPS 12 flanged piping spools that provide the capability to disconnect the NPM in preparation for movement for refueling, and a fixed section of pipe that projects through the pool bay wall into the pipe galleries on each side. The spools include ball joints to allow for small variations in NPM position. The lines are made from stainless steel. In the RXB, the NPS 12 lines include an isolation valve with a NPS 4 bypass line. Pipe routing, weld locations, and placement of hangers have not yet been finalized. To ensure that the RXB and its essential SSC are adequately protected, NuScale has evaluated bounding HELBs and established design requirements for separation. Main steam piping designed to ASME B31.1. 3.3.7 Feedwater System In the NPM bay, two flanged NPS 6 piping spools provide the capability to disconnect the NPM in preparation for movement for refueling. The lines are stainless steel, except for a small section from the inboard flange into the pipe gallery, which is SA-335 PA11. In the RXB, the NPS 6 lines include a check valve. Pipe routing, weld locations, and placement of hangers have not been finalized. To ensure that the RXB and its essential SSC are adequately protected, NuScale has evaluated bounding HELBs and established design requirements for separation. The FSW piping is designed to ASME B31.1. 3.3.8 Decay Heat Removal System Cool water from the passive condenser is returned to the FWS piping by a NPS 2 Schedule 160 line located both inside the CNV and inside the bay submerged in the pool. In the NPM bay, a NPS 8 Schedule 160 line (four lines total) runs from each MSS reducing tee inboard of the main steam CIV through a normally closed, fail open, 6-inch DHRS actuation valve to the passive condenser. 3.3.9 Reactor Component Cooling Water System The RCCWS is a moderate-energy system supplying cooling water to the CRDMs. Inside the CNV, supply and return lines that are part of the CRDS consist of curved headers connected to which are attached flexible cooling hoses to each CRDM. © Copyright 2018 by NuScale Power, LLC 43

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 The effects of any RCCWS rupture are bounded by those of HELBs. 3.3.10 Auxiliary Boiler System The auxiliary boiler system (ABS) is a nonsafety, nonseismic system designed to supply steam to systems where main steam is not available or is not preferred. It is a COL applicant-provided system. The ABS consists of two separate subsystems, neither of which is present in the CNV or NPM bay. The high-pressure system is dedicated to supplying steam to the module heatup system (MHS) heat exchangers during start-up and has two separate headers, with a limited amount of piping in the RXB. The low-pressure portion is outside the RXB and is not discussed further in this report. The high-pressure portion provides up to 18,000 lbm/hr of 575-degree steam at 1100 psig. The two high-pressure boilers can each supply heat for one module on one side of the plant and are equipped with a pressure relief valve. The routing of auxiliary boiler lines is not final. Based upon an estimated NPM heatup time of 24 hours, no need for ABS steam for NPMs going into a refueling outage, a two-year refueling cycle, and a full 12 NPM plant, each header of the ABS has steam in RXB piping for about 72 hours/year (i.e., one module start-up per header every four months). Branch Technical Position 3-4 paragraph B.2.(v) states that leakage cracks instead of breaks may be postulated in those fluid systems that qualify as high energy for only a short operational period. NuScale FSAR Section 3.6.1.1 defines a short period to include being at high-energy pressures or temperatures for less than 1 percent of the plant operation time. The high-energy portion of the ABS in the RXB is expected to operate less than one percent of the year (i.e., 86.4 hours), so the only external effect needing evaluation is from leakage cracks. 3.3.11 Module Heatup System The MHS conveys heat from the ABS to the CVCS to heat reactor coolant for an NPM during start-up until nuclear heat is available. The MHS heating combined with simultaneous heat removal in the SGs drives RCS flow during the heatup. The two MHS subsystems each contain two heat exchangers, with each subsystem serving six modules on one of two sides of the plant. The CVCS recirculation pumps are used to supply reactor coolant through the MHS and then to the respective NPM. Each MHS subsystem provides heat to only one NPM at a time. Although unlikely, the MHS may also be used during shutdown to maintain RCS flow if decay heat is insufficient. Consistent with the ABS, the MHS is expected to operate less than one percent of the year and is, therefore, evaluated as a moderate-energy system for effects of leakage cracks. 3.4 Break Characteristics Where postulated breaks might occur, the characteristics of those breaks (i.e., thermodynamic conditions) are identified as inputs to an evaluation of external effects. In general, bounding conditions are used in analysis of breaks. For example, the CVCS piping has considerable variation of fluid temperature and pressure with location in the © Copyright 2018 by NuScale Power, LLC 44

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 RXB and with plant initial conditions. Rather than evaluating many specific conditions, initial intact system temperature and pressure values are selected to maximize the mass and energy release from the HELB in the affected area of the plant and are, therefore, conservative for evaluating multiple break locations. Postulated breaks are circumferential. Longitudinal cracks are not applicable in the CNV, because piping NPS 4 and larger meets LBB criteria. Also, longitudinal breaks need not be considered in the NPM bay outside the CNV (under the bioshield), based on meeting criteria for not considering circumferential breaks. In the rest of the RXB, effects of longitudinal breaks (with break flow areas equal to the piping flow area) are bounded by circumferential breaks. Table 3-4 summarizes potential break locations in the NuScale plant. © Copyright 2018 by NuScale Power, LLC 45

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 3-3. High-energy and moderate-energy system piping characteristics Max. Operating Design a Largest System Location Line Purpose Press. Temp. Press. Temp. Piping Code PRHA Statusb Remarks Piping (psia) (ºF) (psia) (ºF) Normally RCS CNV Degas 1850 625 2100 650 2 Sch 160 ASME III, Cl 1 BE(iii)/evaluated isolated PZR Spray (2) 1870 455 2100 650 2 Sch 160 ASME III, Cl 1 BE(iii)/evaluated Injection 1870 455 2100 650 2 Sch 160 ASME III, Cl 1 BE(iii)/evaluated Discharge 1850 500 2100 650 2 Sch 160 ASME III, Cl 1 BE(iii)/evaluated CVCS NPM bay Degas 1850 500 2100 650 2.5 Sch 160 ASME III Cl 3 BE(ii)(iii) PZR Spray 1850 543 2250 650 2.5 Sch 160 ASME III Cl 3 BE(ii)(iii) Injection 1875 543 2250 650 2.5 Sch 160 ASME III Cl 3 BE(ii)(iii) Discharge 1850 500 2250 650 2.5 Sch 160 ASME III Cl 3 BE(ii)(iii) B31.1, seismic RXB Various 1875 543 2250 650 4 Sch 160 Evaluated Cat II or III ECCS CNV N/A 1850 543 2100 650 N/A N/A N/A Piping 1 c SGS CNV FWS (2) 550 300 2100 650 5 Sch 120 ASME III Cl 2 LBB FWS (4) 550 300 2100 650 4 Sch 120 ASME III Cl 2 LBB MSS (4) 500 585 2100 650 8 Sch 120 ASME III Cl 2 LBB MSS (2) 500 585 2100 650 12 Sch 120 ASME III Cl 2 LBB MSS NPM bay MSS (2) 500 585 2100 650 12 Sch 120 ASME III Cl 2 BE(ii)(iii) MSS CIV Open for 500 585 2100 650 2.5 Sch 160 ASME III Cl 2 BE(ii)(iii) bypass (2) heat-up B31.1; seismic MSS NPM bay MSS (2) 500 585 2100 650 12 Sch 120 BE(ii)(iii) Cat I B31.1; seismic Bounding RXB MSS (2) 500 575 1000 650 12 Evaluated Cat I analysis FWS NPM bay FWS (2) 511 300 2100 650 4 Sch 120 ASME III Cl 2 BE(ii)(iii) B31.1; seismic FWS (2) 511 300 2100 650 6 BE(ii)(iii) Cat I B31.1; seismic Bounding RXB FWS (2) 540 300 1000 650 6 Evaluated Cat I analysis © Copyright 2018 by NuScale Power, LLC 46

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Max. Operating Design a Largest System Location Line Purpose Press. Temp. Press. Temp. Piping Code PRHA Statusb Remarks Piping (psia) (ºF) (psia) (ºF) 8 Sch 160 4 lines tee DHRS NPM bay Hot leg (4) 1400 635 2100 650 ASME III Cl 2 BE(ii)(iii) 6 Sch 160 to 2 Condensate 1400 310 2100 650 2 Sch. 160 ASME III Cl 2 BE(ii)(iii) 2 lines return CRDM Bounded by Moderate RCCW NPM bay 80 121 165 200 2 B31.1 supply/return HELBs energy Moderate ABS RXB Steam to MHS 1100 575 1250 650 4 B31.1 Operates <1% energy Hot water for Moderate MHS RXB 1850 555 2250 650 3 B31.1 Operates <1% NPM heat-up energy a Systems in more than one region of plant are listed in multiple places. b BE indicates piping is analyzed against break exclusion criteria with (ii) and/or (ii) referring to the applicable criteria of BTP 3-4 B.A. used to determine location of postulated breaks (i.e., BE(iii) means terminal end breaks are assumed but intermediate locations are evaluated against B.A.(iii)). Evaluated means HELB external effects are considered where break are postulated to occur. c Only piping is actuator (trip) line which is normally isolated by IAB while operating. © Copyright 2018 by NuScale Power, LLC 47

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 3-4. Characteristics of blowdown at postulated break locations System Plant Line Purpose Break Flow Fluid Remarks Locationa Locationb Directionc Stated RCS CNV Injection RPV wall From RPVe Flashing Break at mid-height on RPV side: no nearby targets RPV wall From pipe Flashing Break at mid-height on RPV side: no nearby targets CNV head From pipe Flashing CNV head From nozzlee Flashing e Discharge RPV wall From RPV Flashing Break at mid-height on RPV side: no nearby targets RPV wall From pipe Flashing Break at mid-height on RPV side: no nearby targets CNV head From pipe Flashing CNV head From nozzlee Flashing e Degas RPV head From RPV Steam RPV head From pipe Steam Little steam in pipe between break and closed CIV CNV head From pipe Steam CNV head From nozzlee Steam Little steam in pipe between break and closed CIV PZR spray RPV head From RPVe Flashing Blowdown turns to steam after liquid blows from line RPV head From pipe Flashing CNV head From pipe Flashing Blowdown turns to steam after liquid blows from line e CNV head From nozzle Flashing CVCS RXB Supply to RCS High T pipe To NPM Flashing Bounding analysis applicable to any break location Blowdown terminated by check valve adjacent to CIV, From NPM Flashing so system side blowdown is limiting Low T pipe N/A Liquid Bounding analysis applicable to any break location © Copyright 2018 by NuScale Power, LLC 48

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 System Plant Line Purpose Break Flow Fluid Remarks Locationa Locationb Directionc Stated Blowdown terminated by excess flow check valve Discharge High T pipe From RCS Flashing adjacent to CIV, so system side blowdown is limiting From system Flashing Bounding analysis applicable to any break location DHRS CNV Return FWS line From DHRS Flashing From FWSe Flashing e CNV wall From DHRS Flashing From FWS Flashing MSS RXB To turbines Anywhere From NPM Steam Bounding analysis applicable to any break location Anywhere To NPM Steam Return flow bounded by forward flow analysis FWS RXB From turbines Anywhere From NPM Flashing Backflow from SG limited by FW check valve closure Anywhere To NPM Flashing Bounding analysis applicable to any break location From RCCWS CNV CRDM Anywhere Liquid Moderate energy; effects bounded by other analyses upstream From NPM bay CRDM Anywhere Liquid Moderate energy; evaluated upstream Aux. Not Leakage cracks only (high-energy conditions exist less RXB Steam to MHS From boilers Steam Boiler applicable than one percent of plant life) a The evaluation considers three areas of plant, which are inside the CNV, in the NPM bay outside the CNV (under the bioshield), and throughout the RXB. b May be specific location or Anywhere, which means evaluation applies to all potential break locations in that area of the plant; locations satisfying break exclusion or LBB criteria are not included as they are excluded from external effects evaluation (except leakage cracks for LBB and for MSS and FWS piping meeting break exclusion and located in the NPM bay outside the CNV (under the bioshield)). c For the given break location, which end of pipe break is considered. d Flashing is from system having liquid with low enough subcooling to cause two-phase blowdown. e Nozzle does not whip. © Copyright 2018 by NuScale Power, LLC 49

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.5 Restraints, Barriers, and Shields Pipe whip restraints may be used to limit the motion of a broken pipe to prevent it from hitting an essential structure, system, or component. Protection for pipe whip and jet impingement is also available through barriers afforded by walls, floors, and other structures. Sufficiently large and robust SSC can also function as a pipe whip barrier or jet impingement shield. 3.5.1 Pipe Whip Restraints Pipe whip restraints constrain movement of a broken pipe for purposes of preventing or limiting the severity of contact with essential SSC. Restraints installed only for purposes of controlled pipe whip are not ASME Code components; restraints that also serve a support function under normal or seismic conditions are designed to ASME criteria. The design criteria for pipe whip restraints are:

Pipe whip restraints do not adversely affect structural margin of piping for other conditions.

o Restraint design does not restrict thermal expansion and contraction. o The restraint design either: a) does not carry loads during normal operation or seismic events or b) the structural analysis includes a conservative load combination.

Pipe whip restraints are located as close to the axis of the reaction thrust force as practicable. Pipe whip restraints are generally located so that a plastic hinge does not form in the pipe using the methodology of Section C.4.1. If, due to of physical limitations, pipe whip restraints are located so that a plastic hinge can form, the consequences of the whipping pipe and the jet impingement effect are further investigated, as discussed in Appendix C. Lateral guides are provided where necessary to predict and control pipe motion.
Generally, pipe whip restraints are designed and located with sufficient clearances between the pipe and the restraint, such that they do not interact and cause additional piping stresses. A design hot position gap is provided that allows maximum predicted thermal, seismic, and seismic anchor movement displacements to occur without interaction.

o Exception to this general criterion may occur when a pipe support and restraint are incorporated into the same structural steel frame, or when a zero design gap is required. In these cases, the pipe whip restraint is included in the piping analysis and designed to the requirements of pipe support structures for all loads except pipe break, and designed to the requirements of pipe whip restraints when pipe break loads are included.

In general, the pipe whip restraints do not prevent access required to conduct in-service inspection examination of piping welds. When the location of the restraint makes the piping welds inaccessible for in-service inspection, a portion of the restraint is designed to be removable to provide accessibility.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Analysis of pipe whip restraints o Is either dynamic or conservative static.

o Static analysis includes a dynamic load factor of 2.0 potential increase by a factor of 1.1 in loading due to rebound. o Loading combination includes dead weight, seismic, and the jet thrust reaction force. o The criteria for analysis and design of pipe whip restraints for postulated pipe break effects are consistent with the guidelines in ANSI/ANS 58.2-1988. o Design is based on energy absorption principles by considering the elastic-plastic, strain-hardening behavior of the materials used. o Non-energy absorbing portions of the pipe whip restraints are designed to the requirements of AISC N690 Code. o Except in cases where calculations are performed to determine if a plastic hinge is formed, the energy absorbed by the ruptured pipe is assumed to be zero. That is, the thrust force developed goes directly into moving the broken pipe and is not reduced by the force required to bend the pipe. o In that a HELB is an accident (i.e., infrequent) event, pipe whip restraints are single use: allowed to deform provided the whipping pipe is fully restrained throughout the blowdown. Where structural members of a restraint are designed for elastic response, a dynamic increase factor is used. o Allowable strain in a pipe whip restraint is dependent on the type of restraint. Stainless steel U-bar - this one-dimensional restraint consists of one or more U-shaped, upset-threaded rods or strips of stainless steel looped around the pipe but not in contact with the pipe. This allows unimpeded pipe motion during seismic and thermal movement of the pipe. At rupture, the pipe moves against the U-bars, absorbing the kinetic energy of pipe motion by yielding plastically. Structural steel - this two-dimensional restraint is a stainless steel frame encircling the pipe that does not restrict pipe motion for normal operation or earthquakes. Should a rupture occur, the pipe motion brings it into contact with the frame, absorbing the kinetic energy of the pipe by deforming plastically. Crushable material - if used, the allowable energy absorption of the material is 80 percent of its capacity based on dynamic testing performed at equivalent temperatures and at loading rates of +/-50 percent of that determined by analysis. Note that a wall penetration may also serve as a two-dimensional pipe whip restraint, provided the wall has sufficient strength to resist the pipe load. © Copyright 2018 by NuScale Power, LLC 51

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Material properties are consistent with applicable code values, with strain-rate stress limits 10 percent above code or specification values, consistent with NRC guidance (SRP 3.6.2).

