Text

LLC Submittal of the Technical Report, Pipe Rupture Hazards Analysis, TR-121516, Revision 0
	ML23001A002
Person / Time
Site:	99902078, 05200050
Issue date:	12/31/2022
From:	Fosaaen C; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23001A001	List: ML23001A002;
References
	LO-133408 TR-121516-NP, Rev 0
	Download: ML23001A002 (1)
	v • d • e

LO-133408 December 31, 2022 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of the technical report, Pipe Rupture Hazards Analysis, TR-121507, Revision 0

REFERENCES:

1. NuScale letter to NRC, NuScale Power, LLC Submittal of Planned Standard Design Approval Application Content, dated February 24, 2020 (ML20055E565)

2. NuScale letter to NRC, NuScale Power, LLC Requests the NRC staff to conduct a pre-application readiness assessment of the draft, NuScale Standard Design Approval Application (SDAA), dated May 25, 2022 (ML22145A460)

3. NRC letter to NuScale, Preapplication Readiness Assessment Report of the NuScale Power, LLC Standard Design Approval Draft Application, Office of Nuclear Reactor Regulation dated November 15, 2022 (ML22305A518)

4. NuScale letter to NRC, NuScale Power, LLC Staged Submittal of Planned Standard Design Approval Application, dated November 21, 2022 (ML22325A349)

NuScale Power, LLC (NuScale) is pleased to submit the technical report, Pipe Rupture Hazards Analysis, TR-121507, Revision 0. This report supports Chapter 3 of the Standard Design Approval Application, Design of Structures, Systems, Components and Equipment, Revision 0. Chapter 3 supports Part 2, Final Safety Analysis Report, (FSAR) of the NuScale Standard Design Approval Application (SDAA), as described in Reference 1. NuScale submits the report in accordance with requirements of 10 CFR 52 Subpart E, Standard Design Approvals. As described in Reference 4, the enclosure is part of a staged SDAA submittal. NuScale requests NRC review, approval, and granting of standard design approval for the US460 standard plant design.

From July 25, 2022 to October 26, 2022, the NRC performed a pre-application readiness assessment of available portions of the draft NuScale FSAR to determine the FSARs readiness for submittal and for subsequent review by NRC staff (References 2 and 3). The Pipe Rupture Hazards Analysis technical report was not available for NRC readiness assessment review.

Enclosure 1 contains the technical report, Pipe Rupture Hazards Analysis, TR-121507-P, Revision 0, proprietary version. NuScale requests that the proprietary version (Enclosure 1) be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390.

The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 contains the 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

LO-133408 Page 2 of 2 12/31/2022 This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions, please contact Mark Shaver at 541-360-0630 or at mshaver@nuscalepower.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 31, 2022.

Sincerely, Carrie Fosaaen Senior Director, Regulatory Affairs NuScale Power, LLC Distribution: Brian Smith, NRC Michael Dudek, NRC Getachew Tesfaye, NRC Bruce Bavol, NRC David Drucker, NRC Enclosure 1: Pipe Rupture Hazards Analysis, TR-121507-P, Revision 0 (proprietary)

Enclosure 2: Pipe Rupture Hazards Analysis, TR-121507-NP, Revision 0 (nonproprietary)

Enclosure 3: Affidavit of Carrie Fosaaen AF-133409 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-133408 : Pipe Rupture Hazards Analysis, TR-121507-P, Revision 0 (proprietary)

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

LO-133408 : Pipe Rupture Hazards Analysis, TR-121507-NP, Revision 0 (nonproprietary)

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-121507-NP Revision 0 Licensing Technical Report Pipe Rupture Hazards Analysis December 2022 Revision 0 Docket: 52-050 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Pipe Rupture Hazards Analysis TR-121507-NP Revision 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-121507-NP Revision 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-NE0008928.

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

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

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table of Contents Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Executive Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.3 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.1 NuScale Design Features Relevant to Pipe Rupture Hazards Analysis . . . . . . . . . . . . . 8 2.2 Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.2.1 Standard Review Plan Section 3.6.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.2.2 Standard Review Plan Section 3.6.2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.2.3 Standard Review Plan Section 3.6.3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.2.4 Branch Technical Position 3-3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.2.5 Branch Technical Position 3-4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.0 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.1 Identification of Essential and other Structures, Systems, and Components Requiring Protection. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.1.1 Essential Structures, Systems, and Components . . . . . . . . . . . . . . . . . . . . . . . 19 3.1.2 Other Structures, Systems, and Components Requiring Protection . . . . . . . . . 26 3.2 Identification of High and Moderate Energy Systems in Proximity to Essential and other Protected Structures, Systems, and Components . . . . . . . . . . . . . . . . . . . . . . . . 28 3.2.1 Moderate-Energy Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.2.2 High-Energy Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 3.2.3 Postulated Pipe Rupture Locations in High-Energy Lines . . . . . . . . . . . . . . . . . 43 3.3 Rupture Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 3.4 Determination of Potential External Effects of Ruptures Including Dynamic and Environmental Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 3.4.1 Blast Waves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 3.4.2 Jet Reaction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 3.4.3 Jet Impingement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 3.4.4 Pipe Whip Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 3.4.5 Dynamic Subcompartment Pressurization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table of Contents 4.0 Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 4.1 Inside the Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 4.2 Outside the Containment Vessel, in the NuScale Power Module Bay . . . . . . . . . . . . . . 65 4.3 Outside the NPM bay, in the Reactor Building. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 4.4 Outside the RXB, in the remainder of the NPP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 5.0 Results and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 6.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 6.1 Referenced Documents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 Appendix A Break Exclusion - Compliance with Regulatory Criteria . . . . . . . . . . . . . . .A-1 A.1 Application of BTP 3-4 B.1.(ii) Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-1 A.1.1 Steam Generator System Main Steam and Feedwater Piping. . . . . . . . . . . . . .A-2 A.1.2 Containment System Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 A.1.3 Decay Heat Removal System Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 A.2 Connection of Reactor Vent Valves and Reactor Recirculation Valves to the Reactor Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-6 Appendix B Dynamic Amplification and Potential for Resonance . . . . . . . . . . . . . . . . . .A-1 B.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-1 B.2 Necessary but not Sufficient Conditions for Resonance . . . . . . . . . . . . . . . . . . . . . . . .A-2 B.3 Susceptibility to Dynamic Amplification and Resonance . . . . . . . . . . . . . . . . . . . . . . . .A-3 B.3.1 Flat Surface Within 7.5 Diameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 B.3.2 Mach number > 0.7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 B.3.3 Phase Difference Integer Multiple of 2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 B.3.4 Speed of Upstream Propagating Waves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-5 B.3.5 Period Set by Wave Speed and Distance Between Nozzle and Plate . . . . . . . .A-6 B.3.6 Thin Shear Layer near the Nozzle Lip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-6 B.3.7 Large Coherent Structures Play Main Role in Feedback . . . . . . . . . . . . . . . . . .A-6 B.3.8 Dynamic Loading as much as 50 Percent Higher than Non-Resonant Jet . . . .A-7 B.3.9 Jet Must be Axisymmetric . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-7 B.3.10 Jet Axis Normal to Impingement Surface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-8 B.3.11 Addition of Moisture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-8 B.4 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-8

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table of Contents Appendix C Pipe Whip . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 C.1 Inside the Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 C.1.1 Pipe Whip Screening . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 C.1.2 Pipe Whip Evaluation for Inside Containment Vessel Area . . . . . . . . . . . . . . . .C-5 C.2 Reactor Building Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-5 C.2.1 Pipe Whip Screening . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-5 C.2.2 SC Wall Impact by Main Steam System Pipe Whip . . . . . . . . . . . . . . . . . . . . .C-12 C.2.3 Reinforced Concrete Slab Impact by feedwater system Pipe Whip . . . . . . . . .C-15 Appendix D Subcompartment Pressurization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-1 D.1 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-2 D.1.1 NuScale Power Module Bay. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-2 D.1.2 Reactor Building Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-2 D.2 Vent Paths . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-3 D.2.1 NuScale Power Module Bay. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-3 D.2.2 In the Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-3 D.3 Analytical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-3 D.4 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-4 D.4.1 High-Energy Line Breaks evaluated. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-4 D.5 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-5 D.5.1 NuScale Power Module Bay. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-5 D.5.2 Reactor Building Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-6 D.6 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-7 Appendix E Jet Impingement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-1 E.1 Total Force . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-1 E.2 Liquid jets . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-1 E.3 Two-phase jets. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-2 E.3.1 In the Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-2 E.3.2 Example 2913 Calculation of Two-Phase Jet Behavior . . . . . . . . . . . . . . . . . . .E-3 E.4 Steam Jets . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-8 E.4.1 In the Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-8 E.4.2 In the Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-11

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table of Contents E.5 Jet Impingement Force . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-12 E.6 Jet Impingement Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-14 E.7 Jet Impingement Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-18 Appendix F Blast Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 F.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 F.1.1 Blast Wave Behavior . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 F.1.2 Inside the Containment Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-4 F.1.3 In the NuScale Power Module Bay under the Bioshield. . . . . . . . . . . . . . . . . . . F-5 F.1.4 In the Reactor Building: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-5 F.2 Computational Fluid Dynamics Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-5 F.2.1 Computational Fluid Dynamics Code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-5 F.2.2 Verification and Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-6 F.3 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-9 F.4 Results of Blast Effects Modeling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-10 F.4.1 In the Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-10 F.4.2 In the Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-18 F.5 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-28 Appendix G High- and Moderate-Energy Lines in the US460 Plant . . . . . . . . . . . . . . . . G-1 Appendix H Nonmechanistic Breaks in Main Steam and Feedwater Lines. . . . . . . . . . .H-1 Appendix I Pipe Whip Restraint and Jet Shield Design Criteria . . . . . . . . . . . . . . . . . . . I-1 I.1 Pipe Whip Restraints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-1 I.2 Jet Impingement Shields . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-2

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 List of Tables Table 1-1 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Table 1-2 Definitions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Table 3-1 Post-accident Monitoring Instrumentation Sensors inside the Containment Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 Table 3-2 Post-accident Monitoring Instrumentation Sensors outside the Containment Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 Table 3-3 High energy system piping characteristics for piping in the Containment Vessel, NuScale Power Module Bay, and Reactor Building . . . . . . . . . . . . . . . 33 Table 3-4 Barriers in the containment vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 Table 3-5 Summary of essential and other protected structures, systems, or components evaluated for external dynamic effects of pipe ruptures . . . . . . . . 49 Table 3-6 Characteristics of Blowdown at Postulated Break Locations . . . . . . . . . . . . . . . 51 Table 3-7 Break exit plane parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 Table B-1 Range of potential resonance region in the Reactor Building . . . . . . . . . . . . . .A-4 Table B-2 Wavelengths of downstream propagating waves. . . . . . . . . . . . . . . . . . . . . . . .A-5 Table C-1 Maximum hinge length Lh to avoid pipe whip for reactor coolant system lines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-3 Table C-2 Maximum hinge length Lh to avoid pipe whip . . . . . . . . . . . . . . . . . . . . . . . . . .C-6 Table D-1 Reactor Building high-energy line breaks evaluated for dynamic peak pressures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-5 Table D-2 Reactor Building high-energy line breaks evaluated for dynamic peak pressures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-6 Table E-1 Chemical and volume control system steam jet impingement pressure vs.

distance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-11 Table E-2 Shape factors for jet impingement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-12 Table F-1 Summary of average error from validation analysis . . . . . . . . . . . . . . . . . . . . . . F-6 Table F-2 Overview of blast computational fluid dynamics modeling inside the containment vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-11 Table F-3 Maximum total forces on selected components for blasts in the containment vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-18 Table F-4 Overview of modeling scheme for blast analysis in Reactor Building . . . . . . . F-23 Table F-5 Key to Reactor Building structures, systems, or components of interest for blast effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-24 Table F-6 Peak blast wave forces on selected structures, systems, or components for low power cases. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-27

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 List of Tables Table G-1 High- and Moderate-Energy Lines in the containment vessel and NuScale Power Module Bay . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . G-1 Table H-1 Comparison of main steam system and feedwater system piping in containment penetration area. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .H-2

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 List of Figures Figure 3-1 Flowchart of methodology for evaluation of line breaks for each area that contains essential or protected structures, systems, or components. . . . . . . . . 18 Figure 3-2 Proximity of High-Energy Systems to Control Building . . . . . . . . . . . . . . . . . . . 39 Figure 3-3 Safety-Related Underground Duct Bank . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 Figure A-1 Containment penetration areas - steam generator system . . . . . . . . . . . . . . . .A-4 Figure A-2 Containment penetration areas - containment system. . . . . . . . . . . . . . . . . . . .A-5 Figure A-3 Containment penetration areas - decay heat removal system . . . . . . . . . . . . . .A-6 Figure B-1 Normalized pressure of steam jet vs. distance to impingement plate (Reference 6.1.32) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 Figure C-1 Hinge location for elbow less than 90° (i.e., non-perpendicular pipe run) . . . . .C-3 Figure C-2 Resulting pipe whip due to breaks in the reactor pressure vessel discharge and injection lines at the reactor pressure vessel nozzles. . . . . . . . .C-4 Figure C-3 Mass moment of inertia about centroidal axis . . . . . . . . . . . . . . . . . . . . . . . . .C-10 Figure C-4 Mass moment of inertia about hinge location. . . . . . . . . . . . . . . . . . . . . . . . . .C-10 Figure E-1 Thermodynamic properties of water. Temperature as a function of pressure and entropy for a range of pressure and entropy that emphasizes subcooled conditions.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-4 Figure E-2 HEM mass flux as a function of entropy and stagnation temperature for a range of entropy, which emphasizes subcooled stagnation conditions . . . . . . .E-5 Figure E-3 HEM mass flux as a function of stagnation pressure and stagnation temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-6 Figure E-4 Composite target pressure contours . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-7 Figure E-5 Jet zone of influence and expansion for circumferential break with full separation in containment vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-9 Figure E-6 Jet Impingement on flat plate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-13 Figure E-7 Expanding jet impingement on a flat plate . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-14 Figure E-8 Expanding jet impingement on a cylinder. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-14 Figure E-9 Reactor coolant system injection and discharge line breaks at the containment vessel head nozzles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-15 Figure E-10 Pressurizer spray and reactor pressure vessel degasification line breaks at the containment vessel head nozzles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-16 Figure E-11 Reactor coolant system injection and discharge line breaks at the reactor pressure vessel shell nozzles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-17 Figure E-12 Pressurizer spray and reactor pressure vessel degasification line breaks at the reactor pressure vessel head nozzles . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-18 Figure F-1 Characteristic shape of a blast wave and decay with time. . . . . . . . . . . . . . . . . F-2

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 List of Figures Figure F-2 Blast wave reflection coefficient . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-4 Figure F-3 Verification and validation case 8 results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-9 Figure F-4 Simplified NuScale Power Module-160 containment vessel model showing break locations and key structures, systems, or components . . . . . . . . . . . . . F-12 Figure F-5 Cutaway view of the mesh in the center of the model (Case 1) . . . . . . . . . . . . F-13 Figure F-6 Detailed view of the mesh around the pipe break (Case 1) . . . . . . . . . . . . . . . F-14 Figure F-7 Time history of total forces on key structures, systems, or components for containment vessel Case 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-16 Figure F-8 Absolute pressure contours at four time steps for containment vessel blast Case 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-17 Figure F-9 Absolute pressure contours for containment vessel blast Cases 2 & 3 . . . . . . F-18 Figure F-10 Modeled region of Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-20 Figure F-11 Geometry of part of one pipe gallery in Reactor Building showing break locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-21 Figure F-12 Geometry of part of one pipe gallery in Reactor Building showing break locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-22 Figure F-13 Identification of components in Reactor Building . . . . . . . . . . . . . . . . . . . . . . . F-23 Figure F-14 Cross-section view and close-up view of the mesh in case 1 . . . . . . . . . . . . . F-25 Figure F-15 Pressure contours for three time steps for Reactor Building blast Case 1 (Low Power) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-26 Figure F-16 Force time history for various structures, systems, or components for Reactor Building blast Case 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-27

Pipe Rupture Hazards Analysis TR-121507-NP Revision 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 Nuclear Regulatory Commission (NRC) regulations and guidance identification of essential and other SSC required to be protected from the effects of ruptures identification of postulated rupture locations characteristics of ruptures, including break types and size determination of potential external effects of high- and moderate-energy line ruptures 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 NPM bay inside the NuScale Reactor Building (RXB). The PRHA evaluation of the piping beyond the NPM bay in the RXB and through the balance of plant is the responsibility of the applicant that references the NuScale Power Plant US460 standard design.