3.5.1.1 Pipe Whip Barriers Standard Review Plan 3.6.2 identifies that an unrestrained, whipping pipe need not be assumed to cause ruptures or through-wall cracks in pipes of equal or larger NPS with equal or greater wall thickness. By extrapolation, a structure, system, or component made of metal of equivalent or better yield strength, equal or larger diameter, and equal or greater wall thickness does not only not leak or crack but also obstructs further travel of the whipping pipe, protecting SSC farther away from being struck. Table 3-5 provides a comparison of potential whipping pipes and the SSC credited to act as barriers. The numbers in { } brackets are the factor by which the barrier diameter (pipe size) and wall thickness exceed that of the whipping pipe, where a minimum value of 1.0 for both satisfies the SRP 3.6.2 guidance for pipe-on-pipe impact not causing a crack or rupture. Therefore, the SSC listed in Table 3-5 are considered to serve as pipe whip barriers without further evaluation. Concrete floors, walls, and ceilings can also serve as pipe whip barriers but require a more quantitative approach as described in Section 3.9.5 and Appendix C. 3.5.1.2 Jet Impingement Shields NRC guidance does not have specific criteria for judging suitability of a structure, system, or component as a jet shield. Regarding impingement effects, if the following criteria are met, then the structure, system, or component is judged capable of serving as a shield:

The diameter and wall thickness of the shield meet the criteria for a pipe whip barrier with a size equal or greater than that of the broken pipe.
The barrier is of sufficient area and positioned to subtend a solid angle from the pipe break opening (considering potential pipe whip) that covers the structure, system, or component to be protected.
The barrier is solid (without openings) to the extent that no direct line of sight exists from the break opening to the structure, system, or component. This criterion allows for some indirect passage of spray through an opening, but environmental qualification for pressurization and flooding demonstrates functionality. The possibility of pipe whip affecting the location of the pipe break exit must be considered.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 3-5. Comparison of sizes of whipping pipe to potential barriers for high-energy line breaks in the containtment vessel Component Pipe Size Outer Diameter Wall Thickness (in.) (in.) NPS 2 Schedule RCS lines (( 160 CNV N/A RPV N/A CRDM latch housing N/A lower sectiond N/A RXB walls

                                                                                         }}2(a),(c) a without cladding b

varies with vertical location; minimum value in range of pipe break locations shown c minimum in RXB areas containing high energy piping within range of a whipping pipe d along most of its length, the housing is surrounded by magnetic coils that are about 2 inches thick 3.5.1.3 Pipe Whip Shields and Jet Impingement Barriers in NuScale Design The RPV and CNV are thick-walled components that serve as barriers and shields to isolate effects of HELBs. The NuScale RXB includes a functional requirement to accommodate the effects of environmental conditions associated with normal operations, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. Specifically the RXB is to be appropriately protected against dynamic effects, including those of missiles, pipe whipping, and discharging fluids. The design of the RXB considers protection of on-site electric power against water damage, flooding, jet impingement, and pipe whip resulting from failure of nearby piping. In the RXB, concrete walls, floors, and ceilings serve as barriers separating the effects of HELBs from areas containing reliable DC power and cables used for PAM. They also maintain structural integrity of the RXB. This requires that the RXB concrete structures be capable of resisting pipe whip impact and jet impingement. Analysis of pipe whip impact consistent with established methods for missile impact (i.e., tornado missiles) on concrete is provided in Appendix C. No pipe whip impact need be considered in the NPM bay under the bioshield because potential break locations in that area satisfy break exclusion criteria. For jet impingement, the potential for a jet to breach a wall is less than for the pipe whip impact force because the jet expansion distributes the force over a wider area. © Copyright 2018 by NuScale Power, LLC 53

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.5.2 Susceptibility of Essential Structures, Systems, and Components by Plant Location The potential for and consequence of line breaks depends on the location of the rupture being postulated. Three areas of the NuScale plant are considered separately in the following subsections. 3.5.2.1 Inside the Containment Vessel

Essential components are the RCS, MPS, NMS, EPAs, CNV, DHRS, ECCS valves, and CRDMs o If more than one MPS indication of a type (including NMS) loses its signal because its cable is severed by pipe whip or jet impingement, a reactor trip and/or safety component actuation occurs. For example, breaking signal lines for two or more RCS hot temperature or pressurizer level indications causes a reactor trip.

o The electrical penetration assemblies (EPAs) form part of the containment boundary. o Although not essential to ensure long term shutdown and core cooling, NuScale has evaluated the availability of PAM indication following a pipe rupture in order to satisfy NRC guidance. o The reactor safety valves are not needed if a HELB depressurizes the RPV (MSS and FWS ruptures are excluded by LBB). o Essential components inside the CNV are qualified for exposure to 1000 psia saturated steam caused by ECCS initiation and are isolated from the outside environment by the walls of the CNV. o Essential components fail to the safe condition upon loss of power or control signal (i.e., CRDS and ECCS).

A precursor crack can be detected by an increase in CNV pressure or CES liquid accumulation.
Potential for an inadvertent ECCS valve opening due to loss of power to the trip valve (i.e., wire breaks) or breach of the trip/reset line is averted by the IAB.
Piping is not insulated.
The NPM piping, including MSS and FWS, is stainless steel.
Piping runs are short.
The areas at the top of the RPV head and the underside of the CNV head are congested.

3.5.2.2 In the NuScale Power Module Bay

Essential components are the MPS temperature sensor under the bioshield, CNV, CIVs, EPAs, DHRS actuation valves, DHRS condenser (submerged), and ECCS trip/reset valves (submerged)
Essential components fail to the safe condition upon loss of power or signal (i.e., CIVs and DHRS actuation valves).

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Post-accident monitoring indication is also evaluated.
Piping in the containment penetration area is stainless steel.6
MSS and FWS piping has a design pressure and temperature of 2100 psia and 650 degrees, similar to the RCS.
High-energy line breaks are excluded by satisfying break exclusion criteria.

Nonmechanistic breaks in an MSS or FWS pipe in accordance with BTP 3-3 are discussed in Section 3.5.2.5, and effects of leakage cracks are evaluated.

The NPM bay under the bioshield is vented to the RXB to limit peak pressure and temperature.

3.5.2.3 In the Reactor Building (pipe routing in the RXB is not finalized7)

Structural damage to the RXB that could affect pool integrity has been evaluated.
Functionality of PAM indications and reliable DC power is ensured by separation from areas where high-energy piping is present.
No pipe ruptures in the RXB affect the control room, which is located in a separate building
Multi-module effects such as pipe-whip induced ruptures are considered. Three interactions are evaluated to determine if an MSS or FWS HELB in one module could:

o Structurally damage the RXB due to pipe whip impact or subcompartment overpressurization, potentially affecting other NPMs. o Cause a pipe whip to impact an adjoining NPM piping. Fluid release occurs too late to reinforce the initial HELB blast wave, and any secondary rupture blast wave is less severe because the piping is smaller. Ability for unaffected NPMs to be safely shut down and to provide long-term core cooling is not affected by the occurrence of an HELB in one or more other NPMs. Although piping arrangements are not finalized (COL Items 3.6-1, 3.6-2, 3.6-3), Figure 3-4 shows the potential overlap in the pipe galleries of MSS (bright green) and FWS (pale green) piping from adjacent NPMs (light blue): An MSS line impacting an equivalent size and schedule MS line does not cause a rupture or leakage crack per SRP 3.6.2, paragraph III.2. A main steam system HELB could impact the bypass line, causing up to a 4-inch diameter rupture in an adjoining NuScale Power Module MSS. This rupture represents a potential to increase NPM steam mass and energy 6 Feedwater system piping passing through the NPM bay wall is chrome-moly alloy SA-335 P11. 7 The final design for piping systems beyond the NPM bay is the responsibility of the COL applicant, as stated in the NuScale FSAR (Reference COL Items 3.6-1, 3.6-2, and 3.6-3). This includes final equipment location, pipe routing, support placement and design, piping stress evaluation, pipe break mitigation, and evaluation of subcompartment pressurization and multi-module effects. However, this report documents analysis of bounding scenarios that were performed to ensure the design, when finalized, can comply with NRC regulations and guidance. © Copyright 2018 by NuScale Power, LLC 55

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 release, and has been bounded by the overpressurization analysis performed for the pipe gallery. An MSS line impacting another NuScale Power Module FWS line could cause it to break. However, the second modules FWS line break cannot cause additional major ruptures because the other lines are equivalent or larger size and schedule.8 Because of the lower enthalpy compared to MSS lines and the double-ended discharge (in an FWS break, the FWS check valve quickly shuts off flow from the SG), an MSS bypass line rupture causes higher pressures. Therefore, a collateral break of a feedwater line is not limiting.

Environmental conditions such as high pressure and temperature or flooding that adversely affect another NPMs essential equipment are evaluated.

Because avoiding a collateral accident in another module is a design objective, the COL applicant needs to assess the final piping arrangements for the possibility of interaction. Where a rupture in an adjacent module cannot be ruled out, pipe whip restraints or barriers should be included. 8 Damage to small diameter (i.e., instrument) lines could occur but does not affect the ability to shut down and maintain cooling in other NPMs and does not increase compartment pressure. © Copyright 2018 by NuScale Power, LLC 56

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                }}2(a),(c)

Figure 3-4. Adjacent NuScale Power Module overlap of main steam system and feedwater system piping in the Reactor Building pipe gallery 3.5.2.4 Applicable Dynamic and Environmental Effects Dynamic and environmental effects are evaluated based on the postulated rupture location, thermodynamic conditions, and the break mechanism:

Where break exclusion criteria are satisfied, no rupture or leakage cracks are postulated but nonmechanistic breaks are considered (see Section 3.5.2.5).
Where LBB applies, no dynamic effects (i.e., pipe whip, blast, jet impingement, pressurization) are required to be considered, and the leakage effects are negligible because the allowable crack size is small.
Remaining postulated rupture locations are evaluated for pipe whip, blast, jet reaction loads, jet impingement loads, pressurization, and flooding effects.

The magnitude of the jet reaction load and the piping configuration determines if a pipe whips and, if so, its motion and impact force depend on the relative geometry of the pipe, its restraints and barriers, and potential target SSC. The ZOI and pressure force of jet impingement are conservatively evaluated. Inside the CNV, subcompartment pressurization for postulated breaks is bounded by analysis for ECCS. Outside the CNV, pressurization caused by postulated breaks is limited by venting to a differential pressure within the capability of the RXB structure. Dynamic amplification and resonance do not occur as a result of HELBs in the NuScale plant, as discussed in Appendix B. © Copyright 2018 by NuScale Power, LLC 57

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.5.2.5 Non-mechanistic Secondary Line Breaks in Containment Penetration Area Branch Technical Position 3-3 B.1.(a)(1) specifies: Even though portions of the main steam and feedwater lines meet the break exclusion requirements of item 2.A(ii) of BTP 3-4, they should be separated from essential equipment. Designers are cautioned to avoid concentrating essential equipment in the break exclusion zone. Essential equipment must be protected from the environmental effects of an assumed nonmechanistic longitudinal break of the main steam and feedwater lines. Each assumed nonmechanistic longitudinal break should have a cross sectional area of at least one square foot and should be postulated to occur at a location that has the greatest effect on essential equipment. The following considerations form the basis for this guidance:

The MSS and FWS piping are generally the largest, high-energy piping near containment boundary.
The lines have a single CIV outside containment, in accordance with GDC 57 for lines closed inside containment.
Piping is usually made of less-corrosion-resistant material than that used for the NuScale design: MSS and FWS piping in many pressurized water reactors is carbon or low-alloy steel, which have greater susceptibility to degradation than stainless steel.

Analyzing for nonmechanistic ruptures ensures that multiple essential SSC are capable of withstanding the effects of a limited piping failure should one occur. In the NuScale plant, CIVs are located outside the containment and exposed to the same environmental conditions, making protection against unexpected ruptures particularly important. However, the NuScale design has the following characteristics that make nonmechanistic ruptures low risk:

The essential SSC in the vicinity of the MSS and FWS piping to which break exclusion criteria apply are CIVs, DHRS actuation valves, and instrumentation cables and sensors.
Unlike safety-related valves in other plant designs that use motor-operators, the NuScale CIVs are ((

                                                                                                 }}2(a),(c).

The DHRS actuation valves similarly fail open.

Failure of the NuScale MSS and FWS piping is unlikely because o Piping in the containment penetration area is made of stainless steel.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 o The physical length of MSS and FWS piping in the containment penetration area is zero (i.e., there are only valves and fittings).9 o MSS and FWS piping has a design pressure and temperature of 2100 psia and 625-degrees Fahrenheit, respectively, similar to RCS piping. Table 3-6 shows a comparison of new design PWR MSS and FWS piping in the containment penetration area. Table 3-6. Comparison of main steam system and feedwater system piping in containment penetration area Plant MSS Piping FWS Piping Material Size Pressure Temp. Material Size EPR SA-106 Grade C NPS 30 1111 psig 558F SA-106 Grade B NPS 20 AP1000 SA-335 Gr. P11 NPS 38 836 psia 523F SA-335 Gr. P11 NPS 20 APR1400 SA-106 Grade C NPS 32 992 psia 544F SA-335 Gr. P22 NPS 24 APWR SA-106 Grade B NPS 32 931 psia 536F SA-335 Gr. P22 NPS 18 & 16 NuScale SA-312 304/304L NPS 12 500 psia 585F SA-312 304/304L6 NPS 4 & 5 The flow area for the nonmechanistic longitudinal break (1 ft2) specified in BTP 3-3 is disproportionately large for a small modular reactor with smaller pipe sizes. NuScale MSS piping is NPS 12 Schedule 120 and the FWS piping is NPS 6 Schedule 80 where it exits the bay area. For those piping sizes, a 1 ft2 flow area would be about 142 percent for MSS (552 percent for FWS) of the area for a full circumferential rupture, which is unrealistic. Additionally, BTP 3.4 B.3.(iii) specifies postulating leakage cracks with a flow area of one-half of a pipe diameter by one half pipe wall thickness in piping in the vicinity of essential SSC, regardless of system. This guidance yields an equivalent flow area of 2.7 in.2. The NuScale approach for nonmechanistic breaks of MSS and FWS piping in the containment penetration area considers these design differences from the larger LWR plants. 9 There is an approximately one foot long NPS 2 bypass around each MSS CIV, ending in the normally closed MSS bypass valve. © Copyright 2018 by NuScale Power, LLC 59

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.6 Break Exclusion Branch Technical Position 3-4 Paragraph B.A.(iii) identifies specific criteria for which ruptures need not be considered; at terminal ends in piping from the containment wall to and including the inboard or outboard isolation valves (usually referred to as the containment penetration area break exclusion zone). This is necessary due to constraints on the ability to cope with breaks occurring between CIVs. Should a break occur between the CIVs with a single failure of one of the CIVs, then containment bypass results. To preclude bypass, criteria are developed to ensure that the probability of a piping failure was sufficiently low to make it unlikely. The NuScale plant has its dual CIVs in a single valve body located directly outside of containment. Therefore, there are no break locations between the valves. However, the weld between the valve body and the CNV nozzle is equivalent to those to which break exclusion applies. Therefore, interpretation of the allowable extent of break exclusion would limit it to only a few welds in the NuScale design. NuScale has extended this break exclusion boundary outside the CNV slightly to include:

The weld at the outboard CIV nozzle.
The weld at the outboard check or excess flow check valve nozzle in RSC-connected lines.
DHRS piping welds outside the CNV.

Thus, the guidance of BTP 3-4 Paragraph B.A.(ii) is used in piping design to ensure that breaks and leakage cracks can be excluded in the containment penetration area. The BTP 3-3 nonmechanistic breaks of main steam and feedwater piping are also addressed. The remaining high-energy piping under the bioshield applies BTP 3-4 Paragraph B.A.(iii) for ruptures and (v) for leakage cracks. Figure 3-5 is a representation (not all lines are shown) of application of the NRC guidance on break location and size, as applied in the NPM bay and the RXB. © Copyright 2018 by NuScale Power, LLC 60

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Figure 3-5. Application of Nuclear Regulatory Commission break location guidance in the NuScale power module bay and the reactor building The NuScale design has notable differences from the larger LWRs for which BTP 3-4 was developed.

The length of piping and number of welds inside the NuScale CNV is limited and is less than for large LWR break exclusion zones. For the NuScale design, no primary or secondary piping, other than about 160 feet of DHRS piping, is within the break exclusion zone outside containment, compared to approximately 1500 feet of primary and secondary break exclusion zone piping in the AP1000.
The design pressure and temperature of MSS, FWS, and DHRS piping in the break exclusion zone is the same as for the RCS.

Break exclusion is not applied to the piping in the RXB. 3.6.1 Leakage Cracks Leakage cracks are excluded in containment penetration areas where the criteria of BTP 3-4 Paragraph B.A.(ii) are satisfied. Per BTP 3-4 Paragraph B.A.(v), leakage cracks are postulated unless specific criteria are met. For Class 2 piping, the acceptance criterion is for the calculated stress to not exceed 0.4 times the sum of stress limits given in Subarticles NC/ND-3635. Postulated leak locations are isolable by the CIVs and are small (about 2.7 in.2 in the MSS and 0.15 in.2 © Copyright 2018 by NuScale Power, LLC 61

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 for CVCS) if they were to occur. Larger leakage cracks are expected to be detectable by temperature monitoring under the bioshield. 3.7 Leak-Before-Break General Design Criterion 4 includes a provision that the dynamic effects associated with postulated pipe ruptures may be excluded from the design basis when analyses, reviewed and approved by the Commission, demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. This analysis is called LBB. The LBB concept is based on the ability to detect a leak in the piping components well before the onset of unstable crack growth. 3.7.1 Inside the Containment Vessel For the NuScale plant, the application of LBB is limited to the large bore ASME Class 2 SGS (i.e., MSS and FWS) piping inside the CNV. The piping analysis addresses cyclic transients and produces bounding loads for the ASME Class 2 piping with respect to LBB. Methods and criteria to evaluate LBB are consistent with the guidance in SRP 3.6.3 and NUREG-1061, Volume 3 (Reference 1.4.2.11). Potential degradation mechanisms are limited. The piping is stainless steel, uninsulated, and in a hot, dry, evacuated environment, precluding external corrosion during normal operation. LBB analysis methodology and results for MS and FWS piping is provided in FSAR Section 3.6.3. Application of LBB permits elimination of the dynamic external effects of postulated ruptures in high-energy piping; specifically (Reference 1.4.2.7):

blast effects
pipe whip
pipe break reaction forces
jet impingement forces
dynamic or non-static pressurization of cavities, compartments, or subcompartments (not performing a containment function)

Therefore, lines qualifying for LBB are evaluated only for leakage cracks and flooding effects. Because essential components inside the CNV are qualified for the containment design conditions of saturated steam at 1000 psia, flooding in an ECCS transient, and flooding during refueling, the effects of leakage from lines meeting LBB criteria are bounded and do not need to be explicitly analyzed. The methodology, criteria, and results of applying LBB are discussed in detail in NuScale FSAR Section 3.6.3. © Copyright 2018 by NuScale Power, LLC 62

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.7.2 In the NuScale Power Module Bay Leak-before-break is not applied to the piping in the NPM bay. 3.7.3 In the Reactor Building Leak-before-break is not applied to the piping in the RXB. 3.8 Separation Separation is a means of protecting essential SSC from the effects of HELBs and MELBs. The four degrees of separation applied are:

Isolation of essential SSC from high- and moderate-energy piping by placement in different compartments of the plant. An example of this is that components outside the CNV are isolated from rupture effects inside the CNV.
Separation by distance: if essential SSC are distant from the rupture location, it may be shown that there are no effects of blast, pipe whip, and jet impingement. However, pipe break reaction forces and environmental effects such as pressurization and flooding must be evaluated.
Separation through redundancy: multiple, distributed components exist such that a HELB can only affect a number such that, after postulation of a single active failure, necessary functionality remains available.
Separation by intervening obstacle; depending on the obstacle, missiles, blast, pipe whip, jet impingement, or flooding may be mitigated, but not pressurization.

o Pipe restraint: a restraint may limit the movement of a whipping pipe, keeping the pipe or jet from affecting essential SSC. o Plant structure or component: the plant design may include SSC that are large and robust enough to serve as a barrier to HELB effects (see Section 3.5). o Pipe whip barrier or jet shield: these are structural features added for the purpose of intercepting a whipping pipe or jet at specific rupture locations from striking an essential SSC. The NPM and RXB evaluated in this PRHA do not require any features the sole purpose of which is to serve as a pipe whip barrier or jet shield. 3.8.1 Inside the Containment Vessel Three degrees of separation are considered: compartmentalization, distance, and the presence of an intervening obstacle.

The CNV isolates essential components inside from HELB effects outside.
The CNV is a tall, narrow vessel. A pipe break at any given location has a limited reach for pipe whip and jet effects. For example, a whipping pipe or jet caused by a break at the inner CNV head does not affect an RRV about 50 feet below. An NPS 2 Schedule 160 pipe has an inner diameter of 1.687 inches, therefore, for steam

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 discharge, a penetration distance of 25 diameters is estimated to be 3.5 feet. However, this penetration distance is shown to be limited to about 2.2 diameters, or less than 4 inches (see Appendix E).

For some pipe break locations inside the CNV, large structures limit the range and direction of a whipping pipe or jet. Examples are the RPV, CNV, and SGS piping.

3.8.1.1 In the NuScale Power Module Bay No HELBs occur in the NPM bay area based on high-energy lines satisfying criteria of BTP 3.4 for excluding breaks. 3.8.1.2 In the Reactor Building Separation by placement in different subcompartments and redundancy are the degrees of separation considered. Although not essential components, PAM and DC power cables are routed in areas separated from high-energy lines by structural or shield walls. 3.9 Analysis Methodology Figure 3-1 is a flow chart of the process for identifying postulated rupture locations and vulnerable essential and safety-related targets through assessing the relevance and consequences of possible external effects.

Essential targets are identified (see Section 3.2).
High- and moderate-energy systems are identified.
Each of the three regions of the plant (the CNV, the NPM bay under the bioshield, and the RXB) is considered separately.
If potential HELB locations satisfy break exclusion (in CNV or NPM bay) or LBB (MSS and FWS piping in CNV) acceptance criteria, then consideration of HELB dynamic effects is avoided.
For postulated breaks of piping containing steam, the potential for creation of a blast wave is assessed (see Section 3.9.3).
Availability of energy sufficient to cause pipe whip is evaluated (see Appendix C).
If pipe whip is possible, then the vulnerability of essential SSC to being hit is determined based on direction of pipe whip and distance.
If pipe whip impact is possible, the impact of the impact is assessed.
Jet ZOI is defined to determine if any essential SSC are within it (see Appendix E).
For essential SSC within the ZOI, the jet impingement effects are assessed (see Appendix E).
The pressurization effect of the postulated HELB is quantified (see Appendix D).
The consequences of flooding are determined FSAR Section 3.4.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.9.1 Determining Break Locations As described in Section 3.6 above and FSAR Section 3.6.3, potential break locations are evaluated for ability to satisfy BTP 3-4 or LBB criteria. If one of those sets of criteria is met, then HELB effects need not be further evaluated. As a result of break exclusion and LBB evaluations in accordance to regulatory guidance, breaks are postulated only at terminal ends of NPS 2 piping inside the CNV and no locations in the NPM bay under the bioshield. In the RXB, ruptures are assumed to occur anywhere that high-energy piping is present. 3.9.2 Parameters Affecting Severity of High-Energy Line Break Effects The parameters that determine the severity of HELB and MELB effects are:

Thermodynamic conditions of the system before the break occurs (see Table 3-3) -

higher energy fluid generally causes larger magnitude effects. The initial fluid condition in the pipe before rupture is that for normal full power (102 percent thermal) operating conditions for that pipe segment. MSS and FWS lines in the CNV are excluded from rupture by LBB, so hot standby conditions are not considered. This fluid energy in the blowdown is consumed by several phenomena: failing of the material in order to create the rupture opening, accelerating the fluid out the break, irreversible losses, counteracting spray in opposite directions, bending the pipe at its plastic hinge, and accelerating the end of the pipe in a circumferential offset break. None of these are credited in removing energy from the blowdown, except for pipe whip screening.

Size of the pipe that breaks - NuScale piping serving a given function (i.e., feedwater) is smaller than traditional LWRs. This reduces the severity and the range of effects.

For example, a main steam system NPS 38 line in the AP1000 has approximately 27 times the energy per foot of pipe than the NuScale NPS12 line.10

Location of the break (i.e., proximity to essential SSC, ambient conditions) -

o If the break is remote or separated from essential SSC, the effects are negligible. o The flow issues from a straight pipe section downstream of either a long pipe run or a nozzle connected to a reservoir (i.e., the RPV), involving flow resistance and entrance losses.

Break configuration - in accordance with regulatory guidance, assumptions are made (e.g., discharge coefficient of 1.0). The acceleration of a whipping pipe segment depends on the fluid jet thrust force. The blowdown from postulated ruptures provides a bound on dynamic effects.

o No credit is taken for the reality that the end of an actual break is ragged, likely with rough and bent edges that provide flow resistance. 10 AP1000 volume per pipe foot of 7.47 vs. 0.63 ft3; steam density in AP1000 of 1.84 versus 0.73 lbm/ft3; and specific enthalpy in AP1000 of 1198 versus 1290 BTU/lbm yields factor of 27.6. © Copyright 2018 by NuScale Power, LLC 65

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 o The break opens instantaneously, which is physically impossible but conservative. Regulatory guidance sets a maximum break opening time of one millisecond unless otherwise justified.

Duration of blowdown - in accordance with regulatory guidance, credit for reduction in upstream (source) pressure is only considered where justified (e.g., closure of FWS check valve). For estimating jet reaction and pipe whip, the blowdown is from an infinite reservoir at intact system normal operating conditions. This is assumed unless a check valve or normally closed isolation valve is available within a short distance to terminate flow. For subcompartment pressurization, blowdown can be terminated by valve closure after a single active failure or by depletion of the reservoir.
The thrust load acting on the pipe due to a blowdown jet is equal and opposite to the jet. The pipe may pivot at the nearest surface contact point or pipe restraint. In the case of a circumferential break, the force of the jet is directed along the axis. A nozzle does not deflect because of its rigidity, straightness, and short length. Jet thrust load is determined as described in Section 3.9.4.
Occurrence of pipe whip is screened by assessing if the jet thrust load is sufficient to form a plastic hinge, as described in Section 3.9.5 and Appendix C. Because most NuScale high-energy pipe is small diameter but heavy wall (i.e., schedule 160 or 120),

available energy compared to bending moment is less than large LWRs.

The kinetic energy of a whipping pipe is determined by the distance through which the jet thrust can cause it to move. The smaller scale of the NuScale design reduces the pipe size (hence, thrust) and distance of the whipping pipe.

3.9.3 Blast Effects As previously noted, the potential for a blast wave to occur depends on the surrounding environment. The timing of opening of the break and the initial, intact system thermodynamic conditions are also key. Although pipe rupture times of less than a millisecond are unlikely, break opening time is assumed to be instantaneous. Appendix F provides a detailed discussion of blast effects based on three-dimensional CFD modeling that reflects the postulated break characteristics and NuScale plant geometry. Key observations are:

A blast wave is weakly formed if the surrounding environment is at low pressure (less than 1 psia), as is the case inside the CNV. Buildup of pressure as blowdown progresses is not relevant because the blast wave is a prompt and short-lived phenomenon.
The severity of a blast depends on the amount of fluid that can escape within one millisecond of onset of the break because the blast wave forms within that time.

o The NuScale high-energy, steam-filled lines are relatively small, which limits the severity of the blast pressure. The energy available to form the blast is less than 27 times less than that of AP1000 (see prior Footnote 10).

Blast waves are not significant for subcooled discharge because liquid flashing occurs on timescales exceeding that of formation of a blast wave according to J. Lius Investigation on Energetics (Reference 1.4.3.7).

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

A blast wave has well-defined and interrelated characteristics. For example, its peak pressure and speed decrease with distance from its origin.
The pressure load applied by a blast wave is of short duration (i.e., an impulse load as shown in Figure 3-6) and does not apply uniformly across a large SSC at a given instant. Therefore, assuming the peak blast pressure is applied across the entire projected area of a component is inappropriate. The CFD analysis explicitly accounts for the time-varying pressure of the rapidly propagating blast wave.
Reflection off surfaces can reinforce the pressure load, requiring consideration of plant specific geometry. Angled or curved surfaces are loaded differently than a flat surface perpendicular to a line between the blast origin and surface. The pressures applied to surfaces by reflection can substantially exceed the incoming wave pressure. For this reason, use of representative plant geometry is necessary. The CFD analysis includes the interaction of incident and reflected waves with each other and nearby surfaces, including how the shape and orientation of surfaces affect reflection.
A small target has a lower peak pressure due to clearing, which is a phenomenon where some of the blast overpressure is relieved by bleeding off around the edge of the target. From Equation 2-8 of UFC 3-340-02 (Reference 1.4.3.15), clearing distance is equal to the height or half the width, whichever is the smallest, of the side of an object facing a blast wave. Because of both this pressure-relieving clearing and the short load duration as a supersonic blast wave moves over them, small structures are not limiting. The only SSC in the CNV or RXB that are large are the structures (e.g., CNV, RPV, and RXB walls and floors). The CFD analysis models clearing.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.9.4 Blowdown Thrust Loads If the energy in the pipe is insufficient to deform the piping, no whip occurs. If the line is isolated, no sustained jet occurs. 3.9.4.1 Thrust Coefficient The thrust coefficient is defined as the exit plane thrust force divided by the product of source pressure and exit plane flow area. In Reference 1.4.2.12, the NRC states the thrust force may be computed by calculating the force that must be exerted to hold in static equilibrium a plate positioned normal to the flow directly at the break point. The exit plane thrust force for a given break represents the maximum jet impingement force that can be delivered to target SSC. The time dependent thrust force includes the combined effects of the initial pulse, wave propagation and reflection, and the blowdown thrust from buildup of the discharge flow rate. ANSI/ANS 58.2 Appendix B discusses initial behavior of the jet before reaching steady-state. During the initiation of the jet, the peak of a decaying pressure oscillation exceeds the steady-state level that occurs once the blowdown stabilizes. Shock wave pressures in the low-pressure ambient in the CNV are low and their duration is about a millisecond (see Figure F-7 for an example), so this initial pulse is not significant. In the RXB, the jet is not assumed to expand with distance and a conservatively short distance between break exit and target SSC is assumed, eliminating the need to separately model an initial pulse. Therefore, just the total, steady state jet thrust force FT as given by SRP Section 3.6.2 needs to be evaluated:

                                           =                                      Eq. 3-1 where, Po = initial intact system pressure (psia),

AE = the pipe break exit area (in.2) (subscript E refers to break exit), and CT = the thrust coefficient (unitless). Values for CT depend on fluid conditions but otherwise are largely independent of plant design. Standard Review Plan Section 3.6.2 specifies that values should not be less than 1.26 for steam, saturated water, and steam-water mixtures and should be 2.0 for subcooled, nonflashing water. ANSI/ANS 58.2 identifies values of 1.26 to 1.30 for saturated and superheated steam. NuScale uses 1.26 for steam and two-phase jets, which meets the acceptance criteria of NRC guidance. No breaks in the CNV cause high pressure, liquid jets. During operation, CNV pressure is below 1 psia. For pipe ruptures in the CNV, PA varies with time, starting at less than 1 psia and rising for large leak rates (i.e., RRV opening). Because FT is maximized when PA is a minimum, CNV pressure is set to be 0 psia initially. © Copyright 2018 by NuScale Power, LLC 68

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Eq. 3-1 assumes that there is no substantial flow resistance to the discharge and that the upstream reservoir pressure is constant. The latter is generally true for periods of seconds or minutes, except for isolated lines (i.e., a break of the high point vent degasification line at the RPV head) for which the time span for depressurization is equivalent to that for opening of the break, thereby removing the jet thrust before the pipe can move. 3.9.4.2 Damage Potential Single-phase steam jets with upstream pressures of 1200 psia were found to cause damage to pipe insulation at a distance of up to 25 times the pipe exit diameter (i.e., L/D

        = 25). However, insulation is fragile as evident from Reference 1.4.2.10, which reports types of insulation suffering damage for impingement pressures as low as 4 psig.

NUREG/CR-6808 (Reference 1.4.3.14) Table 3-1 provides the impingement pressures found in testing that cause damage to various types of piping insulation used in US PWRs. The damage pressures range from 4 to 40 psi for fibrous insulation to a high of 190 psi for two types of reflective metal insulation. Insulation is more fragile than the solid metal surfaces of SSC inside the CNV. Therefore, jet impingement pressures need to be considerably above 190 psi to be of concern. Impingement loads are only meaningful for hard or relatively hard targets such as ECCS valve bodies, the CNV steel wall, and RXB concrete structure. Thus, impingement pressures must be substantial (above 190 psia) rather than the less than 4 psia needed to protect against dislodging insulation. As such, fewer uncertainties exist in predicting jet impingement effects on piping, and the relevant penetration distance is much shorter than 25 L/D. Jet impingement testing was performed on electrical cable in support of the AP1000 assessment of debris generation. The conclusion was that cables at 4 L/D from a jet simulating an AP1000 loss of coolant are not damaged. The results are given in terms of distance due to the difficulty in accurately measuring impingement pressure. The NRC staff agreed with the conclusion. In Reference 1.4.2.13, the ACRS also agreed, stating The recommended distance of four break diameters from a loss-of-coolant accident jet, at which unprotected cables would not be damaged, has been shown by testing to be sufficiently conservative to bound plant conditions with high likelihood. Although the focus of this testing did not include cable functionality, inspection of test target cables showed no damage at 4 L/D (with exception of one cable). The results are applicable only to the type of cables actually tested, but an AP1000 LOCA jet is considerably larger and higher energy than a NuScale NPS 2 HELB. Therefore, it is likely that even unprotected cable inside the CNV would survive jet impingement from an NPS 2 HELB provided its separation from the break exit exceeded 4 L/D, or 6.75 inches. NuScale cable to be used in the CNV is tested for survival under jet impingement. An overview of the NuScale vulnerability to jet impingement is:

Based on plant operating conditions and smaller size of piping, thrust loads for NuScale line breaks are a fraction of those normally encountered in large LWRs (i.e.,

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 a NuScale 12-inch MSS line has about five percent of the total thrust force of an AP1000 38-inch MSS line break11).

Main steam system HELB occurrence is limited to the RXB, because MSS breaks inside the CNV and in the NPM bay are eliminated by LBB and break exclusion, respectively. The NuScale full power steam pressure is 500 psia (AP1000 is 808 psia),

which is actually a higher enthalpy per pound mass. However, the NuScale MSS steam density is about half, the flow rate driven by the system to ambient differential pressure is about 80 percent, and the full break single-ended flow area is about 11 percent of those of AP1000. The combination of these differences would put NuScale main steam system HELB mass and energy transfer at about five percent compared to AP1000.

Damage to insulation on piping in the RXB is not a concern, and the essential SSC are located in other areas, leaving only building walls as a target SSC. Allowable impingement pressure on SSC is considerably higher than that in large PWRs where insulation stripping is relevant.
With a lower system pressure and more jet resistant target, a main steam system HELB penetration distance of 25 pipe diameters is an overestimate. In the RXB, NPS 12 (inner diameter of 10.75 in.) MSS line breaks are postulated: 25 L/D corresponds to 22.4 feet. Based upon postulated RXB arrangements (not yet being finalized) and possible break locations (not yet defined), jet impingement anywhere within the subcompartments was considered.
For HELBs inside the CNV, only piping of NPS 2 size (inner diameter of 1.687 in.) is susceptible. Presuming 25 L/D, the steam jet range would then be about 43 inches.

However, as shown in Appendix E, the jet pressure drops off rapidly with distance, even with a conservatively low spreading half-angle, such that the effective range of concern is less than 2.2 L/D (4 inches). For unprotected cable, 4 L/D (6.75 inches) is used.

Similarly, although NuScale SSC are packed more closely together, the lower CVCS and MSS pressure, smaller pipe size, lesser distance through which a whipping pipe can travel, and presence of robust structures that serve as pipe whip barriers make the damage potential of pipe whip impact considerably less than other LWRs.

3.9.5 Pipe Whip Loads In the CNV, pipe whip loads are limited because:

Only the NPS 2 locations that do not satisfy break exclusion are considered. These locations are limited to terminal ends. The opposite end of the break is a nozzle, which does not whip.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

The hinge point must be more than 14 inches (or more if the end of the break includes a piping length parallel to whipping motion; see C.4.1) lateral distance from the break to have sufficient energy to form a plastic hinge allowing whip to occur.
The congested arrangement and short piping lengths limit the whip distance and, thereby, limit the energy at impact.

The occurrence and consequences of pipe whip are determined as follows.

For piping meeting the criteria of break exclusion or LBB, pipe whip is not considered because dynamic effects of ruptures are excluded.
If the other end of a terminal end is a RPV or CNV nozzle, it does not whip because the nozzle is short, stiff, straight, and restrained by the component. Similarly, breaks are not postulated to occur in pump and valve bodies because the wall thickness exceeds that of connecting pipe.
The calculation of thrust and jet impingement forces considers no line restrictions (that is, a flow limiter) between the pressure source and break location, but does consider the absence of energy reservoirs, as applicable (e.g., the degasification pipe in the CNV is normally isolated).
If the jet thrust is insufficient to yield the pipe, then pipe whip at that break location is eliminated from further consideration, although jet impingement from a limited separation break is still relevant.
Where pipe ruptures are postulated to occur, the distance is determined from the break location to the nearest restraint that limits range or direction of pipe whip.
Pipe whip is considered to result in unrestrained motion of the pipe along a path governed by the hinge mechanism and the direction of the vector thrust of the break force. A maximum of a 90-degree rotation may take place about any hinge. Pipe whip occurs in the plane defined by the piping geometry and configuration and to initiate pipe movement in the direction of the jet reaction, as identified in BTP 3-3.

For postulated break locations remaining:

The reach of the whipping pipe is compared to the distance from the restraint to the nearest essential SSC and other high-energy lines (the line is not assumed to straighten out because the jet load is trying to compress the piping). If no target of concern is within reach, then pipe whip mitigation at that break location is not needed.

Even if a target is within range, pipe whip impact may be prevented by presence of an intervening SSC that is sufficiently robust to serve as a barrier in accordance with Section 3.5.1. o If the direction of the initial pipe movement caused by the thrust force is such that the whipping pipe impacts an essentially flat surface normal to its direction of travel, it is assumed that the pipe comes to rest against that surface, with no pipe whip in other directions. However, to account of the potential rebound upon impact, a rebound force of 10 percent is added to the impact load.