The PRHA supports the NuScale US460 standard design per U.S. NRC Standard Review Plan, NUREG-0800, Chapter 3, Sections 3.6.1 through 3.6.3, and Branch Technical Position (BTP) 3-3 and BTP 3-4, which serve as appendices to SRP Section 3. Evaluations of the dynamic effects of HELBs are largely covered within this report; however, for environmental effects, this report primarily references other documents for evaluations and results.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 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 Power Plant (NPP) 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 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 essential structures, systems, or components (SSC) susceptible to postulated pipe rupture hazards. Examples of key 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 NuScale Power Module (NPM).

a small-volume, metal containment operated at a vacuum.

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 than the operating industry fleet.

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 results in only NPS 2 line breaks in the containment vessel (CNV) and no breaks under the bioshield in the NPM bay requiring evaluation of dynamic effects (i.e., blast waves, pipe whip, jet impingement). Consideration of nonmechanistic breaks of the main steam system (MSS) and feedwater system piping in the containment penetration area under the bioshield involves evaluation of environmental effects. Mitigation protection is demonstrated through separation and by the robustness and qualification of safety-related and essential SSC.

For the Reactor Building, evaluation of bounding high-energy line breaks (HELBs) and moderate-energy line breaks (MELBs) and cracks is performed to demonstrate that final piping design is 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-121507-NP Revision 0 1.0 Introduction 1.1 Purpose This document describes the NuScale Power, LLC (NuScale) methodology applicable to identification and assessment of pipe rupture hazards and the effects of pipe ruptures on the ability to achieve safe shutdown and cooldown. The following are addressed:

compliance with Nuclear Regulatory Commission (NRC) regulations and guidance identification of essential and other SSC required to be protected from the effects of ruptures identification of postulated rupture locations characteristics of ruptures, including break types and size determination of potential effects of high- and moderate-energy line ruptures identification of essential and other SSC required to be protected from the effects of ruptures criteria for showing acceptability of structures, systems, and components (SSC) exposed to those effects This evaluation addresses external effects of high-energy line breaks (HELBs),

moderate-energy line breaks (MELBs), and leakage cracks in piping in the NuScale Power Module (NPM) and Reactor Building (RXB). It does not discuss effects on core cooling and pressure forces on components internal to the system that has ruptured.

This report addresses the requirements for the as-designed Pipe Rupture Hazards Analysis Report as described in NRC inspection procedure Inspection of Pipe Rupture Hazards Analyses (Inside and Outside Containment) Design Acceptance Criteria (DAC)-Related ITAAC, Inspection Procedure (IP) 65001.21 (Reference 6.1.7) for piping located inside the containment vessel (CNV) and immediately outside, this report satisfies option 1 from Reference 6.1.7. After the plant is built, the as-built pipe rupture hazards analysis report Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) will verify that the as-designed PRHA is still valid.

1.2 Scope Pipe ruptures are addressed for each of four regions of the NuScale Power Plant (NPP) where high- or moderate-energy piping exists:

inside the CNV outside the CNV, in the NPM bay outside the NPM bay, in the RXB outside the RXB, in the remainder of the NPP

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 1.3 Abbreviations 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 CRDM control rod drive mechanism CRDS control rod drive system CVCS chemical and volume control system DAC design acceptance criteria DC direct current DEGB double-ended guillotine break DHRS decay heat removal system ECCS emergency core cooling system EDAS augmented DC power system EQ equipment qualification FWS feedwater system GDC General Design Criteria HELB high-energy line break ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LBB leak-before-break L/D length to diameter ratio LWR light water reactor M&E mass and energy 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 RMS root-mean-square RP reactor pool

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table 1-1 Abbreviations (Continued)

Term Definition RPV reactor pressure vessel RRV reactor recirculation valve RVV reactor vent valve RXB Reactor Building SG steam generator SMR small modular reactor SRP Standard Review Plan (NUREG-0800)

SSC structures, systems, and components SSE safe shutdown earthquake SST shear stress transport UHS ultimate heat sink ZOI zone of influence

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table 1-2 Definitions Term Definition Acceptable interaction A pipe rupture interaction for which, from a systems standpoint, the net required safety functions for a particular rupture are not impaired when assuming a single active component failure.

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

Blast wave A shock wave; namely, 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 design does not have Class 1E electrical power.

Essential systems As defined by Branch Technical Position 3-4, those systems necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without offsite 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 (e.g., pipe whip) and environmental (e.g., increased ambient pressure) effects.

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

Augmented DC Power The EDAS is responsible for providing a continuous, failure-tolerant source of System (EDAS) 125V DC power to assigned plant loads during normal plant operation and for a specified minimum duty cycle (mission time) following a loss of ac power. The design of the EDAS is intended to provide electrical system reliability substantially similar to that of a Class 1E DC power system. For that purpose, EDAS functional requirements have been defined based on requirements, regulatory guidance, and standards that would typically be applied to a Class 1E system.

(Reference 6.1.1)

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

Jet core The region immediately downstream of a break within which fluid striking a target would experience full recovery of the fluid stagnation pressure. The jet core is shown as Region 1 in Reference 6.1.17.

Jet impingement force The force imparted to an object because of 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 Systems that, during normal plant conditions, have: (a) maximum operating system temperature of 200 degrees F or less, (b) maximum operating pressure is 275 psig or less, and (c) minimum operating pressure above atmospheric pressure, or (d) 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.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table 1-2 Definitions (Continued)

Term Definition NuScale Power Module The assembly including the reactor pressure vessel, containment vessel, and all (NPM) directly attached components out to the outboard flange connecting module systems to those in the Reactor Building.

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

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

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.

Single active failure A single active failure is the failure of an active component to complete its intended function upon demand. In Reference 6.1.5 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 which 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 footnote states: 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 NPP 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 A terminal end is defined as 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 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 the zone of influence (ZOI) depends on the direct effect being evaluated (e.g., within physical reach of a whipping pipe of a given length, entire compartment for pressurization).

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 2.0 Background Design requirements for piping such as the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) (Reference 6.1.18) and the ASME B31.1 Power Piping Code (Reference 6.1.20) ensure the probability of a pipe rupture is low. Nonetheless, ruptures are postulated to occur in high- and moderate-energy piping so that their potential effect on the safe shutdown of the reactor can be evaluated. This report addresses the external effects of pipe ruptures on adjacent SSC required for safe shutdown. The report does not address direct effects on core cooling and pressure forces on components internal to the system that has ruptured.

The external consequences of ruptures depend on the thermodynamic conditions in the system at the rupture location and the interaction with the surroundings. The NRC has issued guidance and criteria for determining break locations and assumptions used in assessing the consequences. This guidance, which is reviewed further in Section 2.2, identifies the following external effects of ruptures that must be considered:

Dynamic Effects:

blast waves jet reaction loads jet impingement (including feedback amplification and resonance effects) dynamic subcompartment pressurization pipe whip impact Environmental Effects:

pressure, temperature, and humidity effects spray wetting flooding Additional background provided in subsequent sections includes a discussion of design features especially relevant to pipe rupture hazards, regulatory requirements with some discussion of their historical application, and the approach for regulatory compliance.

2.1 NuScale Design Features Relevant to Pipe Rupture Hazards Analysis The design is an integral, multi-module, small modular reactor (SMR) for which safety is provided by passive cooling 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 design specific review standards (DSRSs) to address what is directly applicable for the NuScale design. Examples of relevant design differences are:

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

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The NPMs are partially immersed in a large pool of water that serves as the ultimate heat sink (UHS) for the NPMs and does not require replenishment for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a design-basis accident.

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

Piping is small compared to the large reactors for which regulatory guidance is developed. Comparing AP1000 MSS conditions, volume per pipe, steam density, and specific density to NuScale's yield a factor of 17. In particular, the energy per foot of MSS pipe is generally less than 1/17th that of equivalent systems in large reactors.

The NPM containment is not a building. It is a pressure vessel designed and fabricated to ASME Code Section III Class 1 requirements.

Piping of the NPM, including secondary system piping, is made of corrosion resistant stainless steel (Type 304 or Type 316).

Main steam (MS) and feedwater (FW) piping inside containment boundary is designed to reactor coolant system (RCS) design pressure and temperature.

Ambient conditions inside the CNV are maintained at less than 1 psia.

Equipment and piping inside the NPM containment are not covered by insulation, which is important because

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

- detection of small leakage cracks is not impeded by retention of moisture in insulation.

- the bare piping can be readily inspected, because insulation does not need to be removed to note deposits, discoloration, or other signs of degradation.

- potentially corrosive substances (e.g., chlorides) cannot be trapped and held in contact with the piping surface.

Essential components inside the NPM containment are qualified to be functional after exposure to saturated steam at containment design pressure up to 1200 psia, requiring designs that are robust.

The small containment results in congestion that makes difficult addition of traditional piping restraints and the separation of essential components from break locations, but whipping pipes have a limited range of motion before encountering an obstacle.

The containment isolation valves (CIVs) are outside of containment. Where two valves in series are required (i.e., General Design Criteria (GDC) 55 and 56), both are in a single-piece valve body (i.e., no piping or welds between CIVs, which precludes breaks in between). Except for the MS lines, there is also a containment isolation test fixture (CITF) valve between the CIVs and the CNV nozzles, but there is no physical piping (i.e., valves are directly welded together and to the safe ends).

Active containment pressure suppression is not required.

During refueling, the NPM is disconnected from supporting systems (i.e., removal of piping spools), transported by crane to a refueling location, and disassembled. This

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 process provides access for inspection to portions of the plant not normally accessible.

Up to six 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 is remote from high-energy piping systems.

The considerations above demonstrate that pipe ruptures have a limited potential to adversely affect the safe shutdown of the reactor. Although the pipe rupture mitigation objective is consistent with that of large LWRs as described in NRC guidance, the unique features and passive safety attributes of the NuScale design involve other considerations as part of pipe rupture hazards analysis, as described in following sections.

2.2 Regulatory Requirements Regulatory requirements relevant to pipe rupture hazards are primarily derived from Environmental and Dynamic Effects Design Bases, GDC 4, of Appendix A to 10 CFR Part 50.

Regulatory guidance is provided in U.S. NRC Standard Review Plan (SRP),

NUREG-0800, Chapter 3, Sections 3.6.1 through 3.6.3 and in the many other documents referenced within those SRP sections. In particular, Branch Technical Position (BTP) 3-3 and BTP 3-4, which serve as appendices to SRP Section 3, are referenced extensively.

The following sections summarize important aspects of the documents above and discuss the design's compliance with the guidance. If certain information is not included in the summaries, it does not imply that the design does not comply with the guidance (note some guidance does not apply to the NuScale design); however, instances where the design does not directly comply with applicable guidance are identified and discussed.

Specifically, the two instances where the design does not directly comply with applicable guidance is the size of the nonmechanistic breaks specified in BTP 3-3 B.1.a(1)

(Section 2.2.4), and the definition of the containment penetration area in BTP 3-4 B.1(ii)

(Section 2.2.5).

2.2.1 Standard Review Plan Section 3.6.1 Plant Design for Protection against Postulated Piping Failures in Fluid Systems outside Containment, SRP 3.6.1, Rev. 3 describes design considerations for protecting essential equipment from adverse effects of pipe ruptures outside of containment. This document primarily points to BTP 3-3 for specific criteria, but provides additional clarification and guidance related to its application to traditional large reactors. The definition of essential systems from BTP 3-3 is clarified in SRP 3.6.1 Section III.4 to exclude safety equipment that is not necessary (i.e., essential) for safe shut down and mitigation for a particular pipe break event. Because much of the guidance from this document is repeated with more detail in BTP 3-3, it is instead discussed in Section 2.2.4.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 As described in Section 2.2.4, the design conforms to SRP 3.6.1 by following BTP 3-3.

2.2.2 Standard Review Plan Section 3.6.2 Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, SRP 3.6.2, Rev. 3 describes the methodology used to postulate pipe break and crack locations and configurations by generally referring to BTP 3-4 for detailed guidance. Therefore, that guidance is discussed in Section 2.2.5.

SRP 3.6.2 also gives detailed guidance for the evaluations of dynamic effects associated with HELBs (e.g., jet impingement and pipe whip), which are discussed further below.

Jet Thrust Guidance for representing jet thrust forces using both time dependent and steady state functions is given in SRP 3.6.2 Section III.4.C. These functions can be used as thrust loads on vessels and pipes (for whip analysis), and as a basis for the total jet impingement force.

The analysis applies a steady state function as described in Section 3.4.2 for pipe whip and jet thrust and impingement loads. For other analyses, such as system blowdown and subcompartment pressurization, which are not the focus of this report, a non-steady discharge based on the characteristics of the upstream reservoir may be used.

Jet Impingement Load Guidance for modeling jet impingement loads is in SRP 3.6.2 Section III.5. This guidance is based on American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard 58.2-1998 (Reference 6.1.17), which has been accepted by the NRC. However, in 2004, following interactions with the Advisory Committee on Reactor Safeguards (ACRS) on jet models, the staff determined that there were potential non-conservatisms in the models described in ANSI/ANS 58.2.

The non-conservatisms involve neglecting blast waves and dynamic loading feedback/resonance effects, and potentially incorrect jet expansion pressure distribution models and are discussed in Appendix A of SRP 3.6.2. The NRC has not incorporated detailed acceptance criteria for these complex phenomena into SRP 3.6.2; however, some guidance is now contained in Jet Impingement in High-Energy Piping Systems, NUREG/CR-7275.

The analysis applies a jet impingement load methodology based on the guidance in SRP 3.6.2 and ANSI/ANS 58.2 modified to address the potential non-conservatisms identified in Appendix A of SRP 3.6.2.

Blast Waves: Three-dimensional computational fluid dynamics (CFD) modeling are performed to bound the severity of blast waves, in locations with postulated HELBs.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Jet plume and zone of influence (ZOI): As noted in Appendix A of SRP 3.6.2, initial spreading rates are highly dependent on ratio of jet source to ambient conditions. Therefore, because NuScale has a unique containment environment (i.e., vacuum conditions) different methodologies are utilized depending on the environment where the HELB occurs. Both methodologies (inside CNV and outside CNV) use approaches selected to bound NRC accepted methodologies, as described in Section 3.4.3.3 and Appendix E.

Distribution of pressure within the jet plume: Inside the CNV, conservative pressures are assumed for steam jets ((2(a),(c) and NUREG/CR-2913 is used for two-phase jets, as described in Section E.3. Outside the CNV, the ZOI is assumed to be anywhere in a room containing a postulated break with impingement pressure being the full pressure at the break exit plane, multiplied by the appropriate thrust coefficient, as described in Section 3.4.2. Jet dynamic loading including potential feedback amplification and resonance effects: This phenomenon is not applicable to pipe breaks in the NPP, as discussed in Section 3.4.3.6 and Appendix B. Pipe Whip SRP 3.6.2 Section III.4 establishes that pipe whip analyses should show that pipe motions resulting from a HELB do not result in unacceptable impact upon or overstress of any safety-related or RTNSS B SSC to the extent that essential functions would be impaired or precluded. Pipe whip dynamic analysis criteria and piping system modeling considerations are discussed here, where jet thrust (SRP 3.6.2 III.4.c) is discussed in the above sections. Criteria are given to determine the need for a whip analysis, the initial condition assumptions if one is required, and if subsequent pipe ruptures should be considered for impacted (i.e., target) pipes. Several options for modeling pipe whip are described, although it is noted that other models may be considered if justified. The analysis uses a pipe whip methodology that is consistent with the SRP 3.6.2 Section III.4, as described in Section 3.4.4 and Appendix C. Detailed whip evaluations are used when applicable in the CNV and in the NPM bay in order to ensure that unacceptable impacts to essential SSC do not occur. In the Reactor Building (RXB) and outside the RXB in the remainder of the plant, where there are fewer essential SSC exposed to HELBs, detailed whip evaluations are generally not used and bounding impacts are assumed to occur. Multi-module effects due to pipe whip impacts on piping belonging to other modules is limited by the SRP 3.6.2 Section III.4 criteria that a whipping pipe does not cause breaks in piping of the same size or larger. Therefore, the additional and simultaneous effects due to a break in conservatively selected smaller pipe is always assumed. 2.2.3 Standard Review Plan Section 3.6.3 Leak-before-break Evaluation Procedures, SRP 3.6.3, Rev. 1 (Reference 6.1.4) provides guidance on performing a leak-before-break (LBB) analysis acceptable to © Copyright 2022 by NuScale Power, LLC 12

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 the NRC staff. The LBB analysis can be used to preclude the need to postulate HELB, and the consequent pipe whip restraints and jet impingement barriers (and analysis of dynamic effects), but still requires the consideration of cracks. The LBB methodology is not applied to the US460 standard design. The design instead eliminates postulated breaks in those portions of the MS and FW line by applying BTP 3-4 B.1(ii) containment penetration area design criteria. 2.2.4 Branch Technical Position 3-3 Protection against Postulated Piping Failures in Fluid Systems outside Containment, BTP 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. BTP 3-3 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) plant arrangements that separate fluid system piping from essential systems and components, including ensuring the MS and FW lines are not routed near the control room. enclosing fluid system piping within structures designed to protect nearby essential SSC, or enclosing essential SSC in protective structures. redundant design features that are separated or otherwise protected from postulated piping failures, or by designing or testing essential systems and components to withstand the effects associated with postulated piping failures. BTP 3-3 also provides guidance for analyzing the effects of pipe breaks outside containment, including certain assumptions such as offsite power being unavailable for certain events and consideration of a single active component failure in systems necessary to mitigate the rupture and shut down the reactor. Each pipe rupture should be considered separately as a single postulated initial event occurring during normal plant conditions. In addition to the pipe rupture locations specified in BTP 3-4 (Section 2.2.5), BTP 3-3 requires consideration of full circumferential breaks of non-seismic moderate-energy piping (because BTP 3-4 only applies during normal conditions, not seismic events). Although portions of the MS and FW lines may meet the criteria of BTP 3-4 B.1.(ii), eliminating the need to postulate pipe ruptures, BTP 3-3 requires the consideration of nonmechanistic breaks in these lines to ensure environmental effects of ruptures in these lines are considered. Plant arrangements generally separate fluid system piping from essential equipment outside containment. Most essential equipment is located either inside the CNV or in © Copyright 2022 by NuScale Power, LLC 13

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 the NPM bay, which is separated from the remainder of the plant piping by the pool walls. Essential equipment outside the NPM bay, inside the RXB, and in the remainder of the plant primarily consists of the module protection system (MPS) which comply with BTP 3-3 by the use of separation design criteria, as discussed in Section 3.2.2.3. Although the compact nature of NPM design requires the essential equipment to be located near the MS and FW line containment penetration areas, nonmechanistic breaks are evaluated to ensure essential equipment is designed to withstand the environmental conditions that would result from ruptures in those lines. The analysis deviates from the sizes of the nonmechanistic breaks specified in BTP 3-3 B.1.a(1), as discussed in Appendix I. Although a description of the plant safety features design and their response to a pipe rupture is not the focus of this report, it is noted that the NPM is able to passively reject decay heat to the UHS following design-basis events, with reactor trip and engineered safety feature actuation functions occurring without the need for safety-related alternating current (AC) or direct current (DC) power. Therefore, a loss of offsite power is not a threat to the safe shutdown the reactor in the event of a pipe rupture. Worst-case single active failures are considered in the thermal hydraulic and core neutronic response evaluations of pipe ruptures per the guidance in Reference 6.1.16. 2.2.5 Branch Technical Position 3-4 BTP 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, (Reference 6.1.6) provides guidance for selecting the location and characteristics of pipe ruptures including HELBs and leakage cracks. There are internal cross-reference errors in BTP 3-4 Revision 3. BTP 3-3 makes similar errors when referencing BTP 3-4. The following clarification is provided to explain how the analysis addresses these errors: Where BTP 3-4 or 3-3 references 2.A(ii), a reference to B.1(ii) is assumed. Where BTP 3-4 or 3-3 references 2.A(iii), a reference to B.1(iii) is assumed. High-Energy Fluid Systems BTP 3-4 B.1 provides specific guidance for postulating HELBs in high-energy fluid systems depending on the situation. Fluid systems separated from essential SSC (BTP 3-4 B.1(i)): Specific ruptures need not be postulated if it can be shown that a break anywhere in the system could not adversely affect essential SSC. Fluid systems in containment penetration areas (BTP 3-4 B.1(ii)): HELBs (including those at terminal ends) and leakage cracks need not be postulated in portions of piping from the containment wall to and including the inboard or outboard isolation valves provided that the piping design and stress analysis results satisfy the criteria listed in BTP 3-4 B.1(ii). Fluid systems in areas other than containment penetration (BTP 3-4 B.1(iii)): HELBs and leakage cracks are postulated to occur at areas of high stress or © Copyright 2022 by NuScale Power, LLC 14

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 fatigue usage as identified in the code stress analysis. The HELBs are also postulated at terminal ends. Alternatively, HELBs may be postulated at each fitting, welded attachment, and valve, and leakage cracks may be postulated at locations that produce the most severe environmental effects. Because the CIVs (and CITFs if applicable) are welded directly to CNV safe ends, the applicability of BTP 3-4 B.1(ii) to physical piping is explicitly limited to the decay heat removal system (DHRS) piping outside containment, which branches off of the MS safe-ends inboard of the main steam isolation valves (MSIVs). However, the design constraints applicable to an integral SMR differ from those of the large scale reactor designs for which the regulatory guidance was originally developed. Because the NPM has limited space for the installation of large pipe whip restraints and jet shields, the application of the conservative design and stress criteria of BTP 3-4 B.1(ii) is extended to the locations listed below (Section 3.2.2.4 and Appendix A provide further discussion): the outboard isolation valve to piping welds for high-energy lines from the containment penetration to the reactor pressure vessel (RPV) safe-end for the MS and FW lines the piping to safe-end welds inside the CNV for the DHRS condensate lines the emergency core cooling system (ECCS) main valve flanged connections to the RPV (although the specific design and stress criteria applied differs significantly from BTP 3-4 B.1(ii) as discussed in Section A.2 For other portions of high-energy systems inside containment and outside containment, inside the NPM bay not listed above, HELBs are postulated according to BTP 3-4 B.1(iii). NuScale piping design criteria require that high-energy lines inside containment and outside containment, inside the NPM bay satisfy the stress and fatigue criteria of BTP 3-4 B.1(iii)(1), (2), and (3), eliminating the need to postulate intermediate breaks. Therefore, for these areas, only terminal end HELBs are postulated inside the CNV and no HELBs are postulated outside the CNV, inside the NPM bay, because terminal end HELBs are eliminated by the application of BTP 3-4 B.1(ii). Leakage cracks in these areas are not selected based on stress criteria, but instead are either postulated to produce the most severe environmental effects or shown to be bounded by other design-basis events (DBEs). Outside the NPM bay, inside the RXB, HELB effects are assumed to occur anywhere in areas that house high-energy piping systems and leakage cracks are postulated at locations which produce the most severe results. Specific rupture locations are not needed for high-energy system piping outside the RXB, as this piping is considered to be separated from essential and other protected SSC per BTP 3-4 B.1(i). © Copyright 2022 by NuScale Power, LLC 15

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Moderate-Energy Systems BTP 3-4 B.2 provides guidance for postulating leakage cracks in moderate-energy fluid systems also using a methodology dependent on the situation. Fluid systems separated from essential SSC (BTP 3-4 B.2(i)): Specific leakage cracks need not be postulated if it can be shown leakage anywhere in the system could not adversely affect essential SSC. Fluid systems in containment penetration areas (BTP 3-4 B.2(ii)): Leakage cracks need not be postulated provided that the piping design and stress analysis results satisfy certain criteria. Fluid systems in areas other than containment penetration (BTP 3-4 B.2(iii)): Leakage cracks are postulated to occur at areas of high stress as identified in the code stress analysis. Moderate energy fluid systems in proximity to high energy fluid systems (BTP 3-4 B.2(iv)): Leakage cracks need not be postulated in moderate energy fluid systems when the resulting environmental effects are bounded by postulated ruptures in nearby high energy systems. For moderate-energy systems inside containment and outside containment, inside the NPM bay, leakage cracks are not postulated as their effects are bounded by ruptures in nearby high-energy systems (or by ECCS actuation inside the CNV). Outside the NPM bay, inside the RXB, leakage cracks are conservatively assumed to occur in areas that house moderate-energy systems at locations that result in the most severe effects. Specific leakage crack locations are not needed for moderate-energy system piping outside the RXB, as this piping is considered to be separated from essential SSC per BTP 3-4 B.2(i). Types of breaks and Cracks in Fluid System Piping Postulated pipe rupture locations are modeled using the guidance of BTP 3-4 B.3, except that outside the NPM bay, inside the RXB, circumferential breaks are always assumed (rather than longitudinal breaks) as this assumption results in a bounding analysis (Section 3.2.3.3). © Copyright 2022 by NuScale Power, LLC 16

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.0 Methodology This section discusses the methodology for demonstrating compliance with GDC 4 and addresses the related regulatory guidance for determining the location and characteristics of pipe ruptures and the analysis of dynamic external effects of pipe ruptures due to HELBs. The methodology for evaluating the environmental effects due to leakage cracks and MELBs is not the focus of this report; however, the applicable analyses are identified for reference. Figure 3-1 includes a flowchart depicting the process for evaluation of the effects of pipe ruptures. The flowchart addresses: Identification of essential SSC and other protected SSC Identification of high and moderate-energy systems in proximity to essential and other protected SSC and rupture locations Characterization of ruptures including type and size Determination of potential effects of ruptures including dynamic and environmental effects Criteria for determining the acceptability of those effects upon essential SSC The following sections describe this methodology in detail. © Copyright 2022 by NuScale Power, LLC 17

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure 3-1 Flowchart of methodology for evaluation of line breaks for each area that contains essential or protected structures, systems, or components ((

                                                                                               }}2(a),(c) 3.1      Identification of Essential and other Structures, Systems, and Components Requiring Protection BTP 3-3 (Reference 6.1.5) defines essential systems and components as those necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without offsite power. As mentioned in Section 3.2.1, essential systems do not include safety equipment that is not necessary for the safe shut down and mitigation of a particular pipe break event. Therefore, essential SSC in the design are a subset of SSC classified as safety-related.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 It should be noted that Revision 3 of SRP 3.6.2 introduced the topic of protection for regulatory treatment of non-safety (RTNSS) Category B SSC; however, the design does not include SSC of this classification. For systems outside the CNV and NPM bay, the following discussions are based on preliminary information to be confirmed or amended in a later revision of this document. 3.1.1 Essential Structures, Systems, and Components Systems that include safety-related SSC are described below. The simplicity and passive safety features of the 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. Containment System This system provides the primary containment for the reactor (i.e., final barrier to the release of fission products), and comprises the CNV, CIVs and other valves, piping, instrumentation, electrical penetration assemblies (EPAs), instrument seal assemblies (ISAs), and support structures. Piping belonging to the containment system (CNTS) connects systems inside the CNV to systems outside the CNV (e.g., connecting the RCS lines inside containment to the chemical and volume control system (CVCS) piping outside). The CNTS boundary ends at the flange where a pipe spool can be removed so that the NPM can be moved. Each CIV ((

                                                        }}2(a),(c)

The CNTS provides structural support for the RPV and other systems and is located in the NPM bay except for the CHPU skids, which are located outside the NPM bay in the RXB. The CNTS provides the following safety functions: provides a barrier to contain mass, energy, and fission product release by closure of the CIVs upon Containment Isolation signal provides a sealed containment and thermal conduction for the condensation of steam that provides makeup water to the RCS transfers core heat from the reactor coolant in Containment to the UHS provides safety-related signals provides structural support for safety-related SSC © Copyright 2022 by NuScale Power, LLC 19

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 provides electrical penetration assemblies for safety-related reactor instrumentation cables through the CNV provides the required pressure boundary for DHRS operation Control Rod Assembly The control rod assemblies (CRAs) are mechanically raised out of and lowered into the reactor core system (RXCS) by control rod drive mechanisms (CRDMs) located on top of the RPV. The rods provide reactivity control when required by plant conditions. The safety function of the CRA is to provide negative reactivity for reactor shutdown. Control Building The Control Building (CRB) houses the main control room and protects safety-related portions of the MPS including portions necessary for post-accident monitoring (PAM). The CRB is physically separate from the RXB except for a safety-related underground duct bank that runs between the two buildings. The CRB provides three safety functions. The CRB houses safety-related, risk-significant equipment and facilities pertinent to the operation and support of the NPMs. The CRB provides anchorage and support for safety-related, risk-significant equipment and facilities pertinent to the operation and support of the NPMs. The CRB protects safety-related, risk significant equipment and facilities from natural phenomena and externally generated missiles. Control Rod Drive System The purpose of this system is to physically raise and lower the CRAs. 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 released and fall into the core (i.e., scram) via gravity, achieving passive insertion of negative reactivity for reactor shutdown. Additionally, the CRDS pressure housing forms part of the RCPB. The CRDS provides two safety functions. The CRDS maintains the pressure boundary of the RPV, and ensures that during a reactor trip the CRDM does not prevent the CRA from fully inserting into the core. Chemical and Volume Control System The CVCS maintains RCS inventory during normal operation, provides purification and chemical injection to the RCS, provides pressurizer (PZR) spray, and supplies heated water to warm the RCS during startup. However, none of these functions are safety-related. Only the potential effect of CVCS misoperation on reactivity and the ability to shut down the reactor requires certain isolation valves, used to isolate boron © Copyright 2022 by NuScale Power, LLC 20

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 sources, to be classified as safety-related. Because a pipe rupture does not have to be postulated to occur simultaneously with a CVCS malfunction or operator error, SSC for which this is the only applicable safety function are not considered essential. Decay Heat Removal System The DHRS comprises two independent and redundant closed loop flow paths, each consisting of a steam generator (SG), a DHRS passive condenser mounted to the exterior of the CNV, and associated piping that provide natural circulation of secondary water flow from the SGs to the passive condensers, where heat is rejected to the UHS (reactor pool). During normal plant operation, the DHRS is in standby with flow blocked by closed DHRS actuation valves in the inlet line. These valves are maintained closed by energized hydraulic actuators. When power to the solenoids is interrupted, the valves (which are of a similar design to the CIVs, except that they passively open rather than passively close) open through the force of the stored energy device. Opening either of two DHRS actuation valves on one of the two redundant trains provides sufficient cooling. The DHRS provides the following safety functions: MPS actuation instrument information signals flow path from MS line through condensers to SG opens the DHRS actuation valves on a DHRS actuation signal 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 UHS. Water in the CNV flows back into the RPV by natural circulation. The ECCS also adds boron to recirculating coolant during ECCS actuation through the supplemental boron dissolvers and mixing is promoted by the lower mixing tubes.The ECCS consists of valves, both inside and outside the CNV, and connecting tubing and fittings, the dissolvers and the lower mixing tubes. Similar to the DHRS, the ECCS is started by interrupting power to the ECCS valves. The majority of this system is contained inside the CNV, except that the trip and reset valves are part of a single manifold located on the exterior of the CNV, partially immersed in the pool. The ECCS performs the following safety functions: depressurizes the RPV to allow recirculated coolant from the containment to enter the RPV for the removal of core heat provides recirculated coolant from containment to the RPV for the removal of core heat © Copyright 2022 by NuScale Power, LLC 21