The loading that results from a break in piping is determined as described in Appendix C.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.9.5.1 Screening for Occurrence of Pipe Whip The method to determine if pipe whip occurs is based on calculation of the minimum internal forces necessary to form a plastic hinge in the pipe, which depends on the thrust at the break exit plane, the strength of the pipe to resist bending, and the distance to the near pipe whip restraint. Jet thrust loads are calculated for applicable breaks using Eq. 3-1 (above) with a CT of 1.26, consistent with SRP 3.6.2. Piping force is estimated by determining the projected length Lh from the pipe break to the nearest pipe whip restraint in the plane perpendicular to the plane of motion. The HELB jet loads applied on the end of the Lh long moment arm are assessed against the plastic moment for the pipe. The methodology assumes the thrust force remains constant, except for an isolated line with a limited length of pressurized piping such as the degasification line, which has insufficient mass and energy to whip. 3.9.5.2 Impingement Pressure The maximum force applied to an impingement target is determined using Eq. 3-1. The only breaks inside the CNV are NPS 2 CVCS and DHRS lines. The limiting break in the RXB is an MSS line. The pressures for these two breaks at the break exit plane are as shown in Table 3-7 and include a factor of 1.26 for the thrust coefficient CT. These values are upper limits for the downstream pressures for real breaks where pressure across the jet drops off as the jet expands and velocity of the jet is reduced by occurrence of turbulence leading to irreversible conversion of kinetic energy to heat. The isentropic expansion of steam jets is discussed in Appendix E. Table 3-7. Break exit plane parameters CVCS Break* MSS Break Inner diameter (in.) (( Intact system pressure (psia) Intact system temperature (ºF) Break exit plane pressure (includes CT of 1.26)(psia) Break exit plane area (in.2) Maximum impingement force (lbf) }}2(a),(c)

DHRS breaks are assumed equivalent although internal pressure is only about 500 psia 3.9.6 Jet Zone of Influence Three types of breaks are considered per regulatory guidance: 1) circumferential breaks with full axial or sideways separation of pipe ends, 2) circumferential breaks with limited separation, and 3) longitudinal breaks. In addition, there are three thermodynamic blowdown conditions: 1) liquid, 2) two-phase, and 3) steam that have different behavior, as described in Appendix E.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 High-energy line breaks are under-expanded when they issue from the end of the break, because the pipe section immediately upstream confines the flow radially. High-energy line breaks expand rapidly into the surrounding medium, with the expansion limited by jet momentum and increasing pressure at the boundary of the jet with the surrounding medium. In the limit, for a slow leak, the discharged fluid disperses uniformly in all directions. The expansion has the effect of reducing the jet pressure at a target below that at the break exit. ANSI/ANS 58.2 provides guidance on this expansion, but the NRC has expressed concern that this guidance is not generally applicable (see SRP Section 3.6.2). Considerable effort has gone into evaluating the jet plume appropriate for HELBs. ANSI/ANS 58.2 presents the modified Moody model in which the conical jet expands at a 45-degree half-angle for a downstream distance of 5 L/DE and at 10 degrees from there on. Some evaluations recommend a hemispherical or even a spherical ZOI. The advantage of wider ZOIs is that they cover repositioning of the pipe exit due to whip and occurrence of redirection. The NuScale approach is to overestimate the extent of the ZOI while underestimating the effect of jet expansion on reducing the pressure on downstream SSC, although these are mutually exclusive. For NuScale, the acceptability of jet impingement pressures is insensitive to the analytical approach. Because piping inside the NuScale CNV is not insulated, the use of non-metallic material inside the CNV is minimized. Most cable is protected by being routed out of range. The RRV intake is directed downward and submerged during ECCS recirculation. Therefore, jet impingement does not present a risk of generating debris capable of blocking ECCS recirculation. Although piping outside the CNV is insulated, insulation stripping presents no hazard to safety-related functions. 3.9.6.1 Inside the Containment Vessel For breaks inside the CNV, expansion of the jet into the low-pressure surroundings results in different behavior than is usually experienced for HELBs. Wider jet spreading occurs because the initially low air density of a CNV pressure below 1 psia removes most of the resistance to jet expansion. The wider jet expands the ZOI but reduces the pressure and the penetration length, because the mass and energy of the jet is more widely dispersed. Although pressure within the CNV increases with time, the pre-existing wide expansion of the jet persists as the jet is already established. The CFD blast modeling discussed in Appendix F shows that a steam jet initially develops a spreading half-angle greater than 60 degrees. Appendix E provides a detailed discussion of the jet modeling applied in the CNV. ((

                                                                  }}2(a),(c)

For two-phase jets, the methodology of NUREG/CR-2913 is applied to determine the jet pressure distribution versus distance from the break exit. This is discussed in Appendix E. © Copyright 2018 by NuScale Power, LLC 73

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Based on the preceding discussion that pressures of at least 190 psi are acceptable for hard components and on the pressure vs. distance behavior for steam and two-phase jets, a distance of slightly more than four inches (2.2 L/D) is sufficient to provide acceptable protection of metal SSC. This distance is sufficiently short that few SSC are within range. For unprotected cable, 6.75 inches (4 L/D) is sufficient. In summary, any SSC in the CNV more than four inches axially or radially (6.75 inches for unprotected cable) requires no further evaluation for jet impingement loading. 3.9.6.2 In the Reactor Building In the RXB, normal atmospheric pressure surrounds any postulated break location, and the venting available limits the buildup of backpressure. No breaks are postulated in the NPM bay. Because piping arrangements are not yet finalized elsewhere in the RXB (COL Items 3.6-1, 3.6-2, and 3.6-3), specifying a particular ZOI is not meaningful. The RXB walls, floors, and ceilings are assumed to be a minimum of two pipe internal diameters from a break exit. This is judged to be closest practical distance from a wall at which to make a field weld (e.g., 22.5 inches for an NPS 12 MSS pipe, 3.5 inches for NPS 2 CVCS pipe). It is also reasonable as the minimum separation necessary to clear the other end of the broken pipe to impinge on an RXB structural surface or another component. Because the evaluation is performed assuming a distance of only 2 L/D, penetration length is not relevant. To focus the impingement pressure, no expansion is considered. For an MSS break, the exit plane pressure is approximately (( }}2(a),(c) from Eq. 3-1. This is small compared to a specified minimum concrete compressive strength of at least 5000 psia. No damage to concrete structure would therefore occur. 3.9.7 Jet Impingement Loads The load on an object exposed to a jet depends on the pressure of the jet upon the objects surface, on the intersection of the jet with the object, and on the shape of the object. To take credit for a limited intersection of the jet with the object, the break exit location must be known. Jet pressure at the nearest target surface is determined, including the thrust coefficient CT (1.26 for steam and two phase jets). If less than 190 psia (or beyond 4 L/D for unprotected cable)12, the impingement pressure is low enough to be non-damaging, and no further analysis is needed. 3.9.7.1 Dynamic Amplification and Resonance Experiments simulating HELBs routinely evince oscillations but not resonance. For dynamic amplification and resonance to occur, a number of criteria must be met, as discussed in Appendix B. These criteria are based on the research referenced in SRP Section 3.6.2 and similar work that identified the physical phenomena leading to 12 No additional factors need be applied because these criteria are based on testing. © Copyright 2018 by NuScale Power, LLC 74

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 resonance. The processes at work during a HELB have fundamental differences from those that occur in a jet with dry, noncondensable gas issuing from a smooth, fixed nozzle. These physical differences involve instability of the discharge, irregular discharge geometry, phase changes that suppress pressure changes, misalignment of jet and impingement target surfaces preventing establishment of a feedback loop, lack of an appropriately flat surface within a sufficiently close distance, and etc. If any one of these criteria is not met, a resonance is unlikely. In an HELB, none of the criteria is satisfied, precluding formation of a resonance. 3.9.8 Pressurization Caused by High-Energy Line Breaks In locations where HELB dynamic effects are not obviated by satisfying break exclusion or LBB criteria, the pressurization transient resulting from the mass and energy (M&E) release to the surrounding volume(s) has been analyzed. As additional M&E is introduced into the surroundings, it increases pressure and temperature. Pressure continues to rise until cooling of the enclosed volume (i.e., the CNV) or venting (e.g., RXB) of the volume is sufficient to offset the blowdown. Inside the CNV, postulated HELB locations involve blowdown from an RCS- or DHRS-connected NPS 2 pipe. The M&E release for these HELBs is less than 10 percent of that from an ECCS initiation that serves as the design basis for the CNV and for environmental qualification of safety-related and essential equipment in the CNV. Therefore, a separate environmental evaluation of HELBs inside the CNV is not performed. In the NPM bay, no postulated HELB locations require evaluation because piping satisfies break exclusion criteria (nonmechanistic breaks of MSS and FWS piping are discussed in Section 3.5.2.5). For HELBs in the RXB, the concern is room pressurization that challenges the structural integrity of the building, due to the combination of the pressure load with other loads (e.g., seismic, deadweight). Based on an assessment of RXB structural capability that considers combination with other loads (i.e., deadweight, structural, etc.), a pressure load that can be sustained on walls, floors, and ceiling of rooms housing high-energy piping is at least 3 psid. Detailed design of RXB piping arrangements is a COL applicant item. To ensure that RXB design is satisfactory for any allowable arrangements, the following approach was used:

Reactor Building design criteria and MPS specifications provide that o The areas through which Class 1E, associated circuits, augmented design circuits and associated-ADC circuits are routed and in which equipment is located are reviewed for potential hazards such as high-energy piping. Separation commensurate with the damage potential of the hazard is provided through the use of features such as separate rooms.

o Pipe failure hazard areas contain piping normally operating at high or moderate energies. For moderate-energy piping, pipe whip and jet impingement need not be considered; however, the wetting and environmental effects must be considered. © Copyright 2018 by NuScale Power, LLC 75

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Protection of nonhazard and limited hazard areas from pipe failure hazard areas are accomplished by the use of barriers, restraints, separation distance, or appropriate combination thereof. o The routing of Class 1E, associated circuits, augmented design circuits, associated-ADC circuits, or raceways in pipe failure hazard areas conforms to specific requirements, unless it can be demonstrated that pipe failure cannot prevent the Class 1E circuits and equipment from performing their safety-related function, and it can be demonstrated that pipe failure cannot prevent the applicable augmented design circuits and equipment from performing their important-to-safety function. Separation criteria depend on the qualification of piping for design basis events, the division(s) it is in and co-located with, and the need for protective action. o Class 1E circuits are routed or protected so that failure of the mechanical equipment of one division cannot disable Class 1E circuits or equipment essential to the performance of the safety-related function of the redundant division(s). The effects of pipe whip, jet impingement, water spray, flooding, radiation, pressurization, elevated temperature, or humidity on redundant electrical systems caused by failure, misoperation, or operation of mechanical systems are considered. o The MPS fails into a safe state or into a state demonstrated to be acceptable on some other defined basis, if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. For practical purposes, the above require that instrument cables used for the MPS or PAM and DC power systems and cables not be located in the same rooms as high-energy piping or through which venting occurs, with the exception of the NPM bay where HELBs are excluded. To bound postulated HELBs anywhere in the RXB, the following scenarios are evaluated, using a GOTHIC model of relevant parts of the building.

A full shear of a NPS 12 MSS pipe of one NPM is postulated to occur in the pipe gallery. Blowdown occurs from both ends of the pipe. Whip of the MSS pipe fails either a NPS 4 MSS bypass line or an NPS 8 FWS pipe (pipe whip of an MS pipe into another MS pipe does not cause a second rupture, in accordance with regulatory guidance),

adding to the M&E release.

A double-ended shear of a high temperature and pressure section of the CVCS discharge line is postulated to occur in the smallest room through which it passes. The highest energy lines in the CVCS system in the RXB are the hot side inlet (i.e.,

discharge) and outlet of the non-regenerative heat exchanger, at maximums of 1840 psia and 500-degrees Fahrenheit and of 1960 psia and 453-degrees Fahrenheit, respectively. The higher temperature of the discharge line results in a greater room pressurization. © Copyright 2018 by NuScale Power, LLC 76

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3.9.8.1 Venting Venting inside the CNV is not required and contrary to the requirements for containment. For rooms in the RXB with postulated HELB locations, GOTHIC analysis is used to determine the vent opening area necessary to limit pressurization to less than the RXB structural limit (Appendix D). Where a room vents into another room, the downstream room also needs its own vent area until venting into the pipe gallery or pool area occurs. Room venting is via existing openings or by door failure. The vent paths are not dependent on the RXB ventilation system. 3.9.9 Effects of Leakage Cracks Leakage cracks are considered in high-energy systems not satisfying break exclusion criteria and in moderate-energy systems. The consequences of a leakage crack in a moderate-energy system are limited to slow, gravity flow of liquid water. Leakage cracks in high-energy systems could release small quantities of steam. 3.9.9.1 Inside the Containment Vessel The effects of leakage cracks are bounded by evaluations of HELBs and ECCS initiation. The SGS lines are shown to meet LBB based on the ability to detect leakage cracks. Very small leak rates can be detected by monitoring CNV pressure and containment evacuation system (CES) sample vessel level. Operation of the CES vacuum pump prevents a continued accumulation of water or steam in the CNV. 3.9.9.2 In the NuScale Power Module Bay High-energy piping in the containment penetration area satisfies break exclusion criteria out to and including the welds attaching the outermost valve body nozzle to the piping section that ends in the spool flange, eliminating the need to evaluate effects of ruptures or leakage cracks, except for those of nonmechanistic breaks of MSS and FWS lines (as discussed in Section 3.5.4.2). Beyond that point and to the first weld on the pipe gallery side of the pool wall (including the piping spools), the piping satisfies the BTP 3-4 paragraph B.A.(iii)(2)(b)(ii) criteria for no breaks at intermediate locations. Leakage crack environmental effects are bounded by those of nonmechanistic breaks of MSS piping in the vicinity. Safety-related and essential SSC (e.g., CIVs, I&C) are qualified for pressure and temperature conditions resulting from leakage cracks. The one moderate-energy system is RCCWS (the CFDS is normally isolated). With an operating temperature below boiling, no pressure or temperature increase in the NPM bay occurs. Continued leakage is detected by a loss of expansion tank level before dilution of borated pool water could occur. 3.9.9.3 In the Reactor Building The effect of leakage cracks in high- and moderate-energy systems is bounded by the HELB pressurization analyses and RXB flooding analysis. © Copyright 2018 by NuScale Power, LLC 77

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 4.0 Results This section discusses postulated HELBs and MELBs and their external effects in the NuScale plant. The effects are based on the methodology discussed in Section 3.0. Because of the different conditions and systems in the CNV, the NPM bay under the bioshield, and the RXB, results are subdivided accordingly. 4.1 Postulated Break Locations 4.1.1 In the Containment Vessel MSS and FWS piping satisfy LBB criteria. Breaks are postulated at terminal ends of RCS-connected pipes and DHRS pipes, as identified in Table 3-4. Intermediate locations satisfy BTP 3-4 paragraph B.A.(iii) criteria or are evaluated for external effects. Longitudinal breaks are not considered because this piping is NPS 2. Leakage crack effects are bounded by those of ECCS initiation and ruptures, and leakage detection in the CNV (i.e., CNV pressure and CES sample tank level) is capable of identifying very small leaks. Appendix A provides details of piping evaluation against BTP 3-4 criteria. 4.1.2 In the NuScale Power Module Bay No breaks or cracks occur in the containment penetration area (other than nonmechanistic breaks in the MSS and FWS), based on application of the break exclusion criteria of BTP 3-4. The containment penetration area extends from the CNV nozzle-to-pipe weld to the outermost CIV or check valve-to-pipe weld, allowing application of the criteria of BTP 3-4 Paragraph B.A.(ii) to exclude breaks at terminal ends. Ruptures at intermediate locations are excluded because the design satisfies BTP 3-4 Paragraph B.A.(iii). Appendix A provides details of piping evaluation against BTP 3-4 criteria. Piping under the bioshield beyond the containment penetration area is not subject to rupture as it is designed to meet BTP 3-4 Paragraph B.A.(iii) criteria, and effects of leakage cracks are bounded by nonmechanistic breaks of MSS and FWS piping considered in the contiguous containment penetration area. 4.1.3 In the Reactor Building Breaks are postulated in any location where high- or moderate-energy MSS, FWS, or CVCS piping is located because piping arrangements are not finalized. Therefore, no piping stress calculations are needed. Effects of longitudinal breaks in main steam system and FWS piping are bounded by this approach. 4.2 Blast Effects Blast effects results are based on three-dimensional CFD analysis discussed in Appendix F. © Copyright 2018 by NuScale Power, LLC 78

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 4.2.1 In the Containment Vessel Because only NPS 2 lines are postulated to break, the M&E release feeding the blast formation is small. Only the degasification line has the potential for forming a blast, because the other CVCS lines contain subcooled liquid. The magnitude of the blast wave pressures is low, and the maximum force imposed on any component is limited to 6,000 lbf. In addition, the load is of very short duration, a few milliseconds. 4.2.2 In the NuScale Power Module Bay Not applicable. 4.2.3 In the Reactor Building Breaks are postulated in MSS lines at three locations in a pipe gallery. Only MSS lines have a potential for forming a blast, because the other CVCS lines contain subcooled liquid. The maximum force on any component is less than 10,000 lbf. Although a force of 103,000 lbf on the pool wall was calculated, it is distributed over a surface area with a radius of about 100 inches, yielding a momentary overpressure of less than 15 psig. No damage occurs as a result, and the shortness of the loading eliminates the need to consider it in load combinations. 4.3 Pipe Whip Results of pipe whip evaluations are detailed in Appendix C. 4.3.1 In the Containment Vessel Pipe whip for breaks at the RPV and CNV terminals ends has been evaluated. The nozzle end does not whip. For the piping end, the motion of the pipe is such that no safety-related or essential SSC are impacted. Even if an impact did occur, the SSC are of heavy wall construction so that they neither leak nor crack. There is one exception: the ECCS trip/reset line. If a whipping pipe strikes a trip/reset line, the line is severed, causing it to vent. This has the same effect as opening the trip valve and allows the ECCS main valve to open once the IAB clears. As the response to the HELB is ECCS initiation, the severance of a trip/reset line has no effect on response to the event. 4.3.2 In the NuScale Power Module Bay Not applicable. 4.3.3 In the Reactor Building Because of their higher internal energy and longer whip arc possible in the pipe gallery, MSS breaks are limiting. To bound future piping arrangements, a large pipe whip arc was evaluated. Penetration of the concrete was minor, and the pool and main structural walls are sufficiently thick to avoid spalling. © Copyright 2018 by NuScale Power, LLC 79

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 4.4 Jet Impingement The small diameter piping in the NuScale plant yields small impingement forces. Testing previously performed as part of GSI-191 research showed reflective metal insulation could withstand at least an impingement pressure of 190 psi. Because stripping of insulation is not a concern in the NuScale design, safety-related and essential components are considered acceptable for jet impingement pressures of 190 psi. Based on industry available cable insulation testing, unprotected cables are considered acceptable if at least 4 L/D from an HELB exit, which is confirmed by testing of NuScale cable. 4.4.1 In the Containment Vessel Only NPS 2 CVCS and DHRS breaks at terminal ends are considered. Pressure of the RCS jet is below 190 psi within 2.2 L/D, or 4 inches for CVCS and within 1 L/D for DHRS. Damage to cables for separation group B & C is a concern, because MPS functionality is satisfactory for loss of signal from one channel (see Section 3.2.2) but PAM functionality requires that both group B & C signals not be lost to HELB effects. The cable separation distance of 4 L/D corresponds to about 6.75 inches. No damage from jet impingement occurs because cables are routed more than 6.75 inches away from postulated breaks. 4.4.2 In the NuScale Power Module Bay Not applicable. 4.4.3 In the Reactor Building The RXB concrete structure is evaluated for the effects of HELBs and MELBs. For effects on concrete, MSS breaks are limiting and are assumed to occur within 2 L/D of a wall, with no reduction in jet pressure with distance from the break. The maximum force of the jet and its maximum pressure is that at the break exit, or ((

                     }}2(a),(c), which is well within the minimum 5000 psi compressive strength of the concrete making up the five-foot thick wall. In addition, the effect of erosion is negligible.

4.5 Subcompartment pressurization 4.5.1 In the Containment Vessel Structures, systems, and components within the CNV are designed and qualified for ECCS initiation (design pressure 1000 psia). Therefore, the effects of the NPS 2 high-energy line breaks are bounded and do not require further evaluation. 4.5.2 In the NuScale Power Module Bay [LATER] © Copyright 2018 by NuScale Power, LLC 80

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 5.0 Conclusions This report documents the methodology and results of evaluations performed to determine postulated break locations and the effects of those breaks. The NRC guidance on relevant effects is identified and how differences in the NuScale design affect application of that guidance. The NuScale design is a compact, integral reactor than relies on passive safety features to ensure safe shutdown and cooldown for design basis events. The absence of large diameter RCS piping and active safety systems leads to a minimal number of safety-related and essential SSC. Examples of key features include

No operator action or electric power is required for safe shutdown and cooldown for design basis accidents.
Absence of essential and safety-related SSC in the RXB in areas containing high- or moderate-energy piping.
Small-volume metal containment operated a low pressure and with sensitive leak detection capability.
No insulation used inside the CNV, therefore there is no concern for dislodged piping insulation blocking core cooling.
Greatly reduced energy of blast, pipe whip, and jet impingement effects due to smaller plant size and lower energy system conditions.
Stainless steel primary and secondary piping within containment and areas where break exclusion is applied.
Ready access for inspection.

Application of the criteria for break exclusion and LBB leaves few locations in the CNV and none in the NPM bay requiring evaluation of external effects of blast waves, pipe whip, jet impingement, subcompartment pressurization, elevated temperature, and flooding. Protection is demonstrated through separation and by virtue of the robustness and qualification of safety-related and essential SSC. Evaluation of bounding HELBs and MELBs demonstrates the RXB structure is capable of withstanding the external effects of HELBs and providing separation from PAM instrument lines and DC electric power. External effects of HELBs and MELBs in the NuScale plant do not adversely affect the ability to shut down and maintain core cooling of the NPM. The following table summarizes the evaluations and results of this report. © Copyright 2018 by NuScale Power, LLC 82

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table 5-1. Summary of approach and result for line break assessment by plant area Inside Containment NPM Module Bay Reactor Building Limits on Containment penetration occurrence of area (all lines out to and breaks RCS: Break exclusion at including outboard valve intermediate locations body to pipe weld): satisfy BTP 3-4 B.A.(iii)(2) None BTP 3-4 B.A.(ii) MSS & FWS: LBB Rest of high-energy lines satisfy BTP 3-4 B.A.(iii)(2) Postulated Any high- or moderate-energy break RCS terminal ends None part of MSS, FWS, or CVCS locations Blast effects Negligible as confirmed by Evaluated bounding MSS Not applicable 3D CFD cases by 3D CFD Pipe whip Insufficient energy to whip, Protection by showing separation, or separation Not applicable separation or acceptable sufficient to avoid impact consequences Jet Pressure of jet: impingement Steam: NUREG-2913 Analysis of impact on building 2-phase: 30to 5 L/D & 10 Not applicable structure assumes no Hemispherical ZOI: pressure reduction. 2.1 L/D for pipe 4 L/D for unprotected cable Dynamic Does not occur amplification Pressurization Bounded by ECCS initiation Not applicable Evaluated bounding HELBs Flooding Not applicable (discharge Bounded by existing analyses Bounded by ECCS initiation goes into pool) in FSAR Section 3.4. Leakage Environmental effects cracks Bounded by ECCS initiation determined and used in Bounded by HELBs equipment qualification © Copyright 2018 by NuScale Power, LLC 83

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Appendix B. Dynamic Amplification and Potential for Resonance B.1 Background ((

                                                                                       }}2(a),(c)

B.2 Necessary Conditions for Resonance ((

                                                                                           }}2(a),(c) 13 ((
                      }}2(a),(c)

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                    }}2(a),(c)

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          }}2(a),(c)
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Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                      }}2(a),(c) 14 ((                                  }}2(a),(c)

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                                                           }}2(a),(c) 15 ((                                                              }}2(a),(c)

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  © Copyright 2018 by NuScale Power, LLC 92

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Appendix C. Pipe Whip This appendix provides the detailed methodology for pipe whip evaluation and results of the evaluations of pipe whip in the CNV and RXB. The methodology includes determination of whether a pipe has sufficient energy to whip, whether a whipping pipe can actually contact a safety-significant target, whether the target is sufficiently robust to withstand the impact, and the consequences of an impact should the previous steps not obviate the possibility of damage. The thrust force caused by the release of fluid from a circumferential break of a high-energy piping system may result in pipe whip, causing the piping to rotate about a plastic hinge point (e.g., pipe restraint, pipe anchor point) and possibly impact nearby SSC. Inside the CNV, the largest pipe size subject to HELB conditions is NPS 2, and target SSC are robust (i.e., RVVs). Other pipe sizes above NPS 2 have been qualified for LBB inside the CNV. Within the NPM bay, piping satisfies criteria of BTP 3-4 to conclude that no breaks occur and does not need to be evaluated for whip. In the RXB outside the bioshield, MSS, FWS, and CVCS lines are subject to a postulated HELB, but the only SSC requiring protection are evaluated. C.1 Considerations for Evaluating Pipe Whip As noted in Section 3.9.5, the following considerations apply to evaluation of pipe whip:

For piping meeting the criteria of break exclusion or LBB, pipe whip is not considered because dynamic effects of ruptures are excluded.
If the end is a RPV or CNV nozzle, it does not whip because the nozzle is short, stiff, straight, and restrained by the component.
In accordance with SRP Section 3.6.2, a pipe struck by another pipe of equal or smaller diameter and schedule (i.e., wall thickness) does not break or crack. In the CNV where HELBs are limited to NPS 2 Schedule 160 pipe, the RPV, CNV, ECCS valve bodies, and CRDMs are equivalent to larger, thicker walled pipe and, therefore, do not crack or break. This is discussed further in Section C.2.
Where pipe ruptures are postulated to occur, the distance is determined from the break location to the nearest restraint that limits range or direction of pipe whip.
The jet thrust necessary to cause pipe whip is determined. The calculation of thrust and jet impingement forces consider no line restrictions (e.g., a flow limiter) between the pressure source and break location, but does consider the absence of energy reservoirs, as applicable (e.g., the degasification pipe in the CNV is normally isolated).
If the jet thrust is insufficient to yield the pipe, then pipe whip at that break location is eliminated from further consideration, although jet impingement from a limited separation break is still relevant.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

Pipe whip is considered to result in unrestrained motion of the pipe along a path governed by the hinge mechanism and the direction of the vector thrust of the break force. A maximum of a 90-degree rotation may take place about any hinge. Pipe whip occurs in the plane defined by the piping geometry and configuration and in the direction of the jet reaction, as identified in BTP 3-3.

C.2 Pipe Whip Impact Inside the Containment Vessel Section 3.5.1.1 discusses the barriers presented by the RPV, CNV, and CRDMs. These, in addition to the ECCS main valve bodies, are robust structures with equivalent wall thicknesses considerably in excess of the NPS 2 Schedule 160 pipe that may whip inside the CNV. Similar to Table 3-5, Table C-1 provides a comparison of the safety-related and essential SSC size and wall thickness to those of the potentially whipping pipe. Figure C-1 is a visual representation of the information provided in the table. Table C-1. Comparison of sizes of whipping pipe to potential targets for high-energy line breaks in the containment vessel Component Pipe Size Outer Diameter (in.) Wall Thickness (in.) Whipping pipe RCS lines NPS 2 Schedule 160 (( SSC CNV N/A RPV N/A CRDM latch housing N/A lower section ECCS main valve N/A

                                                                                                         }}2(a),(c) body a without cladding b varies with vertical location; minimum value in range of pipe break locations shown c minimum in RXB areas containing high-energy piping within range of a whipping pipe d scaled from drawing

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Figure C-1. Visual scale comparison of NPS 2 Sch. 160 pipe to SSC wall thickness In view of the SRP 3.6.2 provision for impact of a pipe on like-size or larger pipe, the RPV, CNV, CRDMs, and ECCS valve bodies experience neither rupture nor crack if struck by a whipping NPS 2 Schedule 160 pipe in the CNV. Because of the large disparity in the thickness of the walls, much of the whipping pipe kinetic energy would be absorbed in crushing of the pipe itself. Regardless, functionality of components with moving parts (i.e., CRDMs and ECCS valves) following impact must still be addressed. Postulated break locations are at the RPV (head for spray and degasification lines and side wall for injection and discharge lines) and CNV heads. © Copyright 2018 by NuScale Power, LLC 95

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 C.2.1 Breaks at Reactor Pressure Vessel The degasification line does not whip for a break at the RPV head because the isolated line is filled with steam that immediately depressurizes as the break begins to open. For the spray line breaks at the RPV head:

Reactor vent valves - pressurizer spray line breaks at the RPV head do not result in pipe impact on an RVV because of the direction of pipe motion and intervening barriers. The two branches of the spray line are close to symmetrical. The image on the left side of Figure C-2 shows the arrangement on one side with the spray line shaded grey. If a break were to occur at the nozzle-to-pipe weld (green oval), the jet thrust force pushes it up and away from the only nearby RVV (dashed green oval). In addition, the steam and feed pipes block movement toward the RVV. Further, breaks at other than the immediately adjacent pipe penetration are blocked by the barrier presented by the CRDMs. Even should an RVV be within range of pipe whip, the RVV functionality is not impaired:

o The center-to-center spacing of RPV head penetrations for RVVs to potentially whipping pipes is at most 19 inches, but the RVVs are about 12 inches in diameter and the pipe is 2.375 inches outer diameter, reducing the separation to less than 12 inches. In this short distance, the whipping pipe gains only a small amount of kinetic energy. o Reactor vent valves in the closed position are held shut by primary pressure, with discs not free to move. The valves are qualified to be functional following seismic accelerations. o Less sturdy components on the outside of the RVV are the IAB and the position indicator housing. The IAB is out of reach of an RCS line whip because it is underneath the main valve body. If the position indicator housing were to be dislodged, the indicator stem might be broken off. This would eliminate indication (not a safety-related function) and might slightly increase frictional resistance to opening if the stem stub rubbed when the disc opened, but ECCS performance is not dependent on the speed of valve opening. o The trip/reset line runs from the arming valve to the CNV wall. A whipping pipe might pass through the trip/reset line, which is small diameter tubing. If the line is severed, it is equivalent to opening the trip valve: the RVV opens once RCS pressure drops below the IAB setpoint. The only concern is if the trip/reset line crimped completely shut on the valve side, preventing depressurization of the main valve control chamber. Crimping could only occur if the whipping pipe uniformly slammed the trip/reset line against a smooth surface, and is unlikely even then. Given the direction of motion of pipe whip at the RPV head, the tubing would break apart rather than crimp. Therefore, pipe whip on the RVVs does not impair RVV functionality.

Reactor recirculation valves - the RRVs are located out of range of the injection and discharge line terminal ends. The image on the right side of Figure C-2 shows the

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 separation of the discharge line (light blue rectangle) from the closest RRV (dashed light blue square). The injection line to RRV separation is farther. Thus, an injection or discharge line break does not impact an RRV.

Control rod drive mechanisms - Referring to Figure C-2, the motion of the spray line is almost parallel to the CRDMs with impact limited to a glancing blow. The center-to-center spacing of RPV head penetrations for RVVs and pipes is at most 17 inches, but the CRDM coil assemblies are about 12 inches across and the pipe is 2.375 inches outer diameter, reducing the separation to about 12 inches, sufficient to preclude contact before the pipe stops against the steam pipe above it. Also, the whipping pipe gains only a small amount of kinetic energy. If any impact were to occur, it would be on the coil assembly surrounding the upper latch housing, so that negligible energy is transmitted into the CRDM internals. In view that the CRDMs are qualified to function after exposure to seismic accelerations, scram functionality is not impaired by pipe whip impact.

C.2.1.1 Breaks at Containment Vessel Pipe whip for postulated breaks at the RCS-connected line nozzles on the inner CNV head also does not result in impact on essential or safety-related SSC, other than the CNV. The only target SSC to be evaluated are I&C cables passing through CNV head penetrations and the tops of the CRDMs. Figure C-3 shows three of the break locations on the interior of the CNV head in two cross-section views (both the pressurizer spray and degasification line break locations are within the circle in Figure C-3(b)). In Figure C-3(a), the injection line could move downward and pivot either toward or away from the CNV wall, depending on which bend(s) yields. The separation between the centerlines of the break and the nearest I&C penetration is 20 inches. However, per Section C.4.1, the pipe does not whip, because the long straight length extending downward from the break location (i.e., L = 37.9 inches16) results in the need for an Lh of 38.2 inches, compared to an actual Lh of less than 30.8 inches. Figure C-3(b) shows the pressurizer spray line (grey) and degasification line (purple). Both of these lines are susceptible to whip, with the broken pipe end moving downward and swinging into the CNV. Therefore, neither of these pipe whip cases contact cabling or the rod travel housings atop the CRDMs. The remaining postulated break is the discharge line. Like the injection line, the pipe end moves downward and pivots either toward or away from the CNV wall. The (L = 19.59 inch) piece at the end results in the need for an Lh of at least 29.7 inches. This is an underestimate because there are two pipe bends, both of which absorb energy between 16 The lengths noted for the end segments exclude the 4.5 inches for the present location of the RCS check and excess flow check valves, so that they are conservative after relocation of the valves to outside the CNV. Also, the determination of whether a pipe whips was done at a consistent temperature and pressure (543-degrees Fahrenheit and 1893 psia); variations in temperature and pressure with break location would have small effects on material properties and on thrust force, but conclusions would not change. © Copyright 2018 by NuScale Power, LLC 97

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 the break and support. Whether or not the pipe whips, the 19.59 inch long end piece is too short to impact the nearest I&C cable. As for the DHRS end breaks, the pipe segments are in the annulus between the CNV and RPV and are at an elevation separated from the RRVs so as to be out of range of the main valve bodies, although they could sever a trip/reset line, which is acceptable as discussed in Section 4.3.1. For the locations discussed, pipe whip impact due to potential HELBs in the CNV does not adversely affect the integrity or functionality of safety-related and essential SSC, and further evaluation of pipe whip impact loads at those locations is not needed. ((

                                                                                                           }}2(a),(c)

Figure C-2. Separation of reactor coolant system line terminal ends from emergency core cooling system valves

© Copyright 2018 by NuScale Power, LLC 98

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                    }}2(a),(c)

Figure C-3. Reactor coolant system breaks on underside of containment vessel head C.3 Pipe Whip Impact in the Reactor Building Ruptures in the NPM bay are excluded. The RXB structural integrity and, in particular, the integrity of the pool wall must be assessed, so pipe whip impact force on concrete surfaces is determined.

  © Copyright 2018 by NuScale Power, LLC 99

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 C.4 Simplified Solution of Pipe Whip Impact Velocity Figure C-4 shows a piping system of two mass segments, m1 and m2 (lower case used to differentiate from moment), and a thrust load applied at the break location immediately after rupture, depicted by position A. The thrust load causes a bending moment over the length of piping, where a plastic hinge forms allowing the pipe segment to rotate at an angular velocity, (rad/sec) and traverse a path, , as depicted by position B. Figure C-4. Pipe whip example © Copyright 2018 by NuScale Power, LLC 100

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 After break-opening, the steady-state jet thrust force, Fb, is (see Section 3.9.4): Fb = CT Po Ae Eq. C-1 where, Fb = Steady state thrust force at the break (lbf), CT = Thrust coefficient (unitless), Po = Internal system pressure (psia)17, and Ae = Pipe flow area (in.2). As identified in Section 3.9.4.1, the thrust coefficients are: CT = 1.26 (Saturated or superheated steam) CT = 2.0 (Non-flashing water jets) The hinge is formed at a distance, Lh ,from the break, resulting in a plastic moment defined by: Mp Lh = Eq. C-2 Fb where, Lh = Distance from hinge point to pipe as shown in Figure C-4, and Mp = Bending moment. The above form is commonly used in static analyses, yet neglects any influence of pipe length from the break to the first elbow, as well as restraint effects. It allows for an estimation of the minimum unrestrained length of pipe that causes the formation of a plastic hinge but leads to short hinge lengths. A more accurate formulation for the hinge length often used in restraint design, based on energy balance (Reference C.7.1) assumes the possibility of an additional length of piping, L, located perpendicular to rotational motion (Figure C-4, portion of piping labeled m2 ): 17 SRP 3.6.2 stipulates that the initial condition used should be the one with the greater of the contained energy at hot standby or at 102% power. For the NuScale design, "hot shutdown" is the equivalent of hot standby and is a lower energy state. © Copyright 2018 by NuScale Power, LLC 101

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 3M p 8LFb Lh = 1 + 1 + Eq. C-3 2 Fb 3M p Nonetheless, if there is no additional length of piping, i.e., setting L = 0, the hinge length equation reduces to: 3M p Eq. C-4 Lh = Fb Here, the plastic bending moment, assuming small deformations, are taken as: M p = Sy Z p Eq. C-5 where Sy = Yield strength of pipe, and Zp = Plastic bending section modulus, which is given by 4 3 3 Eq. C-6 Zp = 3 ( ro ri ) where ro = pipe outer radius, and ri = pipe inner radius. If large deformation is assumed, which includes strain hardening behavior of the material, an approximation to the plastic bending moment capacity is given as: M p = S y Z p + ( Su S y ) Z e Eq. C-7 where Su = Ultimate strength of pipe, and Ze = Elastic bending section modulus, which is given by © Copyright 2018 by NuScale Power, LLC 102

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Ze = 4 ro (ro 4 ri 4 ) Eq. C-8 In solving the pipe whip problem, work-energy principles are applied to the model of Figure C-4, while noting that position A depicts the piping system immediately after rupture and just before any motion takes place, and position B depicts when impact occurs with a target. Therefore, the kinetic energy at position A plus the work done in going from A to B, is equal to the kinetic energy at position B. The effective mass of the system and associated kinetic energy are generally derived from dynamic principles (References C.7.1 and C.7.4) as: m1 meff = + m2 Eq. C-9 3 ( KE ) A + WA B = ( KE )B Eq. C-10 Where the work is defined as the thrust force over the distance traversed minus the plastic moment resistance: WA B = Fb M p Eq. C-11 where

                   = the angle through which the pipe whips.