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 opens the reactor vent valves (RVVs) or reactor recirculation valves (RRVs) when power is removed from their respective trip valves provides a portion of the RCPB for maintaining the RCPB integrity provides a portion of the containment boundary for maintaining containment integrity provides boric acid to recirculated coolant from containment to ensure core remains shutdown for design basis events provides low temperature overpressure protection for maintaining the RCPB integrity Of the above safety-related functions, providing low temperature overpressure protection is not considered essential. This function is only applicable during startup when the RPV is at a low temperature in order to prevent brittle fracture if exposed to high pressures. It is not required to shutdown the reactor in response to a pipe rupture. In-Core Instrumentation system The in-core instrumentation system (ICIS) continuously monitors the neutron flux distribution within the reactor core and provides core temperature information for monitoring core cooling during normal operations and post-accident conditions. The ICIS stringer assembly outer metallic sheath forms part of both the RCPB and the primary containment pressure boundary. The ICIS provides two safety functions. The ICIS provides equipment to accomplish a leak-tight primary containment pressure boundary at the CNV, and provides equipment to accomplish a leak-tight RCPB at the RPV. Module Protection System The MPS provides instrumentation and controls to detect accident situations and initiate engineered safety features when required. Each NPM has a dedicated (i.e., not shared with other modules) MPS. The MPS maintains components in an energized, or non-actuated, state during normal operation. The MPS components are located in the RXB, the CRB, and the safety-related underground tunnel between the RXB and CRB. The MPS provides the following safety functions: removes electrical power from the PZR heaters on a PZR heat trip actuation signal removes electrical power from the trip solenoids of the RVVs on an ECCS actuation signal removes electrical power from the trip solenoids of the RRVs on an ECCS actuation signal © Copyright 2022 by NuScale Power, LLC 22

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 removes electrical power from the trip solenoids of the DHRS Actuation Valves on a DHRS actuation signal removes electrical power from the trip solenoids of the MSIVs, MS isolation bypass valves, FWIVs, and FW regulating valves on a DHRS actuation signal removes electrical power from the trip solenoids of the CIVs on a containment system isolation actuation signal removes electrical power from the trip solenoids of the CVCS CIVs on a chemical and volume control isolation actuation signal removes electrical power from the trip solenoids of the CVCS demineralized water supply isolation valves on a demineralized water system isolation actuation signal (However, this function is not considered essential.) removes electrical power to the CRDS for a reactor trip provides power to the CNTS sensors provides power to the MS pressure sensors provides power to the RCS sensors removes electrical power from the trip solenoids of the MSIVs, MS isolation bypass valves, FWIVs, and FW regulating valves on a Secondary System Isolation actuation signal provides power to the position sensors on the FWIVs removes electrical power from the trip solenoids of the RVVs on a low temperature overpressure protection actuation signal (However, this function is not considered essential.) provides power to the RCS sensors NMS - neutron monitoring system The NMS provides neutron flux data for reactor trip, operating bypasses, and actuation of the MPS and information signals for PAM. The NMS provides two safety functions. The NMS provides neutron flux measurement information for startup, normal operations, shutdown, various reactor trips, operating bypasses, and actuations. The NMS also provides a positioning mechanism to maintain the neutron detector assemblies in a fixed position relative to the CNV during normal operations, anticipated operational occurrences, and design-basis events. Reactor Building Components The Reactor Building components (RBCM) consist of five components housed within the RXB: dry dock gate, RXB equipment door, bioshields, pool liner (including NPM supports), and over-pressurization vents. The overpressure vents relieve and direct © Copyright 2022 by NuScale Power, LLC 23

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 pressure increases from design-basis accidents in order to maintain the RXB internal pressure below design allowable limits. The RBCM provide two safety functions. The RBCM over-pressurization vents provide a means to vent air pressures internal to the RXB that result from a high-energy line break, and the NPM supports provide safety-related anchorage and structural support to the NPM. Reactor Coolant System The RCS primary loop 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 PZR baffle plate, passes downward around the tubes of the SGs where it is cooled, and then passes outside the core to reach the bottom of the RPV to rise again through the core. Portions of the RCS that are part of the RCPB include the RPV, reactor safety valves, and small bore piping (NPS 2), which are functionally part of the CVCS. The RCS provides the following safety functions: removes heat to ensure core and fuel thermal design limits are not exceeded provides instrument information signals for MPS actuation supplies the RCPB and a fission product boundary via the RPV and other components and appurtenances provides RPV overpressure protection provides support for safety-related components mixes a soluble neutron poison provides instrument information signals for low-temperature overpressure protection actuation Of the above safety-related functions, providing RPV overpressure protection is not considered essential. This function is to prevent overpressure of the RPV as required by code, but is not required to shut down the reactor in response to a pipe rupture. Also, providing instrument signals for low temperature overpressure protection actuation is not essential. Reactor Building Portions of the RXB are essential as they are relied upon to provide support and protection for safety-related SSC including the NPM, MPS, and the UHS pools. The RXB provides the following safety functions: houses safety-related, risk-significant equipment and facilities pertinent to the operation and support of the NPMs © Copyright 2022 by NuScale Power, LLC 24

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 provides anchorage and support for safety-related, risk-significant equipment and facilities pertinent to the operation and support of the NPMs protects safety-related, risk-significant equipment and facilities from natural phenomena and externally generated missiles protects safety-related, risk-significant equipment and facilities from internal events and internal generated missiles houses safety-related, not risk-significant equipment and facilities provides anchorage and support for safety-related, not risk-significant equipment and facilities protects safety-related, not risk-significant equipment and facilities from natural phenomena and externally generated missiles protects safety-related, not risk-significant equipment and facilities from internal events and internal generated missiles houses and protects spent fuel Reactor Core System The reactor core system (RXCS) includes the reactor fuel and supporting structures and provides the energy generation source for the RCS and SGS. The RXCS provides the following safety functions: contains fission products and transuranics within the fuel rods to minimize contamination of the reactor coolant maintains a coolable geometry under normal operating and design-basis event conditions provides control rod guide tubes to receive and align the CRA Steam Generator System The SGS transfers heat from the reactor coolant to produce superheated steam while providing a leak-tight pressure boundary between the primary reactor coolant and the secondary side of the SGS. Additionally, the SGS removes residual and decay heat from the reactor core in conjunction with the DHRS following DHRS actuation. The SGS includes FW and MS piping that connects to the CNTS. The FW lines include thermal relief valves to protect the piping from over pressurization during water solid conditions. The SGS provides two safety functions. The SGS supplies part of the RCPB, and transfers heat and provides steam for passive cooling via DHRS. © Copyright 2022 by NuScale Power, LLC 25

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 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 RP, refueling pool (RFP), and spent fuel pool (SFP). The water volume in the RP and RFP portions of the UHS is connected with the water volume in the SFP via the space above the top of the UHS weir wall. The UHS has capacity to remove decay heat following shutdown of six NPMs operating at full power for a minimum of 72 hours and a full SFP. Rather than a safety-related backup water supply, the NPP relies on the volume of water in the UHS to provide a safety-related source of cooling. No electric power is required to fulfill the UHS safety functions. The UHS provides the following safety functions: acts as heat sink to remove heat from NPMs and fuel assemblies removes decay heat from the spent fuel via direct water contact with the spent fuel assemblies provides iodine scrubbing for contents of the RP, RFP, and SFP via the surrounding water provides borated water for reactivity control during refueling 3.1.2 Other Structures, Systems, and Components Requiring Protection Beyond essential SSC, the design is also evaluated for the capability to maintain long-term PAM and the plant augmented DC electrical power system (EDAS). In addition to powering PAM, availability of EDAS ensures that, immediately following a loss of off-site power, multiple modules do not blow down simultaneously from de-energization of the ECCS valve solenoids. BTP 3-3 requires that postulated piping failures should not preclude the safe access to or habitability of the control room. This requirement is evaluated, even though NuScale shutdown and core cooling functions rely on natural forces (i.e., buoyancy driven natural circulation) or local energy storage (e.g., nitrogen accumulator to close CIVs), and initiates without human interaction. 3.1.2.1 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. The PAM information is displayed on the safety display and indication system. Post-accident monitoring does not have a capability to control equipment. © Copyright 2022 by NuScale Power, LLC 26

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Accident monitoring instruments are classified into at least one of the six following variable types: Type A, Type B, Type C, Type D, Type E, and Type F. The design has no Type A variables, which are those that provide information essential for the direct accomplishment of specific safety-related functions that require manual action. Additional variables are identified to provide backup or diagnostic information; however, these backup and diagnostic variables are not considered primary sources of accident information and are not required to meet the PAM design criteria. Design criteria for PAM instrumentation states that only PAM Type A, B, and C variables are required to be designed to meet single failure and common cause failure criteria. Type A, B, and C variables are required to be independent of and separated from the effects of a DBE. Type B and C variables are shown to be protected from the dynamic external effects of pipe breaks (e.g., pipe whip and jet impingement). The MPS and plant protection system, which receive, process, and transmit loads for PAM functionality, are required to ensure that only loads associated with PAM (plus ECCS for the MPS) are energized during a loss of AC power. Subsequent to a loss of AC power the EDAS battery mission time is 24 hours for A and D power channels and 72 hours for B and C power channels. Therefore, protection from pipe ruptures is ensured for channels B and C (for Type B and C variables) as those are the channels that remain available for the full 72 hours. The PAM variables that utilize instrumentation and components or cabling located inside containment are listed in Table 3-1. The PAM variables for the NPM bay, RXB, and Plant areas are listed in Table 3-2. Table 3-1 Post-accident Monitoring Instrumentation Sensors inside the Containment Vessel System Variable Type Location1, 2 ICIS Core Exit Temperature B, C Inside the RPV exiting at the RPV head Wide Range RCS Pressure B, C RPV upper head RPV shell near the top of the RCS Wide Range RCS THOT B downcomer RPV Riser Level B, C Inside the RPV exiting at the RPV head Narrow Range Containment Pressure B Upper CNV CNTS Wide Range Containment Pressure B, C Upper CNV Notes:

1. Instrument cabling (i.e., signal) is routed to penetrations on the CNV head or the CNV upper manway access shell.

2. The cabling for these instruments is routed through the CNV via EPAs or ISAs, through the NPM bay, and into the RXB.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table 3-2 Post-accident Monitoring Instrumentation Sensors outside the Containment Vessel System Variable Type Location1 CNTS CIV Positions B, C NPM Bay NMS Neutron Flux B NPM bay RM Inside Bioshield Area Radiation Monitor B, C NPM bay SDI Safety Display and Indication N/A 2 Plant Notes:

1. Location of sensor obtained from the applicable equipment list, piping and instrumentation diagram (P&ID), CAD model, or other design document.

2. This system only processes PAM signals, but are included for completeness.

3.1.2.2 Augmented Direct Current Power The EDAS is not an essential system; however, it provides augmented DC power to PAM and provides power to the ECCS valves. The location of the powered sensors is listed in Table 3-1 and Table 3-2. The EDAS equipment is located in the RXB and the CRB. 3.1.2.3 Control Room Habitability and Access The habitability of the control room is normally maintained by the control room HVAC system (CRVS) when AC power is available. The CRVS is not safety-related, but performs several important to safety functions including filtering radioactive contamination during a DBE, providing net positive ambient pressure for the CRB, and isolating the CRVS when necessary. The CRVS equipment is primarily housed in the CRB, although some non-important to safety equipment is located exterior to the building. During loss of normal AC power events, toxic gas releases, and radiological accidents, the habitability of the control room is maintained by the control room habitability system (CRHS). The CRHS equipment is in the CRB. The location of the CRB, which houses the control room is depicted in Figure 3-2. 3.2 Identification of High and Moderate Energy Systems in Proximity to Essential and other Protected Structures, Systems, and Components High- and moderate-energy lines are identified and located so that their proximity to the essential SSC identified in Section 3.1 can be determined. High- and moderate-energy are classified per the criteria listed below based on BTP 3-3 (Reference 6.1.5) and BTP 3-4 (Reference 6.1.6). High-energy fluid systems are those that, during normal plant conditions, have either or both a: (a) maximum operating temperature exceeding 200°F, or (b) maximum operating pressure exceeding 275 psig. © Copyright 2022 by NuScale Power, LLC 28

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Moderate-energy systems are those that, during normal plant conditions, have: (a) maximum operating temperature of 200°F or less, (b) maximum operating pressure is 275 psig or less, and (c) minimum operating pressure above atmospheric pressure, or (d) 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. High- and moderate-energy fluid systems in the design inside the CNV and the NPM bays are identified in Appendix H. For systems outside the CNV and NPM bay, the comprehensive list of systems will be added in a later revision of this document. However, some discussions of high-energy lines inside the RXB and yard are included based on preliminary information in order to demonstrate the ability of the plant to safely shut down in the event of a pipe rupture. Systems that are separated from essential and other protected SSC by distance or barriers need not be considered per BTP 3-4 B.1(i) and B.2(i). Additionally, for high- and moderate-energy lines that are in proximity to essential and other protected SSC, pipe ruptures need not be postulated if the criteria of BTP 3-4 B.1(ii) and B.2(ii) are satisfied, respectively. An LBB methodology is another option to eliminate the postulation of breaks in high-energy lines; however, as mentioned in Section 2.2.3, the US460 design does not utilize LBB. 3.2.1 Moderate-Energy Systems As discussed in Section 1.2 this report primarily addresses the external dynamic effects associated with ruptures in high-energy lines. A detailed accounting of moderate-energy systems is available in the applicable equipment qualification (EQ) or flooding evaluations. In these evaluations the following considerations apply to moderate-energy lines. Ruptures in moderate-energy lines are evaluated for the following environmental effects: pressure, temperature, and humidity effects spray wetting flooding Ruptures in moderate-energy lines include leakage cracks per BTP 3-4 B.3(iii) in seismically analyzed piping and full circumferential breaks in non-seismic piping. Per BTP 3-4 B.2(iv), leakage cracks in moderate-energy lines need not be postulated in areas in which a break in a nearby high-energy line results in more limiting environmental effects. For this reason, the moderate-energy lines that are located in the CNV and the NPM bay areas (as identified in Appendix H), are not evaluated for leakage cracks. © Copyright 2022 by NuScale Power, LLC 29