The kinetic energy of the whipping pipe about the hinge point is based on rotational kinematics (see Figure C-5 and Figure C-6): 1 ( KE )B = Ih 2 Eq. C-12 2 where

                   = Angular velocity, and Ih = Mass moment of inertia about hinge-point.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Figure C-5. Mass moment of inertia about centroidal axis Figure C-6. Mass moment of inertia about hinge location. The rotational mass moment of inertia, Ih, of a tubular pipe section about the hinge-point is found from the parallel-axis theorem. At the centroidal axis of the pipe, the mass moment of inertia is: I h = I o + meff d 2 Eq. C-13 where 1 1 Eq. C-14 I o = meff R 2 + L2h 4 12 Lh d= Eq. C-15 2 Then, substituting 1 1 I h = meff R 2 + L2h Eq. C-16 4 3 © Copyright 2018 by NuScale Power, LLC 104

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 The work-energy equation can be re-written and solved for the tangential linear velocity at the point of impact: 1 Fb M p = Ih2 Eq. C-17 2 2 1 V Fb M p = I h t Eq. C-18 2 Lh Solving for the pipe velocity at impact: 1/ 2 2 Vt = Lh ( Fb M p ) Eq. C-19 Ih C.4.1 Screening for Onset of Pipe Whip The ability to cause a pipe whip is dependent on the thrust and distance to the plastic hinge point. Like a lever, a smaller force is needed when applied at a longer distance. Because the thrust force Fb is specific to the system pressure and break flow area, each break type (e.g., chemical and volume control system, main steam system) has a minimum Lh that is necessary to initiate pipe whip. From Eq. C-4, the minimum distance Lh for the thrust force to overcome the pipe resistance to bending can be determined. Alternatively, if there is a substantial length of pipe at the end of the whipping segment, then Eq. C-3 applies. Resulting values for 304 stainless steel at operating temperature are shown in the table below: Table C-2. Maximum hinge length Lh to avoid pipe whip Lh (feet) from Eq. C-3 System Lh (feet) from Eq. C-4 for L = 5*OD18 CVCS 1.1 2.1 MSS 9.2 21.4 Thus, if the distance from the break exit axis to the plastic hinge axis in the plane of rotation is less than the values shown in the table above, then the pipe does not whip. For CVCS lines, any pipe break where the projected plane separation is less than the threshold can 18 Reference 1.4.3.2 Subarticle NC-3642.1 identifies requirements for bend radius from three to six piping diameters. Five piping diameters (5D) are used as an example for an area such as the pipe gallery where tight radius bends are not needed to provide a compact layout. Because a pipe restraint is usually not placed on a curved piping segment, the bend radius is a likely minimum segment L on the end of the pipe. © Copyright 2018 by NuScale Power, LLC 105

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 be omitted for consideration of pipe whip and assessed for a jet from a limited separation break. For the MSS pipe in the RXB, this information cannot be used for screening at this time because pipe arrangements are not finalized, but it could be used to inform the placement of pipe whip restraints in the future. C.4.2 Maximum Impact Force The maximum dynamic impact force on a potential structure, system, or component target is estimated from the tangential pipe velocity by equating the kinetic energy at impact with the potential energy of the target in compression, such that: 1 1 meff Vt 2 = KSCC xmax 2 Eq. C-20 2 2 where, xmax = Maximum compression of impacted spring, and KSSC = Stiffness of structure, system, or component. Solving for xmax : meff xmax = Vt Eq. C-21 K SCC Therefore, the maximum spring force is merely Fi = K SCC xmax Eq. C-22 Simplifying terms, results in a maximum impact force of: Fi = Vt meff K SSC Eq. C-23 The above estimation is conservative because it does not consider the materials plastic deformation, crushing, gaps or strain-rate effects. Additionally, the stiffness estimation is based purely on elastic motion, which tends to overpredict the impact force. C.4.3 Impact on Concrete Wall A steam line pipe whip event within the pipe gallery is capable of applying a large impact force to the concrete wall. Treating the reinforced concrete wall as a spring in compression, as described in the prior section, is not an acceptable method for evaluating concrete and overpredicts impact loads generated. Concrete penetration equations © Copyright 2018 by NuScale Power, LLC 106

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 developed via empirical relationships provide a means of estimating damage based on an impact velocity. The Sandia formula developed by Young in SAND 97-2426 (Reference C.7.5) is an empirical representation taking the form for the depth of penetration, D as: 0.7 W D = 0.00178Kh SN (V 100) for V 200 ft/s Eq. C-24 A where, K h = Mass scaling term for hard targets (use 1.0) (unitless), S = Penetrability index (unitless), N= Nose performance coefficient (unitless), W = Weight of penetrator (lbm), A= Cross-sectional area (in.2), and V = Impact velocity (ft/sec). The nose performance coefficient and penetrability index are: Ln N = 0.18 + 0.56 for tangent ogives Eq. C-25 d Ln N = 0.25 + 0.56 for conic shapes Eq. C-26 d where, L N = Length of penetrator nose (in.), and d = Penetrator diameter (in.) 0.3 5000 S = 0.085 K e (11 P )( tcTc ) 0.06 Eq. C-27 fc where, P= Volumetric percent rebar in concrete (~2 percent), tc = Cure time of concrete (if greater than one year, use 1.0), Tc = Target thickness in penetrator diameters or caliber, and f c' = Unconfined compressive strength at 28 days (psi). © Copyright 2018 by NuScale Power, LLC 107

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 0.3 F Ke = Eq. C-28 W1 where, W1 = Target width in penetrator calibers (unitless), and F = 20 for reinforced concrete (unitless). A concern with a high-energy impact on a concrete wall is spalling, which is the dislodgement of concrete on the far side. Spalling occurs when the impact generates a compression wave that, in turn, causes a tension wave that exceeds the compressive strength of the concrete. According to Reference C.7.8, the spalling depth typically does not exceed the reinforcement depth. The National Defense Research Council formula (Reference C.7.8) for spalling is:

                                   = 2.12   + 1.36                                       Eq. C-29 where hs = Necessary concrete thickness to prevent spalling.

Work has been performed over the years to characterize damage to nuclear plant and other reinforced concrete structures. McLean, et al. in NBSIR 86-338 (Reference C.7.6) state: The investigations found that the high velocity impact of a missile on a reinforced concrete slab or shell was a localized phenomenon as large deformations and damage occurred only in the immediate zone of the impact. Based on this observation, localized penetration and spalling do not adversely affect the structural capability of the RXB. Penetration and spalling should be avoided in regard to the pool walls. Section C.6 provides an example problem of pipe whip for an NPS 12 piping system within the pipe gallery. The target is a reinforced concrete wall, and the maximum depth of penetration is determined. C.5 Reactor Building Piping Arrangements The piping arrangements in the RXB outside the bioshield are not yet finalized, and the final piping analysis is a COL item. Figure C-7 shows a notional layout for MSS (bright green) and FWS (pale green) piping and some surrounding systems. Two major structural walls are shown running the full width of the figure (slightly transparent lavender): the pool wall is 5 feet thick, the RXB outer wall is 3 feet 41/2 inches thick, and the floor is 3 feet thick. © Copyright 2018 by NuScale Power, LLC 108

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 The gallery is congested and much of the MSS piping is in the same plane. It is likely that a main steam system HELB pipe whip impacts another pipe, but indeterminate if it has sufficient length to strike either of the walls. As noted previously, SRP Section 3.6.2 states that a whipping pipe does not cause a break or crack if it hits a pipe of equal or larger diameter and thickness. On this basis, an MSS pipe whip can be assumed to rupture a NPS 4 bypass or NPS 8 FWS line in the pipe gallery. The combination of an MSS rupture with a resultant bypass rupture has the bounding M&E release. ((

                                                                                                      }}2(a),(c)

Figure C-7. Potential high-energy line break locations in pipe gallery C.6 Bounding Main Steam System Pipe Whip This section shows the calculational steps necessary to determine the dynamic impact force of pipe whip on a reinforced concrete wall. Assume that a NPS 12 main steam line ruptures creating a HELB in a pipe gallery near the NPM bay wall. The MSS piping in this area is schedule 80 made of SA-335 P11 material. The hinge length Lh of 20 feet is selected to bound expected arrangements, based on the available space in the pipe gallery. The jet thrust reaction force is sufficient to create a plastic hinge, allowing the pipe to whip. The assumed rotation of the pipe extends over a 30-degree sector, i.e., = 30 degrees (refer to Figure C-4). Table C-3 shows how the information in this appendix is used to determine the speed of the whipping MSS pipe section. In turn, this speed is used to find the depth of penetration into an RXB concrete wall. Finally, the necessary thickness of the concrete wall to avoid © Copyright 2018 by NuScale Power, LLC 109

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 spalling is determined. Even if spalling did occur, the loss of concrete would be limited in area and in depth, having a minor effect on the structural capability of the structure. Table C-3. Example of simplified pipe whip analysis © Copyright 2018 by NuScale Power, LLC 110

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 The results show an impact velocity of ~377 ft/sec for the 12-inch MS line. The impact velocity is applied to a reinforced concrete target through an elbow geometry that is similar to an ogive shape. Again, this result is conservative yet provides a simple way to bound the potential penetration depth, which is calculated to be 6.7 inches in the example. Results show that the NPS 12 main steam line does not penetrate the reinforced concrete wall, as summarized in Table C-4. For the 60-inch thick concrete pool wall, the maximum depth (20 foot pipe length whip through an angle of 90 degrees) represents approximately 22 percent of the overall wall thickness. No spalling occurs for the 60-inch thick structural walls. Table C-4. Reactor building wall penetration depth (inches) for main steam system pipe whip impact Length of Pipe Segment Angle through which Segment Whips L (ft) = 30 = 60 = 90 10 2.5 4.2 5.4 15 4.8 7.7 9.8 20 6.7 10.5 13.4 C.7 References

1. Micheli, I. and P. Zanaboni, An Analytical Validation of Simplified Methods for the Assessment of Pipe Whip Characteristics, Transactions of the 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17), Prague, Czech Republic, August 17-22, 2003.

2. Hayadi, H.M., Simplified Pipe Whip Dynamics, Journal of Pressure Vessel Technology, American Society of Mechanical Engineers, (1984): 106(2):213-215.

3. Beer, F.P. and E.R. Johnston, Vector Mechanics for Engineers, Statics and Dynamics, McGraw-Hill Book Co., New York, NY, 1977.

4. Piersal, A.G. and T.L. Paez, Harris Shock and Vibration Handbook, 6th Edition, McGraw-Hill Book Co., New York, NY, 2010.

5. Sandia National Laboratories, Penetration Equations, SAND-97-2426, Albuquerque, NM, October 1997.

6. US Department of Commerce, Punching Shear Resistance of Lightweight Concrete Offshore Structures for the Arctic: Literature Review, NBSIR 86-338, , Gaithersburg, MD, May 1986.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

7. Dusenberry, D.O., ed. Handbook for Blast Resistant Design of Buildings, John Wiley &

Sons, Inc., Hoboken, NJ, 2010.

8. Szuladzinski, G., Formulas for Mechanical and Structural Shock and Impact, CRC Press, 2009.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Appendix E. Jet Impingement As discussed in Appendix B, jets issuing from pipe breaks in the NuScale plant are not susceptible to dynamic amplification or resonance. However, target SSC potentially in the path of postulated breaks must be assessed for the load imparted by the jet. Three categories of jets are considered:

1. liquid jets

2. two-phase jets

3. steam jets As for other effects, jet behavior and effects differ for the three areas of the plant:

Inside the CNV: breaks are limited to NPS 2 RCS-connected piping because SGS piping meets LBB. Only a degasification line break is initially steam, but spray line break reverse flow almost immediately turns to steam. Other breaks such as injection line or spray line forward flow are two-phase.
In the NPM bay outside the CNV (under the bioshield): no postulated breaks occur because piping satisfies break exclusion criteria of BTP 3-4 Paragraph B.A.(ii) and (iii).
In the RXB: piping arrangements are not finalized, so break locations and jet directions must be assumed to be anywhere in the rooms containing high-energy piping. The piping is limited to NPS 12 and 4 main steam system, NPS 6 feedwater system, and NPS 2 to 3 chemical volume and control system piping at various pressures and temperatures. Main steam system jets are steam only, whereas FWS and CVCS breaks are two-phase jets.

The concern for jet impingement that underlies regulatory guidance is the stripping of insulation with subsequent sump blockage (GSI-191). In the NuScale plant, there is no piping insulation inside the CNV and stripping of insulation outside the CNV has no deleterious safety effects. This raises the impingement damage threshold from four psig to more than 190 psig (NUREG/CR-6808), based on the impingement pressures for which metal insulation sheathing has been found to not be damaged during testing. E.1.1 Total Force The total force by the jet (adjusted for thrust coefficient) cannot exceed that at the break exit plane, which is (( }}2(a),(c) for CVCS and MSS, respectively (Table 3-7). E.1.2 Liquid jets Liquid jets are assumed to not expand and to not droop with distance. The only areas subject to liquid jets are in the RXB where CVCS low temperature, high pressure piping is present. There are no essential SSC in these areas and the liquid jets are considered to have less potential to damage concrete structure than steam jets, which are shown to be acceptable. © Copyright 2018 by NuScale Power, LLC 114

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 E.1.3 Two-phase jets Two-phase jets are assessed using the methodology of NUREG/CR-2913. A bounding approach is taken by applying conservative criteria for jet formation in order to avoid the need to analyze individual break locations in the CNV and RXB. E.1.3.1 In the Containment Vessel Although the low operating pressure of the CNV is a deviation from the experimental and analytical basis of NUREG/CR-2913, the low ambient pressure results in faster expansion of the jet, which is conservative when estimating loading. This is supported by the CFD analysis of blast waves described in Appendix F. Although that analysis is terminated while the jet is still forming, Figure F-8 and Figure F-9 show the half-angle of the 10 percent steam region (grey) already exceeds 60 degrees within the first millisecond. Only RCS-connected NPS 2 pipe breaks need to be evaluated. The inputs needed for the NUREG/CR-2913 (hereafter referred to as just 2913) methodology are the system static thermodynamic conditions, which are shown in Table 3-3.

a. Static temperature and pressure determine the entropy from Figure D.1 of 2913.

b. Entropy and break flow rate are used to obtain the stagnation temperature T0 from either Figure D.4 or D.5 of 2913.

c. Given the stagnation temperature and flow rate Ge, Figure D.6 provides the stagnation pressure P0. However, Figure D.7 is used to find the stagnation quality X0 if blowdown is initially two phase.

d. Given the stagnation pressure P0 determined above, the corresponding saturation temperature at stagnation conditions Tsat,0 is found, which allows the degree of subcooling of the system at the break to be determined from the equation:

                                       =       ,                                         Eq. E-1 The relevant graph of Appendix A of 2913 is selected to obtain target pressure and total force on the target for appropriate values of P0, T0, or X0, and distance to the target in L/D.

Although the graphs can be used to determine the ZOI, the ZOI in the CNV is assumed to be anywhere in the forward facing hemisphere because of the greater spreading angle in the low-pressure CNV and possible pipe whip. Similarly, in the RXB, the generic approach of a ZOI that includes everywhere allows for breaks at any locations determined once pipe routing is finalized and for pipe whip. © Copyright 2018 by NuScale Power, LLC 115

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 E.1.3.2 Example 2913 Calculation of Two-Phase Jet Behavior Find break mass flux for a CVCS break: (break flow) © Copyright 2018 by NuScale Power, LLC 116

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Find stagnation pressure: Find saturation temperature at stagnation conditions:

                                       , ( = 67 ) = 556

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Calculate degree of subcooling:

                               =     ,    = 556   508  = 48 Use plot for  = 50  and  = 80  to bound actual subcooling/pressure.

The final step involves selecting the correct figure representing the pressure contours of a jet most closely matching the thermodynamic conditions of 48-degrees Kelvin subcooling and 67 bar. This is Figure A.39 from 2913. The figure shows pressures at specific points downstream in L/D and radially from the jet centerline in r/D. The origin of the plot is the jet centerline at the break exit plane, and the shaded area at the lower left is the jet core (the region that has not yet begun to interact with the environment). The letters A through D refer to the key for pressure (letters beyond D for pressures above 10 © Copyright 2018 by NuScale Power, LLC 120

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 bar are not plotted). For example, a letter B indicates pressure was 2.5 bar at 4 L/D and 1.5 r/D. Figure A.39 shows that the jet core dissipates within 2 L/D or about 3.4 inches for a thermodynamic condition similar to a CVCS HELB. At 2.5 L/D and 1 r/D, the single D point is a pressure of 10 bar (145 psig), which is already below the conservative NuScale damage threshold of 190 psig. Within 4 L/D or about 6.8 inches, the jet peak pressure has dropped to below 5.0 bar (72.5 psig). The A points representing 1.0 bar correspond to the edge of the jet. The jet persists beyond 7.5 L/D, which is indicative of the concern for fibrous insulation damage at pressures of 4 psig out to a 10 L/D penetration distance. For NuScales design, pressures at about 2 L/D are low enough to cause no damage to the hard components. E.1.4 Steam Jets E.1.4.1 In the Containment Vessel For breaks inside the CNV, expansion of the jet into the low-pressure surroundings results in different behavior than usually experienced for HELBs. Wider jet spreading is expected to occur because the initially low air density of a CNV pressure below 1 psia removes most of the resistance to jet expansion, as seen in the initial jet formation calculated by the blast effects CFD analysis (see Figure F-8 and Figure F-9 which show a half-angle exceeding 60 degrees). The wider jet expands the ZOI but substantially reduces the pressure and the penetration length, because the mass and energy of the jet are more widely dispersed. Although pressure within the CNV increases with time, the pre-existing wide expansion of the jet persists because the jet is already established. For a circumferential break with limited separation, ANSI/ANS 58.2 provides a complicated method to determine the three regions of jet expansion. For Region 3 (beyond the asymptotic plane, which is where jet static pressure approaches ambient pressure), a 10-degree half-angle is specified. For simplicity, 10 degrees is assumed for the entire jet length, which is an underestimate of the expansion when determining drop off of pressure with distance. The pressure in the downstream jet depends on this angle, thermodynamic conditions in the pipe, and the separation of the pipe ends. ANSI/ANS 58.2 specifies that assumed separation of pipe ends be limited to one-half the pipe inner diameter. For NPS 2 chemical and volume control system piping, the maximum separation Wf is 0.844 in. This geometry is depicted in Figure E-1 in the left image. In this case, the pressure drops off with distance in accordance with the increasing circumference of the jet and also with the widening of the disk from its initial value of Wf at the pipe surface. The area Aj of the jet at a given radial distance r from the pipe axis becomes:

                                  =  + 2  (       )  (tan 10)                          Eq. E-2 2
                                           = 2 Eq. E-3