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.2.2 High-Energy Systems 3.2.2.1 High-Energy Systems inside the Reactor Building (Containment Vessel, NuScale Power Module bay, and Reactor Building areas) The following sections describe the piping in high-energy systems that could experience a rupture in the CNV, NPM bay, and RXB areas. The characteristics of these systems are summarized in Table 3-3. Reactor Coolant System The RCS is wholly contained within the CNV and has no loop or other large piping to rupture. However, inside the CNV, lines that are functionally part of the CVCS are included in the RCS. These lines run between the RPV nozzles and the CNV nozzles for PZR spray (one line from the CNV penetration tees into two lines to the RPV penetrations), RPV high point degasification (hereafter, just degasification or degas), discharge, and injection. The RCS lines are NPS 2 Schedule 160. The degasification line is normally isolated with its CIVs closed during operation. Welds and fittings are minimized through use of pipe bends. Containment System Piping belonging to the CNTS connects systems inside the CNV to systems outside the CNV (e.g., connecting the RCS lines inside containment to the CVCS piping outside). The CNTS boundary ends at the flange where a pipe spool can be removed so that the NPM can be moved. Each line that is connected to the RCS or is open 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 safe end. The CNTS-reactor component cooling water also has an identical two-CIV configuration, despite being a closed loop inside containment. There is an additional CITF valve between the CIVs and CNV nozzles, but there is no physical piping between the CIVs and nozzles. The MS and FW lines are closed loop systems inside containment and utilize a single CIV in accordance with GDC 57. Because of the need for DHRS steam supply connections, the main steam CIVs are each separated from the CNV by safe ends tee fittings. There is an additional CITF valve between the feedwater CIVs and CNV nozzles, but there is no physical piping between the CIVs and nozzles. In addition to the CIVs and CITFs, each normally open line directly connected to the RCS has an additional isolation valve (check valve, air operated valve, or solenoid valve) outboard of the CIV body. Each feedwater CIV body also contains an integral check valve that shuts upon flow reversal caused by a FW line break outside the NPM. © Copyright 2022 by NuScale Power, LLC 30

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The CIV actuators utilize pressurized hydraulic fluid contained in tubing. Although these non-process lines can represent hazards for personnel safety or become a fire source, the external effects of ruptures in these lines are not considered hazards to essential SSC as they are small in size and do not produce significant dynamic effects. Chemical and Volume Control System The CVCS includes the RCS-connected lines off the CNV, and consists of the PZR spray, injection, discharge, and degas. In the NPM bay, the lines are NPS 2 1/2, 2, and 1/2 and run between the inboard disconnect flanges and the pool wall. For PZR spray, injection, and discharge, NPS 2 flanged piping spool pieces with ball joints provide the capability to disconnect the NPM from system piping in preparation for movement for refueling. The degas line reduces from NPS 2 to NPS 1/2 in the CNTS portion of the line with the remaining CVCS degas piping being NPS 1/2. From the NPM bay, the CVCS pressurizer spray, injection, and discharge piping is routed down through a vertical pipe chase to the CVCS heat exchanger rooms on the 55 ft elevation, and then to the ion exchange and reactor coolant filter room on the 25 ft level. After purification, the piping runs to the recirculation pumps on the 40 ft elevation, continues back to the regenerative heat exchangers on the 55 ft elevation, and then finally back to the NPM bay. Emergency Core Cooling System The ECCS has no physical piping, other than the trip/reset lines, which are small (i.e., less than 1 inch diameter), and supplemental boron components that are open to the CNV environment (i.e., not high- or moderate-energy). Because ruptures in lines NPS 1 and smaller do not need to be postulated, the ECCS tubing is not discussed further in this report. Though not a pipe rupture and not within the scope of this report, a design-basis blowdown for the RCS and CNV is an inadvertent opening of an ECCS valve. The ECCS main valves are directly bolted onto the RPV and do not including physical piping. However, for the purposes of pipe rupture hazards analysis, the ECCS main valve flanged connections are considered piping. These connections are discussed Appendix A. Steam Generator System The SGS is the in-containment SG and MS and FW piping. In the CNV, four NPS 8 Schedule 120 SGS steam lines from the SG outlet plena connections on the RPV merge into two NPS 12 Schedule 160 steam lines to the CNV safe ends. The two NPS 5 Schedule 120 feed lines from the CNV safe ends split into two NPS 4 Schedule 120 lines (total of four) that supply feed water to the SG plena. Just upstream of the split, the DHRS return line tees in. There are thermal relief © Copyright 2022 by NuScale Power, LLC 31

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 valve connections on both NPS 5 lines; however, these connections are not larger than NPS 1. Main Steam System In the NPM bay, two MSS 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. In the RXB, the MS piping is routed through the pipe galleries on the 100 ft elevation and then through the outer wall into the yard. The main piping remains NPS 12 Schedule 160, however there are smaller NPS 4 bypass lines in the piping galleries as well. Feedwater System In the NPM bay, two flanged NPS 4 piping spools incorporating ball joints provide the capability to disconnect the NPM in preparation for movement for refueling. The lines are NPS 4 stainless steel in the containment penetration area and up to the NPS 4x6 reducer near the NPM bay wall. Beyond and including the reducer, the piping is chrome-moly. A dissimilar metal weld joins the stainless pipe to the chrome-moly. In the RXB, the FW piping is routed through the pipe galleries on the 100 ft elevation and then through the outer wall into the yard. The main piping remains NPS 6 Schedule 160; however, there are smaller NPS 4 bypass lines to the MSS in the piping galleries as well. Decay Heat Removal System In the NPM bay, an NPS 4 Schedule 120 line runs from each MSS safe-end tee inboard of the MSS CIV, splits into two parallel lines (four lines total with two per DHRS train), through normally closed, fail open, 4-inch DHRS actuation valves, before rejoining and continuing to the passive condenser. Cool water from the passive condenser is returned to the FW pipe by a NPS 2 Schedule 160 line located inside the CNV and NPS 2 Schedule 80 line inside the bay submerged in the pool. Note that the line expands to NPS 4 to connect to the exterior CNV safe-end. Module Heatup System The module heatup system (MHS) is the only high-energy system in the RXB to serve all NPMs, via connection to each NPMs CVCS. The MHS mechanical equipment and piping is located on the 55 ft elevation, with the largest pipe size being NPS 3. © Copyright 2022 by NuScale Power, LLC 32

Table 3-3 High energy system piping characteristics for piping in the Containment Vessel, NuScale

    © Copyright 2022 by NuScale Power, LLC Power Module Bay, and Reactor Building Max. Operating1        Design Seismic System   Location     Line Function   Press. Temp     Press. Temp. Largest Piping       Material         Piping Code     Category (psia3)    (ºF)    (psia3)    (ºF)

Degas 2000 635.8 2200 650 NPS 2 Sch. 160 stainless steel ASME NB I PZR Spray 2000 635.810 2200 650 NPS 2 Sch. 160 stainless steel ASME NB I RCS CNV Injection 2000 5456 2200 650 NPS 2 Sch. 160 stainless steel ASME NB I Discharge 2000 518 2200 650 NPS 2 Sch. 160 stainless steel ASME NB I ASME III ND / Degas 2067 641 2200 650 NPS 2 Sch. 160 stainless steel I / II B31.1 ASME III ND / PZR Spray 2123 5456 2850 650 NPS 2 Sch. 160 stainless steel B31.1 I / II CVCS / CNTS NPM bay2 ASME III ND / Injection 2123 5456 2850 650 NPS 2.5 Sch. 160 stainless steel B31.1 I / II ASME III ND / Discharge 2067 518 2200 650 NPS 2.5 Sch. 160 stainless steel I / II B31.1 CVCS RXB Various 21775 5455,6 28225 5755 NPS 3 Sch. 160 stainless steel B31.1 II / III FW 701 250 2200 650 NPS 5 Sch. 120 stainless steel ASME III NC I SGS CNV MS 700 575 2200 650 NPS 12 Sch. 160 stainless steel ASME III NC I MSS / CNTS NPM bay2 MS 700 575 2200 650 NPS 12 Sch. 160 stainless steel B31.1 I stainless steel / MSS RXB MS 700 575 2200 650 NPS 12 Sch. 80 B31.1 I / II / III chrome moly FWS / chrome moly / CNTS NPM bay2 FW 701 250 2212 650 NPS 47 Sch. 120 stainless steel B31.1 I FWS RXB FW 701 250 2212 300 NPS 6 Sch. 808 chrome moly B31.1 I / II Pipe Rupture Hazards Analysis TR-121507-NP 33 Revision 0

Table 3-3 High energy system piping characteristics for piping in the Containment Vessel, NuScale

    © Copyright 2022 by NuScale Power, LLC Power Module Bay, and Reactor Building (Continued)

Max. Operating1 Design Seismic System Location Line Function Press. Temp Press. Temp. Largest Piping Material Piping Code Category (psia3) (ºF) (psia3) (ºF) NPM bay Steam 700 575 2200 650 NPS 4 Sch. 1204 stainless steel ASME III NC I Condensate 9 4 DHRS NPM bay 700 120 2200 650 NPS 4 Sch. 120 stainless steel ASME III NC I return Condensate CNV 700 250 2200 650 NPS 2 Sch. 80 stainless steel ASME III NC I return Hot water for MHS RXB 2177 450 2822 575 NPS 3 Sch. 160 stainless steel B31.1 III NPM heat-up Notes: 1. Conditions are selected to bound applicable system line list or values for low power conditions (102 percent to 20 percent corresponding to normal operating load following and 1 percent power to represent hot shutdown) shown in Appendix A and Appendix B. T Cold is used for discharge line temperatures and FW temperature and steam temperature are used for the FW/DHRS condensate and MS/DHRS steam lines respectively.

2. For lines in the NPM bay area, the values bound those from the CNTS line list and the other system line lists.

3. When converting from PSIG to PSIA, 12 psi is used.

4. Assumes that the NPS 4 DHRS piping is Schedule 120.

5. Higher pressures and temperatures exist in the CVCS, but only in lines NPS 1 and smaller. There are also Seismic Category I lines in the CVCS, but they are also NPS 1 or smaller.

6. Hot shutdown conditions also considered

7. The transition from NPS 6 to NPS 4 is located in the wall penetration, behind the plate guide support. Therefore, the largest pipe size in the bay is NPS 4.

8. The line list for the condensate and FWS does not list pipe schedule.

9. Conservative maximum pool temperature used. Note that temperatures during DHRS operation do not apply here because, once DHRS is operating the NPM is passively cooled (i.e., no longer in hot shutdown).

10. The PZR spray line generally has the same subcooled conditions (i.e., not saturated) as the injection line. However, the line list uses bounding values.

Pipe Rupture Hazards Analysis TR-121507-NP 34 Revision 0

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.2.2.2 High-Energy Systems outside the Reactor Building (Plant Area) Because there are generally no essential SSC outside the RXB in proximity to these lines (Section 3.2.2.3.2), postulated rupture locations and evaluations of external dynamic effects are not needed per BTP 3-4 B.1(i). Therefore, detailed descriptions of these systems are not provided at this stage in the design. However, a comprehensive list of high-energy lines outside the RXB will be developed to verify this analysis. 3.2.2.3 Separation Criteria for High-Energy Systems Separation is a means demonstrating that essential and other SSC are protected from the effects of HELBs. Separation from external dynamic effects may be achieved by: Barriers: Isolation of essential SSC from high- and moderate-energy piping by placement in different compartments of the plant. Barriers are discussed in Section 3.2.2.3.1. Distance: If the essential SSC are distant from the rupture location, it may be possible to show that there are no effects of blast, pipe whip, and jet impingement. 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. Additionally, intervening obstacles, depending on the obstacle, may separate SSC from certain dynamic effects. An example is a larger or equal sized pipe restraining the whipping of another pipe from reaching essential SSC, but not protecting it against blast or jet effects. Pipe whip restraints and jet shields are considered intervening obstacles and are discussed in Appendix I. 3.2.2.3.1 Barriers For SSC to act as barriers the following criteria must be met: Per BTP 3-3 B.2.b The SSC are designed to Seismic Category I requirements

                           -   The SSC are designed to withstand the effects of a postulated pipe rupture in combination with the SSE Per BTP 3-4 B.1.(iii)(4):
                           -   For high-energy lines serving as barriers, SSC are designed to withstand the consequences of the pipe break that produces the greatest effect on the SSC, regardless that the criteria identified in BTP 3-4 B.1(iii) might not require such a break location to be postulated.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Structures acting as barriers to protect against dynamic effects in the design include the RPV, the CNV, and RXB structures and components (RXB and RBCM). Other SSC may act as barriers only for environmental effects, such as HVAC dampers, but are not discussed in this report. The RPV and CNV are considered to act as barriers for dynamic effects of HELBs based on the extrapolation of guidance from SRP 3.6.2, which 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, SSC of equal or larger diameter and equal or greater wall thickness do not only not leak or crack, but also obstruct further travel of the whipping pipe, protecting SSC farther away from being struck. Table 3-4 provides a comparison of potential whipping pipes and the RPV and CNV that are credited to act as barriers. The numbers in ( ) parentheses 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 RPV and CNV are considered to serve as barriers. The loads due to pipe whip and other applicable dynamic effects must be considered for inclusion in SSC load combinations as required by their design requirements. Table 3-4 Barriers in the containment vessel Component Pipe Size Outer Diameter Wall Thickness1 (in.) (in.) Whipping pipe NPS 2 (( RCS lines Schedule 160 SSC Barrier CNV N/A RPV N/A }}2(a)(c) ((

                                                                                         }}2(a),(c)

In the NPM bay and the RXB, concrete floors, walls, and ceilings can also serve as pipe whip barriers but require additional evaluations for pipe whip penetration and jet erosion. A concrete structure, whether reinforced concrete (RC) or steel-plate composite concrete, serves as a barrier to dynamic and environmental effects if it is not fully penetrated by the whipping pipe or eroded by the jet. Pipe evaluations are described further in Appendix C, while erosion is discussed below. © Copyright 2022 by NuScale Power, LLC 36

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Erosion due to continuing jet impingement is acceptable based on work of others; Reference 6.2.59 notes: to determine the dynamic impact and erosive effects of high temperature pressurized water and of steam jets from ruptured pipe lines, Westinghouse conducted a series of tests with subcooled water at 2250 psia/500 degrees F and with saturated steam at 1030 psia, released through nozzles of 3 different diameters, impinging on reinforced concrete structures, at various angles. Evaluation of the results indicates that erosion of concrete by a primary coolant or steam line break definitely does not impose a design consideration. WCAP-7391 is a proprietary Westinghouse report that NuScale does not have access to. However, Reference 6.1.24 provides further details: Westinghouse testing of subcooled jet impingement on bare concrete surfaces from a fluid supply of 2250 psig and about 550 degrees F is summarized in WCAP-7391 (proprietary). This testing, which was performed with subcooled water at pressures approximately twice those of the ABWR, showed no observable concrete damage at an L/D ratio of 3.2. The jet impingement pressure loading from the test conditions is compared to that of the ABWR using the ANSI/ANS 58.2-1988 jet expansion model. The model was used to calculate a stagnation pressure at PWR conditions (fluid conditions of 2250 psia and 540ºF and an L/D = 3.2) and compared to a stagnation pressure at ABWR conditions (fluid conditions of 1070 psia and 540ºF and an L/D = 2.67). The use of slightly more subcooled conditions is conservative as it provides for the jet to remain intact longer and results in a slightly higher stagnation pressure at a given, value for L/D. The results of these calculations are summarized below; For PWR conditions: Stagnation Pressure p =2250 psia, T =540 °F and L/D = 3.2 > 171 psig For ABWR conditions: Stagnation Pressure p = 1070 psia, T = 540 °F and L/D = 2.58 psig NuScale MSS conditions are slightly different, at 700 psia and 575 degrees F. Westinghouse MSS pipe diameter is 38 inches versus 12 inches for the NuScale design. At a distance of 1 length to diameter ratio (L/D) for the NPP, the stagnation pressure is below that for which no concrete erosion occurred in the Westinghouse testing. In the RXB, any MSS break exit within 1 L/D of a concrete surface would be directed at the pool or RXB outer wall, each of which is 4 feet thick and, therefore, has sufficient structural margin for erosion. © Copyright 2022 by NuScale Power, LLC 37

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Additionally, these walls are steel-plate composite construction (encased in steel plate reinforcement), further reducing the concern for erosion. 3.2.2.3.2 Distance The congested areas within the CNV, NPM bay, and RXB make it difficult to justify separation by distance without detailed evaluations of postulated pipe ruptures (i.e., for pipe whip, jet impingement). However, outside the RXB, where it is less congested, and there are fewer essential and other protected SSC, and justification of separation from dynamic external effects by distance is evaluated. This evaluation is based on preliminary information to be verified at a later design phase. The only essential SSC outside the RXB in the Plant area belong to the MPS and CRB. Additionally, per Section 3.1.2.1, these SSC include MPS components that serve as PAM variables Type B and C. Finally, as discussed in Section 3.1.2.3, components that ensure control room habitability that are part of the CRHS and CRVS are housed in the CRB. These SSC can be shown to be separated from the external dynamic effects of ruptures based on being remote from high-energy lines and potential HELBs with the exception of the piping associated with the high pressure breathing air storage bottles in the CRH itself. Figure 3-2 shows the layout of the yard area between the RXB (blue) and CRB (red). The high-energy MS and FW lines that run between the RXB and the Turbine Generator Building (not shown but to the right of the pipe rack) utilize the pipe rack highlighted in red. From the graphical scale, it can be shown that the CRB and enclosed essential or other protected SSC are at least 150 ft away from these high-energy lines. © Copyright 2022 by NuScale Power, LLC 38

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure 3-2 Proximity of High-Energy Systems to Control Building ((

                                         }}2(a),(c)

Additionally, to facilitate the routing of cabling for essential and other protected electrical and I&C systems (MPS, PAM, and EDAS) between the RXB and CRB, a safety-related tunnel is provided to separate these components from the potential effects of pipe ruptures and other outside hazards. This tunnel (purple) is shown below in Figure 3-3. This section demonstrates that separation by distance is implemented in the design outside the RXB; however, an evaluation using detailed design information will be performed at a later stage in the design to verify adequate separation has been implemented. © Copyright 2022 by NuScale Power, LLC 39

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure 3-3 Safety-Related Underground Duct Bank ((

                                                                                                    }}2(a),(c) 3.2.2.3.3              Redundancy Redundancy is used in system design to mitigate single active failures in the plant response to pipe rupture events, but is relied on less to mitigate the external dynamic effects of HELBs, which is the focus of this report.