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 where, Aj = the area of the jet (in.2), r = the radial distance from the centerline of the broken pipe (in.), Wf = the width of the break (i.e., distance between pipe ends) = De/2 (in.), Wj = the width of the jet at distance r (in.), DE = the pipe inner diameter (in.), and t = pipe wall thickness (in.). Eq. E-2 results in a drop off in pressure imposed on more distant targets exposed to the jet, as shown in Figure E-2 and quantified in Figure E-2 Jet zone of influence and expansion for circumferential break with full separation in containment vessel Table E-1. Regulatory guidance is that the ZOI is assumed to extend to a diameter of 25 times Wf (i.e., 21 in.) and also to 25 times Wf axially centered on the break, as shown in the right side of Figure E-1. Considering that at a distance of less than 4 inches from the outer wall of a CVCS break, exit pressure has decreased to nearly below the 190 psig damage threshold, based on a CVCS break exit total pressure of 2330 psia (Table 3-7), the jet does not damage SSC more than four inches away. Figure E-1. Jet expansion and zone of influence for circumferential break with limited separation For circumferential breaks with full separation, it is assumed that any essential system or component is within the ZOI if it is located within the forward-facing hemisphere (see right image of Figure E-2) based on the original pipe orientation. As noted for the limited separation case, applying the break exit pressure over a large ZOI would be a large overestimation of the possible jet impingement loading. ((

                                                                                                  }}2(a), (c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                       }}2(a), (c)

((

                                                                                                       }}2(a), (c)

Figure E-2. Jet ZOI and expansion for circumferential break with full separation in CNV Table E-1. CVCS jet impingement pressure versus. distance for limited separation in CNV Distance r Distance from Jet Width for 10º Jet Area Pressure % of Pressure at from Pipe Pipe Outer Half-angle (in.) (in.2) at r Pipe Inner Wall Axis (in.) Wall (in.) (psia) 0.84* N/A 0.84 4.5 (( 1.19** 0.00 0.84 6.3 2 0.81 1.13 14 3 1.81 1.48 28 4 2.81 1.84 46 5 3.81 2.19 69 12 10.8 4.66 351 }}2(a),(c)

Inner diameter

** Outer diameter Includes 1.26 thrust coefficient CT

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

                                                  =                                                   Eq. E-5 4
                                              =                                                     Eq. E-6 where, Dj = Jet diameter at distance L/DE (in.),

L/DE = Distance of nearest point on impingement surface in L/D (unitless), DE = Inside diameter of break exit (in.), Aj = Total cross-sectional area of the jet at the target SSC (in.2), Pj = Applied jet pressure at nearest target surface, CT = Thrust coefficient (unitless), Po = Internal system pressure (psia)19, and AE = Pipe flow area (in.2). Applying Eq. E-4 and Eq. E-5, the jet pressure variation with distance is given in Table E-2. ((

              }}2(a), (c) 19 In accordance with SRP 3.6.2, jet thrust load is based on operating pressure and temperature.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                       }}2(a),(c)

There are no subcooled, nonflashing jets for HELBs inside the CNV. Table E-2. CVCS steam jet impingement pressure versus distance Distance L (in.) L/DE Aj30 / AE Total P30 (psia) Total P30/Total P60 0 0 (( 1 0.6 2 1.2 3 1.8 4 2.4 5 3.0 6 3.6 7 4.1 8 4.7 9 5.3 10 5.9 15 8.9 20 11.9 25 14.8 30 17.8 35 20.7 40 23.7 }}2(a),(c) Includes 1.26 thrust coefficient CT E.1.4.2 In the Reactor Building In the RXB, the distance between a break and a target structure, system, or component is not defined because RXB piping arrangements have not been finalized. To verify suitability of the design of the RXB, bounding HELB scenarios have been identified. The MSS lines are much larger and contain more energy than any other potential sources in the RXB. Demonstrating passing performance for MSS breaks provides confidence that final HELB analysis results are satisfactory. Therefore, a conservative approach is taken in which the jet impingement pressure is assumed to be the same as that at the break exit (i.e., no reduction for spreading with distance). For an MSS HELB, the break exit pressure is 500 psia. Applying the thrust coefficient CT of 1.26 yields a jet impingement pressure of 630 psi, which is about one-eighth of the minimum compressive strength of the concrete. © Copyright 2018 by NuScale Power, LLC 125

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 E.2 Jet Impingement Force The force delivered by an impinging jet is highly dependent on geometry:

Intersection of the area of the jet with the projected area of the target perpendicular to the jet.
Angle of the jet to the surface.
Shape of the surface.

This dependency is usually represented by:

                                       =        cos                                     Eq. E-7 where, PI = Impingement pressure (psia),

AI = Area of intersection of the jet and the projected target surface area perpendicular to jet axis (in.2), Yj = Normal load applied to a target by the jet (lbf), SF = Shape factor for target SSC (unitless)(see Table E-3), DLF = Dynamic load factor (unitless), and

                   = Angle made by jet axis and line perpendicular to predominant target surface.

Table E-3. Shape factors for jet impingement Target Shape Shape Factor Reference Jet impinging on flat surface 1.0 N/A Circular jet on pipe with jet diameter > pipe diameter 0.576 ANSI/ANS 58.2 Elliptical cylinder 2:1 major:minor axis ratio (CD = 0.6) 0.3 ANSI/ANS 58.2 Square cylinder (CD = 2.0) 1.0 ANSI/ANS 58.2 Eq. E-7 is based on the assumption that the jet is not spreading, as shown in Figure E-3. The left side of the figure shows a non-spreading jet impinging on a flat surface normal to the jet. This scenario results in a maximum impingement force. If, however, the jet is not normal to the surface, then the jet force is reduced as the cosine of the angle from normal, as shown in Figure E-3(b). In the extreme, for an angle of = 90 degrees, the jet is parallel to the surface and imparts no force. © Copyright 2018 by NuScale Power, LLC 126

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 However, the situation is more complicated for an expanding jet, as shown in Figure E-4. If the jet is spreading with a half-angle , then all flow lines except the jets axis (short dash arrow in Figure E-4 (b)) interact with the surface at angles that increase with distance from the axis. This is just like having all off-axis portions of the jet impinging a surface at increasing angles. If the jet to surface angle is not normal, then there may be no flow line that is normal to the surface (short dash arrow in Figure E-4(b) such that the force is farther reduced. In addition, the angled surface points are at different distances from the jet exit, such that the jet has spread more widely by the time it encounters the surface, thereby again reducing the pressure. If the target surface is large and intersects the entire jet, then this has no effect. Where the intersection is not complete, the distance at which the jet pressure is determined is important, at least within 5 L/D where the jet is expanding at the greatest half-angle. (a) (b) Figure E-3. Jet Impingement on flat plate (a) (b) Figure E-4. Expanding jet impingement on a flat plate © Copyright 2018 by NuScale Power, LLC 127

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Introducing the complication of an angled, spreading jet off-center to an angled, limited size, non-flat surface results in an overestimate of the impingement force. This is shown graphically by comparing Figure E-5(a) to (b) and (c). In each part, the jet spreading, the target size, and the break-target separation are the same, but (b) and (c) show that much of the jet misses the target, even if the cross-sectional areas of the jet and target are similar. (a) Side on (b) Offset side on (c) Offset, end on, inclined Figure E-5. Expanding jet impingement on a cylinder As noted in the previous section, the RVV impingement pressure would be no more than 43.6 psia. The RVV is about 12 inches in diameter and 20 in. long, and the jet diameter is also about 12 in. Even assuming optimum alignment for maximum interaction of the jet with the valve, the valve is angled downward at about (( }}2(a),(c) from horizontal. Assuming the areas of the jet and valve cross-section subject it to the full CVCS exit plane force of (( }}2(a),(c) (Table 3-7) applied at (( }}2(a),(c) with and a cylindrical shape factor (Table E-3) of 0.576, the total force is only about ((

                     }}2(a),(c) times its dead weight.

But this force is transitory, because there is no obstacle to stop the pipe whip with the jet pointed at the valve. Impingement effects testing fixes the jet exit pointed at the target. In in postulated HELB, the speed of the end of the whipping pipe increases with the angle through which it moves. Although the speed depends on the exact pipe configuration, within 10 degrees of starting its whip it should be moving more than 100 ft/sec, with the jet cross-section sweeping even faster as distance from the break lengthens. For 100 ft/sec, a 12-in. diameter jet sweeping across a 12-in. diameter target exposes the target to a load for a maximum of 1/100th of a second, with most of that time being a partial load. With its compact size and heavy metal walls, an RVV is a very stiff component. Because of the sinusoidal application of the jet force and its rapid passing, the jet impingement is an impulse with a duration short in comparison with other loads and need not be combined with them. Further, the dynamic load factor DLF in Eq. E-7 can be set to 1. © Copyright 2018 by NuScale Power, LLC 128

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 E.3 Jet Impingement Summary Jet impingement is of low significance in the NuScale design:

The total impingement force is small because of the small size of CVCS and MSS piping.
A conservatively wide ZOI is applied.
In the CNV:

o The trajectory of postulated whipping pipes does not result in a jet pointed directly at an essential target structure, system, or component, except possibly the CNV, which is capable of withstanding the much higher pipe whip impact load. o Insulation stripping concerns do not apply, so the threshold for essential SSC damage is set at 190 psi, based on testing showing that metal reflective insulation is not damaged. o A conservatively shallow jet expansion half-angle is assumed for steam jets, and NUREG/CR-2913 is used for two-phase jets. Considering the decrease of jet pressure with distance from the break exit, impingement pressure has dropped below the component damage threshold of 190 psi within four inches. At closest expected approach to an RVV of about 10 inches, the impingement pressure would be less than 45 psia. o The maximum total load is (( }}2(a),(c). For an RVV, the impingement force would be further reduced by the target shape factor and angle to below ((

                             }}2(a),(c).

o The rapid movement of the whipping pipe limits the imposition of this small load to less than 1/100th of a second.

In the RXB, no credit is taken for reduction in pressure with distance. Impingement pressures and total force are small compared to the load capacity and erosion would be negligible.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Appendix F. Blast Effects F.1 Background F.1.1 Blast Wave Behavior Standard Review Plan 3.6.2 requires assuming a maximum break opening time (i.e., the duration that it takes for an HELB to fully open) of one millisecond, unless a combined crack propagation time and break opening time greater than one millisecond can be substantiated. A very rapid break opening time for a HELB can cause a blast (i.e., shock) wave to form, driven by a rapid release of mass and energy. If the rupture opens over a period of more than a few milliseconds, the mass and energy release rate is too slow to create a blast wave. A blast wave could occur as a HELB injects mass and energy rapidly into the surroundings, creating a region of high density. The pressure differential accelerates material (fluid from the HELB and air in the immediate vicinity) to spread outward at the speed of sound. This material continually interacts with the undisturbed atmosphere impeding its expansion, creating higher pressure, temperature, and density at the interface. A sharp peak of pressure, temperature, and density is formed that travels at the speed of sound for the high density region, which is faster than the speed of sound (i.e., supersonic) of the surrounding atmosphere. The compression created by the blast leaves behind it a low density region into which the continuing HELB blowdown is injected. A HELB does not cause a large blast. Once the wave forms, it is moving at supersonic speed, which keeps it out ahead of the on-going blowdown, preventing additional fluid from contributing to the blast. Break initiation creates a depressurization that can move upstream in the pipe no faster than the speed of sound of the fluid in the pipe. This fluid upstream in the pipe farther than the distance traveled at the speed of sound at intact system conditions (i.e., pressure and temperature) cannot contribute to the initial blast. Therefore, defining the initial energy and mass contributing to the formation of the blast wave involves conservatively estimating the volume of fluid in the pipe that can physically escape before the blast wave initiates. Figure F-1 shows the characteristic features of a blast wave. The region of blast wave pressure above the surrounding ambient pressure PO is the positive specific impulse. It has a peak side-on pressure PSO at its leading edge and a time duration (to or td). The product (area under the curve) of peak pressure and pulse duration is the positive specific impulse. Blast wave spatial extent grows and its speed decreases away from the source, causing the pulse duration to lengthen and the peak incident pressure to decrease. The speed of the blast front depends on the pressure and density, and peak pressure can be determined from the speed of travel and vice versa, using the Rankine-Hugoniot relationship. The area of the positive specific impulse is the energy carried by the wave. © Copyright 2018 by NuScale Power, LLC 130

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Reference:

V. Krlos & G. Solomos, Calculation of Blast Loads for Application to Structural Components, European Commission Joint Research Center, EUR 26456 EN Figure F-1 Characteristic shape of a blast wave and decay with time Predicting the behavior of a blast wave is further complicated if the wave reflects off of objects, as would occur during an HELB event. Reflection can influence the loads caused by a shock wave in two ways:

The presence of condensable vapor can lead to shock-induced condensation that has been found to reduce peak pressure. Reference 1.4.3.7 states Vapor condensation at the shock front causes the coolant to be in single phase (liquid). As a result, the pressure shock is retarded and energy conversion ratio is reduced.
The damage potential of a blast wave depends on the magnitude of the overpressure upon reflection and its duration, and also on the responsiveness and projected surface area presented by the target.

F.1.1.1 Effects of Wave Reflection Reflection of an incoming wave exerts more force than blast overpressure due to the change in momentum of the gas in the blast wave. In reflection of normal sound waves (like jet impingement), the imposed load is up to twice the incoming sound pressure. For a blast wave, the accumulation of mass and energy in the vicinity of the surface is reinforced by the higher speed (i.e., momentum) of the incoming wave compared to normal sound waves. Blast wave reflection off of a surface amplifies the pressure, which is a function of both incoming blast wave speed and angle. This is shown in the Figure F-2 graph (Reference F.6.2) of the reflection coefficient Cr, which is the ratio of the reflected (outgoing) pressure to that of the incident (incoming) wave pressure. For example, an © Copyright 2018 by NuScale Power, LLC 131

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 incident wave of 100 psi encountering a surface at 30 degrees would have a Cr of about 4.5, so the reflected wave pressure imposed on the surface would be about 450 psi. A HELB is a relatively slow release of energy (compared to chemical explosions) with peak incident pressures of less than 100 psi that result in mild amplification of five or less for a wave perpendicular to the surface. Because separation distances are short within a plant, the spherically expanding blast wave is never perpendicular to an SSC surface at more than one point, so the SCC encounters a range of amplifications. An incident wave may be reinforced by overlapping of waves that have previously reflected off other surfaces. This is a complex interaction in congested areas, but is less significant where SSC are more widely spaced. Normal intersection of a shock wave with a SSC is the exception: (a) most SSC have curved surfaces, and (b) flat surfaces are rarely normal to the blast wave. Oblique reflection is when the blast wave arrives at other than normal to the surface. If the surface is not smooth, flat, and large, then the blast wave is distorted. For example, a blast wave striking a cylindrical surface encounters that surface at a different angle at each point around the circumference, with a different reflected pressure being the result (this is similar to the shape effect for jet impingement). The wave pressure drops below the ambient pressure PS0, in which the high density region is followed by a depleted zone: the negative specific impulse that can be considered similar to the troughs of ocean waves. Therefore, as a blast wave washes over a surface, the initial peak pressure at a point drops off rapidly and goes subatmospheric, while other portions of the surface farther from the blast origin are still being subjected to the high pressure portion of the wave. The net effect is that the component is not loaded at the full pressure implied by the wave peak. Blast positive impulse durations are short, usually on the order of a few milliseconds. The loading imposed is short-lived and therefore not treated as a static load. Finally, if the blast wave is created in an enclosed space, the waves reflected from different locations constructively and destructively combine, arriving at subsequent surfaces from a variety of angles and at different points in the wave transient. These interactions make the pressure loading on a surface very geometry dependent, which requires knowledge of the blast wave formation initial pressure, the distance to the reflection surface, and the angle between the incoming blast wave and the surface. Because of these interactions, the best method to determine the pressures created by a HELB blast is to perform a three-dimensional CFD analysis. However, a three-dimensional CFD is time-consuming, making it impractical to use for every possible HELB location and orientation. In view of this, NuScale defined bounding cases in the CNV and RXB and conservative inputs for each to be analyzed. © Copyright 2018 by NuScale Power, LLC 132

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Figure F-2 Blast wave reflection coefficient F.1.2 Inside the Containment Vessel The NuScale plant has a unique feature of operating the CNV at very low pressure. The low air density removes most of the medium (air molecules) necessary for a shock wave to have any substantial pressure. Postulated HELBs are limited to an NPS 2 (1.687 in. inner diameter) pipe break in the degasification line at two locations. Larger piping inside the CNV (i.e., MSS and FWS) meets LBB criteria and is excluded from need to consider a sudden rupture causing a blast wave. Other NPS 2 piping initially contains subcooled fluid with negligible blast potential (Reference 1.4.3.7). The potential for blast effects in the CNV is limited for three reasons:

a. The low atmospheric pressure means few air molecules are present to support formation of the blast wave. In other words, there is no medium to support propagation of the blast wave. By the time sufficient mass has been deposited in the CNV, the opportunity to form the blast wave has passed.

b. The only piping not excluded from pipe rupture is NPS 2, with a small mass and energy input. Also, the piping except for the degas line contains subcooled liquid; the presence of liquid in blowdown takes energy away from the blast (Reference 1.4.3.7).

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

c. Although safety-related components (e.g., ECCS valves) and instrumentation cables are nearby, they are hardened to withstand the design pressure (1000 psia) and temperature of the CNV resulting from ECCS initiation, and the cables are enclosed in conduit in the vicinity of the RPV head.

Table F-3 shows maximum force results for an HELB inside the CNV from an NPS 2 degas line rupture. Blast effects are deemed negligible and not evaluated further. F.1.3 In the NuScale Power Module Bay under the Bioshield Piping in this portion of the plant is excluded from need of consideration of dynamic effects through satisfying BTP 3-4 break exclusion criteria. F.1.4 In the Reactor Building Separation of essential components in compartments not containing high-energy piping eliminates most potential for negative effects. Piping routing in the RXB is subject to change, which could affect the postulated HELB locations. In any case, there would be a considerable number of potential locations, so NuScale has taken the approach of identifying a bounding scenario:

NPS 12 pipe break in the MSS - This is the largest diameter steam line in the RXB.

Feedwater system and CVCS pipes are smaller than MSS piping and contain subcooled liquid at intact system conditions, which moderates formation of the blast wave.

Break surroundings - Because routing of piping within the RXB has not been finalized, a conservative but hypothetical arrangement is used in which the break is postulated to occur close to another similar pipe at three different locations within a pipe gallery. This allows for developing a conservative loading on building structure and on a pipe representing a nearby line for another NPM.