Redundancy is typically not used to show separation for mechanical components or structures, but may be applied to Class 1E circuits and PAM Type B and Type C variables per design criteria as described in Section 3.2.2.3.4. © Copyright 2022 by NuScale Power, LLC 40

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.2.2.3.4 Separation for Cables For this report, detailed cable routing for the MPS, PAM Type B and C variables, systems that provide associated instrumentation, and associated portions of the EDAS are not considered and may not be complete at this stage in the design. However, appropriate separation is ensured by the use of design criteria for these systems. Plant design criteria require that Class 1E DC circuits in the CNV and in the NPM bay be qualified for environmental conditions but must be evaluated for pipe whip and jet impingement if routed unprotected within the ZOI. Class 1E DC 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 shall be 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 shall be considered. The potential hazard of missiles resulting from failure of rotating equipment or high-energy systems shall be considered. Protection of nonhazard and limited hazard areas from pipe failure hazard areas is accomplished by the use of barriers, restraints, separation distance, or 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 shall be limited to a single division. Where the pipe failure requires protective action, Class 1E equipment, associated circuits, or raceways shall not be 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, 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 shall be limited to the same division as the piping. Where the piping involved is not qualified for design-basis events, Class 1E equipment, associated circuits, or raceways shall not be located in the © Copyright 2022 by NuScale Power, LLC 41

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 area, except for those cables that must terminate at devices or loads within the area. 3.2.2.4 Break Exclusion for High-Energy Systems BTP 3-4 (Reference 6.1.5) B.1.(ii) identifies criteria for which ruptures need not be considered from the containment wall to and including the inboard or outboard isolation valves (usually referred to as the containment penetration area break exclusion zone). However, previous designs have extended this break exclusion zone to beyond piping in containment penetration areas in certain instances as long as the additional design criteria of BTP 3-4 B.1.(ii) are satisfied. Break exclusion areas in the design are described below. For lines that require two containment isolation valves, the NPP has both CIVs in a single valve body and a CITF welded directly between the CIVs and the CNV nozzle. There are no break locations between the valves. However, the weld between the valve bodies and the CNV safe end is equivalent to those to which break exclusion is intended to apply. Therefore, strict interpretation of the allowable extent of break exclusion would be limited in application in the design. However, this boundary is extended inside the CNV and NPM bay outside the CNV to include the areas listed below. Break exclusion zones are not used outside the NPM bay. The outboard weld at the CIV - Included to eliminate the need to postulate terminal end pipe breaks and dynamic effects in the area under the bioshield, where it would be difficult to place large whip restraints in a congested area. DHRS piping welds outside the CNV - Included as this piping branches off the CNTS main steam lines between the CNV and MSIV The SGS-MS and SGS-FW lines inside the CNV - Included to eliminate the need to postulate terminal end breaks for large lines inside the CNV, where it would be difficult to place large whip restraints in a congested area. The welds between the DHRS piping inside the CNV and the CNV safe-end - Included to eliminate the need to postulate breaks in the NPS 2 DHRS line which could impose large loads on the SGS-MS and SGS-FW lines, which are themselves in a break exclusion zone. The ECCS main valve flange connections - Included to eliminate postulated catastrophic failures of the ECCS main valve to RPV flanges. Appendix A provides detailed descriptions of these break exclusion zones including their adherence to the criteria in BTP 3-4 B.1.(ii). For the ECCS valve flanges, which are not piping and cannot use the criteria in BTP 3-4 B.1.(ii), a justification for why this connection provides confidence that the probability of gross rupture is extremely low is included in Section A.2. As discussed in Section 2.2.4, non-mechanistic breaks in the MS and FW lines are postulated in the portions of those lines classified as break exclusion zones in order to ensure essential SSC located in those areas are evaluated for © Copyright 2022 by NuScale Power, LLC 42

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 environmental effects. The non-mechanistic breaks are discussed further in Appendix H. 3.2.3 Postulated Pipe Rupture Locations in High-Energy Lines Pipe rupture locations are postulated in high-energy lines that are in proximity to essential or other protected SSC (i.e., separation is not demonstrated) and are not included in break exclusion zones. Pipe rupture locations (circumferential breaks, longitudinal breaks, and leakage cracks) in ASME Class 1, 2, and 3 high-energy lines are determined using the criteria of BTP 3-4 B.1.(iii) and B.3. This criteria requires, at a minimum, circumferential breaks be postulated in high-energy lines, larger than NPS 1, at piping system terminal ends. For piping that is stress analyzed, intermediate breaks and leakage cracks may be eliminated. For non-ASME piping, the criteria of BTP 3-4 B.1.(iii) may be used to determine rupture locations per BTP 3-4 B.1.(iii)(3), which is based on ASME Class 2 and 3 analysis requirements. For B31.1 piping that is stress analyzed, the following stress criteria based on B31.1 analysis requirements from Reference 6.1.20 may be used in place of the ASME Class 2 and 3 stress criteria:

a. Circumferential high-energy line breaks shall be postulated in high-energy piping, exceeding a NPS of one inch, at terminal ends and at intermediate locations where, Eq. 16 (excluding SSE) + Eq. 17 > Sc + 1.16

Sh

b. Longitudinal high-energy line breaks shall be postulated in high-energy piping, equal or larger than a NPS of four inches, at intermediate locations where, Eq. 16 (excluding SSE) + Eq. 17 > Sc+ 1.16

Sh

c. Cracks shall be postulated in high- and moderate-energy piping, exceeding a NPS of one inch, at intermediate locations where, Eq. 16 (excluding SSE) + Eq. 17 > 0.5 *Sc+ 0.58

Sh.

The application of separation, break exclusion zones, and rupture postulation criteria are discussed below in the context of each area. 3.2.3.1 Inside the Containment Vessel As discussed in Section 3.2.2.3.1, the CNV acts as a barrier that protects SSC inside the CNV from high-energy lines outside the CNV. Therefore, only HELBs in the high-energy lines listed as being located inside the CNV in Table 3-3 could affect SSC located inside the CNV. As discussed in Appendix A.1, break exclusion criteria is applied to the SGS-MS and SGS-FW lines, leaving only the NPS 2 RCS and non-terminal end portions of the DHRS condensate lines subject © Copyright 2022 by NuScale Power, LLC 43

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 to HELB postulation. Design criteria for these lines require that BTP 3-4 B.1.(iii) stress criteria be met, eliminating intermediate break locations. Therefore, the only HELBs inside containment are the terminal ends of the NPS 2 RCS lines. The effects of leakage cracks in these lines are bounded by ECCS actuation. Also as discussed in Section 3.2.2.3.1, the RPV acts as a barrier that protects SSC inside the RPV from high-energy lines outside the RPV. Because the entirety of the RXCS and CRA systems are located inside the RPV, they do not need to be considered further. This consideration also applies to certain components of other systems that are located inside the RPV including: the control rod drive shaft assembly, RCPB portions of ICI stringer assemblies, SG tubes, SG tube supports, RCPB portions of the feed plenum access ports, RCPB portions of the integral steam plenum, integral steam plenum caps, and flow restrictors. Additional considerations regarding the susceptibility of the remaining SSC exposed to potential HELB external dynamic effects are: RCS components that are part of the RCPB, such as the RPV, RPV supports, and RCS piping systems are robust. Their safety function, which is to maintain or support a pressure boundary, is not impaired by the dynamic effects of breaks and needs no further evaluation in this report. Nevertheless, SSC subjected to loads because of the external effects of pipe breaks are evaluated as applicable per their design specification. CRDM components that are part of the RCPB and are necessary to maintain a geometry to allow the control rods to drop freely are robust and able to act as barriers. Nevertheless, if subjected to loads because of the external effects of pipe breaks, the CRDMs are evaluated as applicable per their design specification to ensure functionality. Control rods fail to their safe position because of a loss of power or control signal to CRDS. The magnetic jack assemblies are located inside the CRDM pressure housings and are not susceptible to damage from HELBs. CNV components that are potentially exposed to damage are those that form part of the primary containment boundary including the EPAs and instrument seal assemblies. Most CNV components are robust and able to act as barriers, except that the EPAs and instrument seal assemblies are susceptible to damage. The ECCS valves, although robust, are susceptible to damage. If exposed to the effects of pipe breaks, the valves will be evaluated per their design specifications to ensure functionality. The tubing and fittings are especially susceptible to damage; however, upon failure (i.e., rupture), cause the ECCS main valves to fail in their safe-position. The ECCS main valves also go to their safe position because of a loss of power or control signal to ECCS. The ECCS venturis are located inside the ECCS main valve inlets and are not susceptible to damage from HELBs. The MPS is susceptible to physical damage, but perform its safety function if failure occurs. If more than one MPS indication of a type loses its signal because its cable is severed by pipe whip or jet impingement, a reactor trip or © Copyright 2022 by NuScale Power, LLC 44

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 safety component actuation occurs. For example, breaking signal lines for two or more RCS hot temperature or PZR level indications would cause a reactor trip. Nevertheless, protection is applicable for the MPS. Detailed cable and instrument routing is left to future design phases, and protection is demonstrated at this stage by the inclusion of design and routing requirements (Section 3.2.2.3.4). The ICI cable outer metallic sheaths, which form part of the RCPB when within the RPV and the primary containment boundary beyond the RPV compression fitting connections, are exposed to potential damage where they are routed between the RPV and CNV heads. However, their safety function is not likely to be impaired due to damage from external effects of breaks, because the cables are packed with mineral insulation that acts to mitigate leakage beyond the containment boundary. Nevertheless, protection is applicable for the ICIs. Detailed cable and instrument routing is left to future design phases, and protection is demonstrated at this stage by the inclusion of design and routing requirements (Section 3.2.2.3.4). Although not essential to ensure long term shutdown and core cooling, NuScale evaluated the availability of PAM indication following a pipe rupture in order to satisfy NRC guidance. Detailed cable and instrument routing is left to future design phases, and protection is demonstrated at this stage by the inclusion of design and routing requirements (Section 3.2.2.3.4). For protection against pipe whip and jet impingement inside containment where separation by large distances or barriers may not be possible, cables are routed clear of pipe whip paths and are at least 6.75 inches away from an RCS line terminal end break location to mitigate the effects of jet impingement. The cable is qualified by testing for the CNV environment including jet impingement effects, if applicable (i.e., routed unprotected within the ZOI of a jet from any source). 3.2.3.2 In the NuScale Power Module bay As discussed in Section 3.2.2.3.1, the RXB walls acts as a barrier that protects SSC inside the NPM bay from the external dynamic effects of high-energy lines outside the NPM bay (environmental effects are not protected against because of the wall penetrations). Also as discussed in Section 3.2.2.3.1, the CNV acts as a barrier that protects SSC inside the NPM bay from the effects of HELBs inside the CNV. Therefore, only HELBs in the high-energy lines listed as being located inside the NPM bay in Table 3-3 could affect SSC located inside the NPM bay. As discussed in Appendix A.1, break exclusion criteria are applied to the containment penetration areas of all high-energy lines, including the DHRS system piping. Additionally, break exclusion criteria are applied to the outboard weld of all CIVs (including the primary MSIV and FWIV) in order to ensure terminal end breaks are eliminated in the NPM bay. Design criteria for high-energy lines in the NPM bay classified as ASME Class 2 or 3 require that BTP 3-4 B.1(iii) stress criteria be met, eliminating intermediate break locations. © Copyright 2022 by NuScale Power, LLC 45

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 For other high-energy lines in the NPM bay, design criteria require that BTP 3-4 B.1(iii) stress criteria be met, eliminating intermediate break locations in these lines as well. Therefore, all HELB locations in the NPM bay area are eliminated, and there are no essential or other protected SSC in proximity to the external dynamic effects of HELBs. Nonmechanistic breaks in a MSS or FWS pipe in accordance with BTP 3-3 are evaluated as discussed in Appendix H, and effects of leakage cracks are evaluated; therefore, essential SSC located in the bay are only evaluated for environmental effects. Although, as discussed above, the pool wall generally protects SSC inside the NPM bay from the dynamic external effects of breaks inside the RXB, whip restraints are included on high-energy lines to prevent the lines from whipping through the penetration through the 4-ft-thick pool wall should a break occur at the terminal end outboard of the wall. 3.2.3.3 In the Reactor Building Systems that include essential or other protected SSC in the RXB area other than cables includes the CNTS, MPS, RBCM, RXB, UHS, and EDAS. These systems are discussed below: The only CNTS target SSC are the hydraulic actuator assembly skids and associated cables and tubing. There are two hydraulic actuator assembly skids for each NPM, one located on the 125 ft elevation above the pipe gallery (separated from high-energy lines) and one located in the pipe gallery. For CNTS lines with dual CIVs, one CIV is actuated by one skid while the other is actuated by the other skid. The hydraulic actuator assembly supplies a constant pressure via small diameter hydraulic tubing to keep the CIVs open and the DHRS actuation valves shut. Should the hydraulic lines be breached, the fluid would vent and the associated valves would be put into their safe position (e.g., CIVs shut) by the stored energy device attached to each valve or remain in the safe position if already there. Similarly, if electric power were lost or wires to the CIV skid actuators were dislodged, the hydraulic fluid would vent when control actuators de-energized, and the valves would move to their safe position. The hydraulic lines and electric cables are the portions of the skid most vulnerable to damage by pipe whip or jet impingement (blast wave loading is low and of such short duration that it has less damage potential) and, therefore, would cause valves to go to their safe position (and an associated reactor trip) should they be struck and breached during a HELB. Complete crimping of the hydraulic lines by HELB impact is not considered a credible failure mechanism because of their high internal pressure resisting crimping and, as this small tubing is fragile when compared to the relatively large forces due to HELB effects would cause them to breach rather than crimp if struck. Partial crimping could slow CIV closure. © Copyright 2022 by NuScale Power, LLC 46

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The presence of one of the hydraulic actuator assembly skids near postulated HELB locations is deemed an acceptable interaction: while there is potential for damage to the skid, that damage would not result in a failure for that skid to perform its safety function. The MPS and EDAS equipment in the RXB area is separated from external dynamic effects of HELBs in the RXB area by structures (walls and slabs). The RCBM components in the RXB area that are essential (pool area rupture disks and vents, steam gallery blow off panels, CVCS pipe chase blow off panels, CVCS heat exchanger room vents, and MHS equipment vents) function to vent pressure internal to the RXB that result from high-energy line breaks and therefore are designed for HELB effects. The RXB walls and slabs act as barriers against the dynamic external effects of HELBs in rooms that contain high-energy systems from essential and other protected SSC in other rooms (MPS, EDAS, NPM bay SSC, and others). These walls and slabs must be evaluated to ensure they act as barriers. Some structures may only be credited with mitigating external dynamic effects, while others may be credited with acting as a barrier to environmental effects as well. At this stage in the design, doors that are credited as HELB barriers are evaluated for dynamic subcompartment pressurization forces only, and these doors are not expected to be exposed to jet impingement or pipe whip. The only essential component of the UHS is the volume of pool water itself. Its safety function cannot be impaired by the external dynamic effects of HELBs. Cables belonging to systems may be routed through the RXB in order to connect to the MPS or EDAS; as discussed in Section 3.2.2.3.4, cable routing criteria ensures appropriate separation. High-energy systems with lines larger than NPS 1 in the RXB area include the MS, FW, CVCS, and MHS. As break exclusion criteria are not used anywhere in the RXB, these lines are subject to HELB postulation. However, because stress analysis is not yet complete for these systems, and these systems do not have design criteria requiring satisfaction of BTP 3-4 B.1(iii) stress criteria, circumferential breaks are assumed to occur anywhere in these systems. The effects of longitudinal breaks are bounded by the circumferential breaks. Leakage cracks are analyzed at locations which maximize long term environmental temperatures in rooms distant from the crack location. These discussions of high-energy lines inside the RXB are based on preliminary information in order to demonstrate the ability of the plant to safely shut down in the event of a pipe rupture. The conclusions reached in this report regarding the RXB area will be verified in a later stage in the design. 3.2.3.4 In the Plant Outside the RXB, separation of essential and other protected SSC are discussed in Section 3.2.2.3.2. Although the RXB outer wall is not separated from © Copyright 2022 by NuScale Power, LLC 47

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 high-energy lines, the RXB outer walls are designed to withstand aircraft impact, and therefore are considered adequate to withstand the external dynamic effects of pipe ruptures. Discussions of high-energy lines inside the Plant area are based on preliminary information in order to demonstrate the ability of the plant to safely shutdown in the event of a pipe rupture. The conclusions reached in this report regarding the RXB area will be verified in a later stage in the design. 3.2.3.5 Summary of Essential and Other Protected Structures, Systems, and Components Evaluated for External Dynamic Effects of Pipe Ruptures Table 3-5 summarizes the above discussions for use in the downstream evaluation of external dynamic effects. © Copyright 2022 by NuScale Power, LLC 48

Table 3-5 Summary of essential and other protected structures, systems, or components evaluated

    © Copyright 2022 by NuScale Power, LLC for external dynamic effects of pipe ruptures High Energy Lines Area      with Postulated                           Essential and Other Protected SSC in Proximity to HELBs HELBs RPV             CRDM support         PZR pressure sensors              containment level instruments CNV             ECCS venturis  6     CRDM pressure boundary            MS nozzles and piping EPAs            ECCS trip lines      CNV lower mixing tube             feed plenum access port covers RCS Injection ISAs            ECCS reset lines     FW nozzles and piping             RCS hot temperature instruments RCS Discharge CNV                          RVVs            RCS flow sensors ICI stringer assemblies               RCS cold temperature instruments PZR Spray RRVs            RPV support ledge RCS pressure instruments             steam plenum access port covers Degasification Line RCS lines       PZR level sensors CRDM magnetic jack assembly6 containment NR pressure instruments DHRS piping RPV level sensors CDRM seismic support plates              various cables PZR Heaters thermal relief valves supplemental boron dissolver NPM Bay none                   N/A1 MSS FWS RXB                          walls, slabs, doors4, hydraulic actuator assembly skids 5 CVCS MHS Plant   various2             none3 Notes:1. Not applicable as there are no postulated HELBs in this area.

2. A comprehensive list of high-energy systems outside the RXB will be generated in a later phase in the design

3. No essential systems are evaluated at this stage in the design outside the RXB, based on separation by distance.

4. Considered in the context of subcompartment pressurization only.

5. Only applies to skids on the 100 ft elevation. As noted in Section 3.2.3.3, external dynamic effects on the hydraulic actuator assembly skids are deemed an acceptable interaction.

6. Component is protected from HELB effects as described in Section 3.2.3.1.

Pipe Rupture Hazards Analysis TR-121507-NP 49 Revision 0

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.3 Rupture Characteristics Pipe rupture characteristics for circumferential and longitudinal breaks and leakage cracks are determined using the criteria of BTP 3-4 B.3. The application of this guidance is clarified below. All breaks are circumferential and assume at least a one-diameter lateral displacement of the pipe when calculating the effective flow area. Longitudinal breaks are not applicable in the CNV, because piping NPS 4 and larger meets break exclusion criteria. Also, longitudinal breaks need not be considered in the NPM bay, based on meeting break exclusion or stress criteria for not considering breaks. In the RXB, circumferential breaks are assumed to occur anywhere and bound the effects of longitudinal breaks. Leakage cracks in high-energy lines are modeled according to the criteria of BTP 3-4 B.3(iii) inside the NPM bay. Inside the CNV, the effects of leakage cracks are bounded by ECCS actuation, and outside the NPM bay leakage cracks are analyzed at locations that maximize long term environmental temperatures in rooms distant from the crack location. Leakage crack dimensions are not the focus of this report, but are calculated in the applicable mass and energy (M&E) calculations. Where postulated breaks might occur, the characteristics of those breaks (e.g., thermodynamic conditions) are identified as inputs necessary for the evaluation of external effects. In general, bounding conditions are used in analysis of breaks. For example, the CVCS piping has variation of fluid temperature and pressure with location in the RXB and with plant initial conditions. Rather than evaluating many specific conditions, initial intact system temperature and pressure values are selected to maximize the M&E release from the HELB in area of the plant and are, therefore, conservative for evaluating multiple break locations. The initial conditions are selected to bound load following from full power operation (102 percent thermal power) to 20 percent power and hot standby operation, for which the NuScale equivalent is referred to as hot shutdown. The combinations of pressures and temperatures listed in Table 3-3 need not be assumed to occur simultaneously to perform a bounding evaluation, but conditions should be chosen to maximize the particular dynamic effect being evaluated. Table 3-6 summarizes the postulated break locations and characteristics. The following applies for jet reaction forces: For MSS and feedwater HELBs, operating conditions for load following at 20 percent power produce higher calculated thrust loads that bound those at hot shutdown. For CVCS and reactor coolant system HELBs, full operating temperature and pressure produces the most severe blowdown. However, calculated thrust loads are not based solely on system pressure, because the jet thrust load is also dependent on the thrust coefficient, which is 2.0 for nonflashing blowdown, compared to 1.26 for steam. At hot shutdown conditions, the CNV is operated at a vacuum (<1 psia); therefore, to be nonflashing, the coolant temperature would need to be less than © Copyright 2022 by NuScale Power, LLC 50

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 about 100 degrees F. However, inside the CNV, during hot shutdown, the discharge and degas lines are both greater than 350 degrees F (RCS conditions), and the range of injection and PZR line temperatures is 365 degrees F-545 degrees F, eliminating the need to evaluate non-flashing jets. Table 3-6 Characteristics of Blowdown at Postulated Break Locations Blowdown Fluid Area System Break Location Type Flow Direction State1 From safe end Flashing RPV shell nozzle weld Circumferential From pipe Flashing injection From pipe Flashing CNV head nozzle weld Circumferential From safe end Flashing From safe end Flashing RPV shell nozzle weld Circumferential From pipe Flashing discharge From pipe Flashing CNV head nozzle weld Circumferential From safe end Flashing CNV RCS From safe end Steam RPV head nozzle weld Circumferential From pipe Steam2 degas From pipe Steam CNV head nozzle weld Circumferential From safe end Steam2 From safe end Flashing3 RPV head nozzle weld Circumferential From pipe Flashing PZR spray From pipe Flashing3 CNV head nozzle weld Circumferential From safe end Flashing Anywhere in a high CVCS temperature pipe Circumferential Both Flashing4 RXB MS Anywhere Circumferential Both Steam FW Anywhere Circumferential Both Liquid5 MHS Anywhere Circumferential Both Steam Notes:1. Based on conditions in Table 3-3. Flashing is from system having liquid with low enough subcooling to cause 2-phase blowdown.

2. Small reservoir of steam in pipe between break and closed CIV.

3. Blowdown turns to steam after liquid blows from line. No credit is taken for the significant flow restriction at the spray head.

4. Although the CVCS may contain high pressure low temperature lines, a higher temperature line is bounding for subcompartment pressurization analysis.

5. Corresponding to low power (20 percent) conditions.

3.4 Determination of Potential External Effects of Ruptures Including Dynamic and Environmental Effects External effects that are evaluated for postulated rupture locations are categorized as either dynamic or environmental effects. Environmental effects are considered for all types of postulated ruptures (breaks and leakage cracks in both high- and moderate-energy lines including the nonmechanistic breaks of the MS and FW lines) and include changes in pressure, temperature, and humidity, as well as spray wetting and © Copyright 2022 by NuScale Power, LLC 51

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 flooding. These environmental effects are not the primary focus of this report and are addressed in flooding evaluations and the EQ program. External dynamic effects are only evaluated for HELBs and include blast waves, jet loads (reaction and impingement), dynamic subcompartment pressurization, and pipe whip impact. The potential severity of external dynamic effects depends on parameters that are summarized below. Subsequent sections describe external dynamic effects in more detail. The parameters that determine the severity of HELB external dynamic effects are: Thermodynamic conditions of the system before the break occurs- higher energy fluid generally causes larger magnitude effects. This fluid energy in the blowdown is potentially consumed by several phenomena: failing the material 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. Size of the pipe that breaks - because the NPP is an SMR, piping serving a given PWR function is considerably smaller than a typical PWR. This smaller size reduces the relative severity and range of effects. For example, an MSS nominal pipe size 38 line in the AP1000 has an order of magnitude more energy per foot of fluid in the pipe than the NuScale NPS12 line, as discussed in Section 2.1. Location of the break (e.g., proximity to essential SSC, ambient conditions)

             -    If the break is sufficiently remote or separated from essential SSC, the effects are negligible.
             -    Inside the CNV, the initial ambient pressure is a vacuum, compared to other areas where the initial pressure is atmospheric pressure
             -    The flow may issue from a straight pipe section downstream of either a long pipe run or a nozzle connected to a reservoir (e.g., the RPV), which determines resistance and entrance losses.

Break configuration - in accordance with regulatory guidance, conservative assumptions are made (e.g., discharge coefficient of 1.0). No credit is taken for the reality that the end of an actual break likely has rough and bent edges that provide flow resistance. The break opens essentially 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 FW check valve). For estimating jet reaction and pipe whip, the blowdown is from an infinite reservoir at intact system normal operating conditions. This blowdown is assumed unless a check valve or normally closed isolation valve is available within a short distance to terminate flow. For subcompartment pressurization, blowdown may be terminated by valve closure after a single active failure or by depletion of the reservoir. © Copyright 2022 by NuScale Power, LLC 52

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 If the energy in the pipe is insufficient to deform the piping, no whip occurs. If the line is isolated, no sustained jet occurs. 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. 3.4.1 Blast Waves 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 realistic 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 NPP geometry. General points regarding the severity and evaluation of blast effects are summarized below: A blast wave is very weakly formed if the surrounding environment is at low pressure (less than 1 psia), as is the case inside the CNV. The severity of a blast depends on the amount of fluid that can escape within about one millisecond of onset of the break because the blast wave forms within that time.

                  -    The high-energy steam-filled lines are relatively small, which limits the severity of the blast pressure. As previously noted, the energy available to form the blast is less than 1/17th that of AP1000 (Section 2.1.)

Blast waves are not significant for subcooled discharge because liquid flashing occurs on timescales exceeding that of formation of a blast wave (Reference 6.1.22). A blast wave has well-defined and inter-related characteristics. For example, its peak pressure and speed decrease proportionately with distance from its origin (Figure F-1). The pressure load applied by a blast wave is of very short duration 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. In addition, the load has dissipated before the other loads (e.g., pipe whip, pressurization) can occur. 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. © Copyright 2022 by NuScale Power, LLC 53

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 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 Reference 6.1.27, 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 very 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. 3.4.2 Jet Reaction Jet reaction forces are due the pressure difference and momentum effects of expelled fluid at HELBs. Depending on the location of the break, these forces may be resisted by components (e.g., vessels) or pipe supports and connected structures, and may cause rapid acceleration of the broken pipe (i.e., whip) if the force is large enough to form plastic hinge in the pipe. The reaction force for a given break also represents the maximum jet impingement force that can be delivered to target SSC. The thrust coefficient is defined as the exit plane thrust force divided by the product of source pressure and exit plane flow area. Reference 6.1.10 also describes it as being determined by the force that must be exerted to hold in static equilibrium a plate positioned normal to the flow directly at the break point. 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 conditions in the CNV are low and their duration is about a millisecond (Figure F-7 shows 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, only the total, steady state jet thrust force FT as given by SRP Section 3.6.2 needs to be evaluated: FT = CT x Po x AE Equation 3-1 Where P o = initial intact system pressure (psia); in most formulations for F T . P o is usually given in terms of the difference between system and surround pressure, PA A E = the pipe break exit area (in,2) (subscript E refers to break exit) © Copyright 2022 by NuScale Power, LLC 54

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 C T = the thrust coefficient (unitless) Values for C T depend on fluid conditions but otherwise are largely independent of plant design. SRP 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. Reference 6.1.17 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 (e.g., RRV opening). Because F T is maximized when P A is a minimum, CNV pressure is set to be 0 psia initially. The equation 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 (e.g., 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. 3.4.3 Jet Impingement Jet impingement loads are imparted onto an object because of 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. 3.4.3.1 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 6.1.9, which reports types of insulation suffering damage for impingement pressures as low as 4 psig. NUREG/CR-6808 (Reference 6.1.26) Table 3-1 provides the impingement pressures found in testing to 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, such as ECCS valve bodies and the CNV steel wall. 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 © Copyright 2022 by NuScale Power, LLC 55