F.2 Computational Fluid Dynamics Model F.2.1 Computational Fluid Dynamics Code This analysis was performed with the ANSYS CFD program CFX Version 18.0 on the servers running the RHEL Release 6.5 operating system. Correct program function was verified by the ANSYS Certificate of Conformance stored in the ANSYS users controlled software file. Installation verification was documented and validation of the applicability of CFX for the analysis of HELB blast effects was performed as described in the next section. F.2.2 Verification and Validation Reference F.6.16 provides the verification and validation (V&V) of the CFX code for blast effects. Eight test cases were analyzed to validate the CFD methodology for analysis of supersonic flows and shock waves. The CFD methodology is applicable to the analysis of © Copyright 2018 by NuScale Power, LLC 134

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 the effects of blast waves generated from sudden pipe ruptures as postulated for the NuScale reactor module design. In each test case, comparison between the simulation results obtained on three levels of grid refinement and either experimental data or theoretical predictions was performed. The methodology of ASME V&V 20 (Reference F.6.8) was used to estimate the model error () in each case. Typical model error, which is expressed as the average ratio of comparison difference and uncertainty to simulation results, is presented for each case in Table F-1. Table F-1. Summary of average error from validation analysis Case Quantities Compared Average Error (/)

1) Shock reflection Pressure, heat flux 21%

2) Oblique shock Density, Mach number, temperature, pressure 1%

3) Transient shock wave Mach disc location 5%

Mach number, pressure, temperature, density,

4) Steam-air shock tube 1%

contact surface

5) Supersonic steam nozzle Pressure 13%

6) Jet impingement - single phase Force 16%

7) Jet impingement - multiphase Force 8%

8) Blast into low pressure Pressure 8%

F.2.2.1 Phenomena Identification The formation of a blast wave and its propagation in a nuclear plant HELB features complex, interactive phenomena with limited data available to characterize the shock loads. The important aspects of modeling are the transfer of fluid mass and energy into the surrounding air, the formation and propagation of the shock wave, reflection and amplification in the crowded confines within the plant, and loading of SSC within range. Based on the fundamental physics involved in the flow, the following characteristics are relevant to be present in a validation test suite:

1. supersonic compressible flow

2. shock behavior

3. transient shock propagation

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

4. multi-component gas behavior

5. real gas effects

6. shock reflection

7. phase change (minor effect)

8. environment initial pressure The eight physical processes listed above guide the selection of the test cases. A test case may entail the modeling of more than a single process. Phase change due to rapid temperature and pressure fluctuations is not included in a test case because nonequilibrium condensation in supersonic jets downstream of the nozzle throat has been shown to increase total pressure loss in the jet (Reference F.6.9). Therefore, neglecting condensation effects is conservative for the analysis of loads due to HELB blast.

F.2.2.2 Test Case Selection Validation of the CFD method and CFX code for modeling blast effects is achieved by running test problems and comparing the results to either theoretical or experimental results. Agreement between the CFX results and the reference values provides validation and confidence that the numerical approach and mesh adequately model the associated phenomena. This process validates the ability of CFX to predict the behavior of supersonic flows of both air and steam which are possible mechanisms that would govern fluid behavior following a pipe rupture in the NuScale plant. To this end, the following eight cases were evaluated:

1. ((

                  }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                      }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Figure F-3. Verification and validation case 8 results F.3 Methodology Each break scenario is analyzed in two parts: a steady-state simulation and a transient simulation. The steady-state represents the conditions before the pipe breaks. The transient simulation starts from the steady-state results and models an instantaneous, open-ended break of the pipe. The transient CFD results are then used to generate transient load profiles on several nearby SSC of interest. Meshing is performed using the ANSYS Workbench Meshing module. The mesh is built with sufficient density to capture the relevant physics of the blast. Refinement is added around the postulated break using the Sphere of Influence method. Multiple concentric spheres are used to transition from the finest mesh directly around the break to the coarser mesh further away from the break location. Inflation layers are added to key surfaces to improve the flow resolution near surfaces. As part of the V&V of CFX for use in evaluating blast waves, the effects of mesh density were investigated. The mesh size in the vicinity of the pipe break is chosen to match the typical element size relative to the characteristic length scale of the meshes used in the V&V simulation. © Copyright 2018 by NuScale Power, LLC 138

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 F.4 Results of Blast Effects Modeling F.4.1 In the Containment Vessel F.4.1.1 Containment Vessel Break Scenarios Three scenarios were selected to provide loads that bound the potential HELBs within the CNV. High point degasification line breaks were analyzed to bound any CVCS breaks in the CNV because lines filled with subcooled liquid do not cause a significant blast. Although blast effects are geometry dependent, the degasification line break locations are representative of the geometry of any of the CVCS lines at the RPV head or CNV head. Table F-2 summarizes the key modeling parameters. Three different breaks of the degas line are considered as shown in : Case 1: upward-oriented break at the RPV nozzle. Case 2: downward-oriented break close to the RPV nozzle. Case 3: upward-oriented break immediately inside the CNV head. Table F-2. Overview of blast CFD modeling inside the CNV Parameter Selection Discussion Dimensionality 3D Model is too complex for reduced dimensionality Turbulence model SST SST model was validated as appropriate for blast waves Energy model Total energy Total energy is required for modeling supersonic flows Equation of state Ideal gas Appropriate per Section 1.4.3.1.2 of Reference F.6.15 Wall roughness Smooth Solid surfaces modeled smooth (zero sand grain roughness) Buoyancy None Buoyancy effects not considered due to short timescales Second order Time discretization Recommended setting for accuracy backward Euler Primarily a 2nd order accurate discretization that blends 1st Space discretization High resolution order terms to ensure boundedness Solver precision Double Reduces truncation error Adaptive Time step Time step adjusted by solver to increase performance 10 10-5 s RMS residuals < 10-5 Convergence criteria for RMS residual of all equations <10-5 Compressibility High speed Improved performance and stability for high speed flows control numerics (transient portion only) Topology estimate 1.05 Increases internal memory estimate (expert parameter) factor The ambient pressure in the CNV is assumed to be 0.95 psia, although normal CNV pressure is below 0.1 psia. This is a reasonable upper limit for plant operation, because it © Copyright 2018 by NuScale Power, LLC 139

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 corresponds to a saturation temperature of 100-degrees Fahrenheit, a likely maximum CNV wall temperature. A higher pressure is conservative because a higher density medium transmits the blast energy more effectively. F.4.1.2 Containment Vessel Computational Fluid Dynamics Model The simulation domain is generated from an NPM computer model of the CNV and RPV, and is simplified to remove unnecessary detail and to improve runtime of the simulations. The simplifications include removal of small components and reduction of detail for select larger components. Removed features include bolts, cables, and small pipes. These simplifications do not significantly affect the behavior of the blast wave. The SGS steam and feedwater pipes, for which loading is determined, and the degas line, which is postulated to break, are retained in the model. The overall geometry shown in is tailored to the different break locations by trimming the geometry. To simulate the blast propagation through the air space, the solid model is inverted to produce a model of the fluid domain. This process uses the simplified model as a mold from which the air space is created. Figure F-5 and Figure F-6 show a visual representation of the computational mesh for Case 1. © Copyright 2018 by NuScale Power, LLC 140

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                               }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                    }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                       }}2(a),(c)

Figure F-6. Detailed view of the mesh around the pipe break (case 1) Immediately following the break, a blast wave is formed when high pressure steam is released from the pipe. The steam quickly accelerates to supersonic velocities and propagates a supersonic pressure wave that takes the form of a blast in air. The blast expands radially outward from the break location. In each case, the blast is biased in the direction along the pipe axis. Targets in the immediate vicinity of the break are subject to the highest pressure loads. The blast loads for close targets are quickly surpassed by the jet that imparts higher loads on the targets. The opposite end of the ruptured pipe receives a significant load due to both the blast and jet. The blast is reflected by solid surfaces and may reach areas that are shielded from the initial blast. The reflecting surface is loaded by a pressure greater than the incident pressure. However, the pressure magnitudes are small because the vacuum conditions inside the CNV do not propagate the blast wave well. The effects of the HELB on the surrounding structures and components can be divided into two separate physical phenomena: blast and jet. The blast is created when high pressure steam expands into the lower pressure surroundings. It is characterized by a supersonic shock front that causes a sudden pressure increase as it propagates through the surrounding medium. The blast is not associated with bulk mass transport. Conversely,

  © Copyright 2018 by NuScale Power, LLC 143

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 the jet is characterized by bulk mass transport and forms a continuous region that is connected to the break location. Because the medium inside and outside of the pipe are different, mass fractions provide a convenient way to distinguish the blast and jet. Based on post-processing of the results, a cutoff of 10 percent steam is reasonable to distinguish between blast and jet. This distinction is used to visually separate the blast and jet effects in the contour plots provided below, where grey shading is indicative of steam from the jet. The forces on selected components are monitored continuously during the simulated transient. The calculated forces are plotted in Figure F-7 for Case 1. The traces for most loads show three distinct regions: 1) a distinct spike indicative of the sudden arrival of the blast wave and the associated load, 2) a decrease of loading as the blast wave clears the component, and 3) a sustained rise in load which eventually reaches a steady state that is caused by the impingement. Figure F-8 provides pressure contour plots at four time steps for Case 1. The results show blast pressures are low, dissipate quickly, and have a short range. Figure F-9 provides pressure contours at one time step for Cases 2 and 3, showing similar behavior. Because of the weak blast front in the low-pressure surroundings, the peak blast loads from the three CNV cases are low (Table F-3) and are bounded by the jet impingement loads. ((

                                                                                                     }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                         }}2(a),(c)

Figure F-8. Absolute pressure contours at four time steps for CNV blast Case 1

  © Copyright 2018 by NuScale Power, LLC 145

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                                 }}2(a),(c)

Figure F-9. Absolute pressure contours for CNV Cases 2 & 3 Table F-3. Maximum total forces on selected components for blasts in the containment vessel Component CNV RPV MS Piping Support ECCS Bounding FWS Pipe Head Head (Upper/Lower) Beam Valve CRDM Tube Force (lbf) (( }}2(a),(c) F.4.2 In the Reactor Building F.4.2.1 Reactor Building Blast Scenarios Given that the design of the NuScale RXB and pipe layout is not final, the following three different breaks of the main steam line are considered to generate a diverse set of break conditions with bias towards maximizing blast wave reflection and dynamic loads on representative components (e.g., valve bodies, MSS line, FWS line): Case 1: break at a MS line in the mid-gallery with the blast traveling horizontally from the turbine side towards the pool wall. Case 2: break at a MS line in the mid-gallery with the blast traveling horizontally from the reactor side towards the RXB wall. Case 3: break at a MS line in the gallery corner with the blast traveling horizontally from the turbine side towards the pool wall. F.4.2.2 Reactor Building Blast Model Table F-4 summarizes key modeling parameters for the RXB blast analysis. Geometry of the modeled region of the RXB is shown in Figure F-10. The three break locations are shown in Figure F-11. Breaks in MSS lines are analyzed because of their large diameter and high-energy content. Figure F-12 identifies SSC of interest in the modeled region, and Table F-5 is the key identified which SSC correspond to each number. Figure F-13 depicts the mesh used for RXB Case 1.

  © Copyright 2018 by NuScale Power, LLC 146

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table F-4. Overview of modeling scheme for blast analysis in reactor building Parameter Selection Discussion Turbulence Model Shear Stress SST model was validated as appropriate for blast wave Transport (SST) simulation Energy Model Total Energy Total energy is required for modeling supersonic flow Buoyancy None Buoyancy effects not considered due to short timescales Time High Resolution Primarily a 2nd order accurate discretization that blends 1st Discretization order terms to ensure boundedness Space High Resolution Blend of 1st and 2nd order terms to ensure robustness and Discretization accuracy. Blend factor is based on solution values. Solver Precision Double Reduces the truncation error Time Step Adaptive Time step adjusted by solver to achieve appropriate Courant 10 10-5 s number. RMS Residuals < 10-5 Convergence criteria for RMS residual of all equations less than 10-5 per Section 15.10.1.1.1 of Appendix F Reference 5. Solver Control High Speed Improved performance and stability for high speed flows Figure F-12. Identification of components in reactor building © Copyright 2018 by NuScale Power, LLC 149

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 Table F-5. Key to reactor building SSC of interest for blast effects Component Label Component Name Description in Figure F-12 A MS7B Valve MS line 7B isolation valve B MS7B BPV#1 MS line 7B bypass valve #1 C MS7B BPV#2 MS line 7B bypass valve #2 D FW Line 7B Feedwater line 7B E MS Line 7B EW MS line 7B east-west section F MS7A Valve MS line 7A isolation valve G MS7A BPV#1 MS line 7A bypass valve #1 H MS7A BPV#2 MS line 7A bypass valve #2 I MS Line 7A NS MS line 7A north-south section J MS Line 7A EW MS line 7A east-west section K MS Line 8B EW MS line 8B east-west section L FW Line 8B Feedwater line 8B M FW8B Valve Feedwater line 8B isolation valve N MS8A Valve MS line 8A isolation valve O MS8A BPV#1 MS line 8A bypass valve #1 P MS8A BPV#2 MS line 8A bypass valve #2 Q MS Line 8A NS MS line 8A north-south section R MS Line 8A EW MS line 8A east-west section S FW Line 8A Feedwater line 8A T FW8A Valve Feedwater line 8A isolation valve U MS Line 9B EW MS line 9B east-west section V MS Line 9A EW MS line 9A east-west section W FW7B Valve Feedwater line 7B isolation valve X FW7A Valve Feedwater line 7A isolation valve © Copyright 2018 by NuScale Power, LLC 150

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                         }}2(a),(c)

Figure F-13. Cross-section view and close-up view of the mesh in case 1 F.4.2.3 Reactor Building Blast Results The blast wave propagation from the MS8B break for Case 1 is provided in Figure F-14. Regions with steam content higher than 10 percent are colored white to distinguish between regions with blast effects and jet effects. Figure F-15 provide the force-time histories for SSC. The curves show an initial peak when the leading blast wave impacts the object. The duration of this largest, initial peak is in general about one millisecond, characteristic of an impulse load that is applied and gone too quickly for the SSC to be damaged. The subsequent peaks are associated with reflected waves that arrive after the leading wave. MS Line 8A and MS8A Valve are the two components that experienced the highest forces due to blast waves during the transient. The maximum forces on MS Line 8A NS section and MS8A Valve, which are parallel to the broken pipe, are ((

              }}2(a),(c), respectively. The maximum force exerted on the pool wall is ((
        }}2(a),(c) , which is induced by the combination of the jet and blast shock front.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                  }}2(a),(c)

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 ((

                                                                                                                                      }}2(a),(c)

Figure F-15. Force time history for various SSC for reactor building blast Case 1 Table F-6. Peak blast wave forces on selected SSC Case Component(1) Peak Force (lbf) MS Line 8A Isolation Valve (( Case 1 MS8B break towards Pool Wall MS Line 8A (north-south section) Pool Wall MS Line 9B (east-west section) Case 2 MS Line 8A (north-south section) MS8B break towards Reactor Building Wall MS Line 7B Bypass Valve #1 Case 3 MS Line 7A Isolation Valve MS7B break towards Pool Wall Pool Wall }}2(a),(c)

4. See Figure F-12 and Table F-5 for components locations.

5. The force is induced by the combination of the jet and blast shock front over the surface on the pool wall that is centered at the break point with a radius of 100 in. This radius corresponds to the spherical propagation of the shock front at approximately 2.3 ms.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0 F.5 Conclusions Three-dimensional CFD analysis of blast wave formation in the CNV and RXB has been performed using conservative modeling assumptions that bound the pressurization effects that may occur for any HELBs in the plant. Blast wave force time histories were calculated for nearby SSC of interest. The results show:

Peak forces are low and bounded by the jet thrust forces that subsequently develop. The low values are because NuScale HELBs are relatively small diameter and deposit a small amount of mass and energy in the less than one millisecond that it takes for a blast wave to form. The forces inside the CNV are particularly low because the initial low ambient pressure does not support formation of a significant blast wave.
The forces of the passing shock wave are of very short duration.

Therefore, detrimental effects of HELB-induced blast waves anywhere in the NuScale plant can be ignored. F.6 References

1. Karlos, Vasilis and George Solomos, Calculation of Blast Loads for Application to Structural Components, European Commission Joint Research Center, EUR 26456 EN, 2013.

2. Federal Emergency Management Agency, Building Design for Homeland Security, Course: IS-156; Lesson: 6 - Explosive Blast; https://emilms.fema.gov/IS0156/course/156_m7_print.htm

3. ANSYS Fluid Dynamics Verification Manual, Release 18.0, January 2017.

4. Pal, S., et.al, Verification and Validation of CFD Model to Predict Jet Loads and Blast Wave Pressures from High Pressure Superheated Steam Line Break, Paper No.

POWER2016-59675, Proceedings of ASME 2016 Power Conference, 2016.

5. Moore, M. J., et al, Predicting the fog drop size in wet steam turbines, Institute of Mechanical Engineers (UK), Wet Steam 4 Conf. University of Warwick, 1973, paper C37/73.

6. Hopkins,H.B., W. Konopka, and J. Leng, Validation of scramjet exhaust simulation technique at Mach 6, NASA Contractor Report 3003, 1979.

7. White, F.M., Fluid Mechanics, 3rd Edition, McGraw-Hill, New York, NY, 1994.

8. American Society of Mechanical Engineers, Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer, ASME V&V 20-2009, New York, NY.

9. ANSYS CFX Release 18.0 Documentation.

Pipe Rupture Hazards Analysis TR-0818-61384-NP Rev. 0

10. Alam, M.M.A., T. Setoguchi, and S. Matsuo, S., Numerical Analysis of Ideally-Expanded Supersonic Jets with Nonequilibrium Homogenous Condensation, International Journal of Computational Methods, (2013): 10(5).

11. Center for Chemical Process Safety, Guidelines for Vapor Cloud Explosion, Pressure Vessel Burst, BLEVE and Flash Fire Hazards, 2nd Edition, December 2011.

12. Zheng, H.T., et al, Computational fluid dynamics simulation of the supersonic steam ejector. Part 1: Comparative study of different equations of state, Proceedings of the Institution of Mechanical Engineers, Part C: Journal of Mechanical Engineering Science, (2012): 226(3):709-714.

13. Kitade, K., et al, Experimental Study of Pipe Reaction Force and Jet Impingement Load at the Pipe Break, International Association for Structural Mechanics in Reactor Technology (SMiRT-5), 1979.

14. Moody, F.J., Prediction of Blowdown Thrust and Jet Forces, ASME Paper 69-HT-31, August 1969.

15. ANSYS CFX Modeling Guide, Release 18.0.

16. Validation of CFD Method for Use in HELB Analysis, 1206-0082-CALC-001, Rev. 0.

LO-0918-61827 : Affidavit of Thomas A. Bergman, AF-0918-61828 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 Thomas A. Bergman I, Thomas A. Bergman, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying topical report reveals distinguishing aspects about the methodology and process by which NuScale performs its pipe rupture hazards analysis. NuScale has performed significant research and evaluation to develop a basis for this PRHA methodology 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 Pipe Rupture Hazard 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. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon AF-0918-61828 Page 1 of 2

the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4 ). (6) Pursuant to the provIsIons set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and , to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect , and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies , customers and potential customers and their agents , suppliers , licensees , and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties , including any required transmittals to NRC , have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on October 3, 2018. AF-0918-61 828 Page 2 of 2}}

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