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 impingement effects on piping, and the relevant penetration distance is much shorter than 25 L/D. Westinghouse performed jet impingement testing on electrical cable in support of the AP1000 assessment of debris generation. The conclusion is that cables at 4 L/D from a jet simulating an AP1000 loss of coolant were not damaged. The results were given in terms of distance because of difficulty in accurately measuring impingement pressure. The NRC staff agreed with the conclusion. In Reference 6.1.11, 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 were applicable only to the type of cables actually tested, but an AP1000 RCS break jet would be 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. If cable is routed unprotected through a jet ZOI then it will be tested for survival under jet impingement. An overview of the NuScale vulnerability to jet impingement is based on plant operating conditions and small size of piping, thrust loads for NuScale line breaks are a small fraction of those normally encountered in large LWRs (e.g., a NuScale 12-inch MSS line has about 6 percent of the total thrust force of an AP1000 38-inch MSS line break1). Main steam HELB occurrence is limited to the RXB, because MSS breaks inside the CNV and in the NPM bay are eliminated by break exclusion. The NuScale operating conditions and pipe sizes result in significantly less M&E release as compared to traditional PWRs. damage to insulation on piping in the RXB is not a concern, so few essential SSC need to be evaluated. Allowable impingement pressure on SSC is considerably higher (at least 190 psia) than that in large PWRs where insulation stripping is relevant. with a lower system pressure and more jet resistant target, a MSS HELB penetration distance of 25 pipe diameters is a substantial overestimate. In the RXB, NPS 12 (inner diameter of 10.126 in.) The MS line breaks are postulated: 25 L/D corresponds to 21.1 feet. Only MS jets are evaluated in this phase in the design as their effects bound those of the FW, CVCS, and MHS lines inside the RXB. 1.AP1000 MSS is NPS 38 vs. NuScale NPS 12 and AP1000 MSS pressure is 836 vs, 700 psia, yielding a relative thrust of 16 to 1. © Copyright 2022 by NuScale Power, LLC 56

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 for HELBs inside the CNV, only NPS 2 piping (inner diameter of 1.687 in.) is susceptible. Presuming 25 L/D, the steam jet range is 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 2.3 L/D (3.9 in.). For unprotected cable, 4 L/D (6.75 in.) is used. similarly, although NuScale SSC are packed more closely together, the lower CVCS and MSS pressure, smaller pipe size, shorter 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 in AP1000. 3.4.3.2 Impingement Load / Pressure The maximum force applied to an impingement target is determined using Equation 3-1. The maximum break plane pressures for MS and CVCS/RCS (using maximum values regardless of specific line) 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. Table 3-7 Break exit plane parameters CVCS/RCS Break MSS Break (( Inner diameter (in.) Intact system pressure (psia) Break exit plane pressure (includes CT of 1.26)(psia) Break exit plane area (in.2) Maximum jet reaction / impingement force (lbf)

                                                                                                               }}2(a),(c)

The load on an object exposed to the jet depends on pressure of the jet upon the objects surface, on the intersection of the jet with the object, and the shape of the object. Jet pressure at the nearest target surface is determined, including the thrust coefficient CT (1.26 for steam and two phase jets). If it is less than 190 psia (or beyond 4 L/D for unprotected cable)2, the impingement pressure is low enough to be non-damaging, but a load may be determined in accordance with SSC design requirements for use in load combinations. Dynamic load factors are used where applicable to account for the dynamic nature of the loading. Y j = P l x A l x S F x D LF cos Equation 3-2 2.No additional factors need be applied because these criteria are based on testing. © Copyright 2022 by NuScale Power, LLC 57

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Where P l = Impingement pressure (psia) A l = Area of intersection of the jet and the projected target surface area perpendicular to jet axis (in2) Y j = Normal load applied to a target by the jet (lbf) S F = Shape factor for target SSC (unitless)(Reference 6.1.17) D LF = Dynamic load factor (unitless)

                        = Angle made by jet axis and line perpendicular to predominant target surface 3.4.3.3            Jet Zone of Influence Two types of breaks are considered per regulatory guidance: circumferential breaks and longitudinal breaks. The analysis assumes circumferential breaks are full separation because of the absence of nearby rigid restraints on both sides of postulated break locations. 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.

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. This under-expansion means that they 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 (via SRP Section 3.6.2) has expressed concern that this guidance is not generally applicable. 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. If the wider ZOI is considered in analysis, the drop off of pressure with distance is faster. © Copyright 2022 by NuScale Power, LLC 58

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The 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. The acceptability of jet impingement pressures is relatively insensitive to the analytical approach. Because piping inside the CNV is not insulated, the use of non-metallic material inside the CNV is minimal, most cable is protected by being routed out of range, and the RRV intake is directed downward and submerged during ECCS recirculation, jet impingement does not present a risk of generating debris capable of blocking ECCS recirculation. Although some piping outside the CNV is insulated, insulation stripping presents no hazard to safety-related functions. Methodologies are discussed for the different plant areas below. Because no breaks are postulated in the NPM bay under the bioshield (except for nonmechanistic breaks in MSS of FWS piping), no discussion is provided for that area. 3.4.3.4 Inside 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. The wider jet expands the ZOI but substantially reduces the pressure and the penetration length, because the M&E 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. The CFD blast modeling discussed in Appendix F shows that a steam jet initially develops a spreading half-angle more than (( }}2(a)(c). Appendix E provides a detailed discussion of the jet modeling applied in the CNV. A conservative ZOI of a forward facing hemisphere from the pipe ends original location is applied. In addition, the pipe whip trajectory is assessed for possible essential SSC that might be impinged by the jet as the forward facing hemisphere moves with the pipe end. For steam jets issuing from full separation circumferential breaks, a conservatively small initial spreading half-angle ((

                                        }}2(a),(c) is used to determine the jet pressure with distance.

This formulation yields jet pressures two to six times those for the predicted spreading half-angle of more than (( }}2(a)(c). It is judged sufficiently conservative that explicit allowance for radial pressure variations within the jet need not be included. For two-phase jets, the methodology of NUREG/CR-2913 (Reference 6.1.8) is applied to determine the jet pressure distribution versus distance from the break exit. This methodology is discussed further in Appendix E. © Copyright 2022 by NuScale Power, LLC 59

Pipe Rupture Hazards Analysis TR-121507-NP Revision 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 3.9 inches (2.3 L/D) is sufficient to provide acceptable protection of metal SSC. This distance is sufficiently short that few SSC are within the ZOI. For unprotected cable, 6.75 inches (4 L/D) is sufficient. In summary, SSC in the CNV more than 4 inches axially or radially (6.75 inches for unprotected cable) requires no evaluation for damage from jet impingement. 3.4.3.5 In the Reactor Building In the RXB, normal atmospheric pressure surrounds postulated break locations, and the available venting (Appendix D) limits the buildup of backpressure. Final piping designs in the RXB are not considered at this stage in the design; therefore, specifying a particular ZOI is not meaningful. To focus the impingement pressure, no expansion is considered. A bounding scenario for the MS line breaks in the piping gallery on the RXB structure concludes that because pipe rupture loads are not localized, they do not cause loss of function to any essential systems. 3.4.3.6 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 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, for example, instability of the discharge, irregular discharge geometry, phase changes that suppress pressure changes, misalignment of jet and impingement target surface preventing establishment of a feedback loop, and lack of an appropriately flat surface within a sufficiently close distance, as examples. If any one of these criteria are not met, a resonance is implausible. In a HELB, none of the criteria are satisfied, precluding formation of a resonance. 3.4.4 Pipe Whip Effects Pipe whip occurs when the jet thrust load at a HELB location is sufficient to form a plastic hinge in the portions of the piping system nearby the jet location, which would resist the jet load by bending. Therefore, the occurrence of pipe whip is dependent on the thrust force and the distance to the plastic hinge point on piping perpendicular to the direction of the jet. If a support is located on the pipe closer to the break location than the calculated plastic hinge point, whip is prevented, as long as the support is designed to withstand the load due to the jet thrust. Also, if the piping system does not extend beyond the required distance for a plastic hinge to form, whip does not occur. © Copyright 2022 by NuScale Power, LLC 60

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The calculation of thrust 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 degas pipe in the CNV is normally isolated). If the jet thrust is insufficient to form a plastic hinge in the pipe, then pipe whip at that break location is eliminated from further consideration. 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 90° 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. 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 3.2.2.3. 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. Similar to the design of pipe whip restraints discussed in Appendix I, loads for impacted SSC may be determined using static or dynamic analysis. Static load analyses must justify the applicability of the dynamic load factors. Note that a target cannot be subjected to both a pipe whip and a jet impingement load because the whip direction and jet vector are opposite. In other words, for a whipping pipe to strike a target, the jet driving the whip must be pointed away from that target. The jet from the other end of the break would be intercepted by a whipping pipe. Per Reference 6.1.3, an unrestrained whipping pipe is considered capable of causing HELBs in impacted pipes of smaller NPS, and of developing leakage cracks in equal or larger NPSs with thinner wall thickness. 3.4.4.1 Inside the Containment Vessel In the CNV, pipe whip loads are limited because only the NPS 2 locations that do not satisfy break exclusion need be considered. These locations are limited to terminal ends, so the opposite end of the break is a safe end, which does not whip. The congested arrangement and short piping lengths limit the whip distance and, thereby, limit the energy at impact. © Copyright 2022 by NuScale Power, LLC 61

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 3.4.4.2 Inside the NuScale Power Module Bay There are no postulated HELBs inside the NPM bay that could cause pipe whip. However, a HELB on the outboard side of the pool wall penetration in the RXB could potentially result in a pipe accelerating through the wall penetration and into the NPM bay. For this reason, restraints to prevent pipes from whipping through the wall have been added to high-energy lines that penetrate the wall. The pipe whip screening methodology described above is not used to eliminate postulated pipe whip for these configurations because of the presence of piping ball joints in the NPM bay, which greatly increase piping flexibility and are not compatible with the screening criteria. 3.4.4.3 Inside the Reactor Building In the RXB, HELBs that could potentially cause pipe whip are assumed to occur in any area that houses high-energy lines. Because the pipe is assumed to whip to any location in the room, a conservative assumption is made that pipe whip causes an additional HELB in one additional high-energy line in that room that is of smaller NPS. This assumption is utilized in the subcompartment pressurization analyses and in determining possible multimodule effects. This report does not consider final piping routings, cable routings, or mechanical SSC locations in the RXB at this time, and instead a bounding evaluations are performed in Appendix C, to ensure the RXB structure can act as barrier. The effects of pipe whip in the RXB will be confirmed at a later stage in the design. 3.4.5 Dynamic Subcompartment Pressurization In locations where HELB dynamic effects are not obviated by satisfying break exclusion, the pressurization transient resulting from the M&E release to the surrounding volume(s) is 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 (i.e., RXB) of the volume is sufficient to offset the blowdown. Inside the CNV, postulated HELB locations involve blowdown from an RCS-connected NPS 2 pipe. The M&E release for these HELBs are much less than 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. Loads resulting from the asymmetric pressurization of the space between the CNV and RPV during a HELB inside the CNV are calculated. In the NPM bay, no postulated HELB locations require evaluation because the piping satisfies break exclusion criteria. 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). Where required, safety vents or blow-off panels are added to reduce the peak pressures resulting from HELBs. © Copyright 2022 by NuScale Power, LLC 62

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Detailed methodologies for M&E release and subcompartment pressurization analyses are outside the scope of this report, but are summarized in Appendix D. © Copyright 2022 by NuScale Power, LLC 63

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 4.0 Summary This section summarizes the key information of Section 4.0 and the various appendices of this report with respect to each area of the plant (as defined in Section 1.2). 4.1 Inside the Containment Vessel Identification of Essential Structures, Systems, or Components and Other Protected Structures, Systems, or Components Essential systems are is identified in Section 3.1.1 and instrumentation necessary for PAM Type B and C variables are identified in Table 3-1. The location of the essential and other protected SSC other than cables are known so that detailed jet impingement and pipe whip evaluations can be performed as necessary. The RPV acts as a barrier protecting SSC inside of it from the effects of pipe ruptures, leaving only the SSC listed in Table 3-5 exposed to the effects of ruptures. Identification of High- and Moderate-Energy Systems in Proximity to Essential and Other Protected Structures, Systems, or Components The CNV is a barrier protecting SSC inside of it from ruptures in lines outside of it. High-and moderate-energy lines inside the CNV are listed in Table G-1 and additional information is provided in Table 3-3 for the high-energy lines. The SGS lines and the terminal end welds for the DHRS condensate lines are classified in as break exclusion zones, and design requirements require that stress criteria are met to eliminate intermediate breaks in the remaining lines. Rupture Characteristics As listed in Table 3-6, only terminal end circumferential breaks are postulated in NPS 2 RCS lines. These HELBs result in steam or two phase jets. Leakage cracks are not postulated because environmental effects are bounded by ECCS actuation. Determination of Potential External Effects of Ruptures Including Dynamic and Environmental Effects Environmental effects are bounded by ECCS actuation. The evaluations of the various external dynamic effects are summarized below: Blast waves - Because only NPS 2 lines are postulated to break, the M&E release feeding the blast formation is small. Only the degasification line has a potential for forming a blast, because the other CVCS lines contain subcooled liquid. The magnitude of the blast wave pressures is low, and based on the analysis summarized in Appendix F, a bounding force imposed on any component is limited to 6,000 lbf. However, the load is of extremely short duration (a few milliseconds) and is considered insignificant. © Copyright 2022 by NuScale Power, LLC 64

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Jet reaction - Maximum jet reaction loads for the RCS breaks (pipe side and vessel side) can be found in Table 3-7. These loads are to be addressed in NPM component stress evaluations as applicable. Jet impingement - Appendix E documents the jet impingement evaluations for the CNV area. Impingement forces and ZOIs are small, even when using conservative assumptions, because of the small size of the RCS piping. The only essential or other protected SSC to come into contact with jet ZOIs are the edges of the RPV instrument seal assembly flanges (but not the instruments penetrating the flange) and a small portion of some CRDMs. These components are required to address these loads per their design specifications; however, both components are robust perform their safety function despite the jet impingement loads, which are very small. Cables that are not yet routed are required to be located outside of these zones unless protected (e.g., conduit). There is no insulation inside containment that can be stripped to impact ECCS performance. Pipe whip impact - Two potential instances of pipe whip are identified due to terminal end breaks at the RPV nozzles of the RCS injection and discharge lines in the annulus between the RPV and CNV, where there is little room to whip. Both lines whip less than a foot before immediately contacting the CNV shell. Because of the concave cylindrical surface, the pipe comes to a rest after hitting the wall, and does not come into contact with essential or other protected SSC. Dynamic subcompartment pressurization - The CNV pressure temperature response to primary or secondary line breaks is outside the scope of this report. The additional effects of asymmetric pressurization are evaluated in HELB analyses. Dynamic pressurization effects of RCS lines are bounded by ECCS actuation. 4.2 Outside the Containment Vessel, in the NuScale Power Module Bay Identification of Essential Structures, Systems, or Components and Other Protected Structures, Systems, or Components Essential systems are is identified in Section 3.1.1 and instrumentation necessary for PAM Type B and C variables are identified in Table 3-2. The location of the essential and other protected SSC other than cables are known. Identification of High- and Moderate-Energy Systems in Proximity to Essential and Other Protected Structures, Systems, or Components The CNV is a barrier protecting SSC outside of it from ruptures inside. Additionally, the pool/bay walls act as barriers that mitigate the dynamic effects from ruptures inside the RXB. High- and moderate-energy lines inside the NPM bay are listed in Table G-1 and additional information is provided in Table 3-3 for the high-energy lines. Break exclusion criteria is extended to the outboard weld of CIVs (includes DHRS) in order to ensure terminal end breaks are eliminated in the NPM bay. Design criteria require that BTP 3-4 B.1.(iii) stress criteria be met, eliminating intermediate break locations and HELBs from consideration. © Copyright 2022 by NuScale Power, LLC 65

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Rupture Characteristics Only leakage cracks in high-energy lines and nonmechanistic breaks of the MS and FW lines are considered in the NPM bay. The analysis deviates from regulatory guidance regarding the size of the nonmechanistic breaks, due to the small piping size, as described in Appendix I. Leakage cracks in moderate-energy lines are not postulated as the effects are bounded by those of high-energy lines. Determination of Potential External Effects of Ruptures Including Dynamic and Environmental Effects Although not the focus of this report, pressure, temperature, and humidity effects due to leakage cracks in high-energy lines and nonmechanistic breaks in MS and FW lines are evaluated in the EQ program. Flooding is not applicable as the NPM bay drains into the RP. Because there are no HELBs in the bay, most external dynamic effects evaluations are not applicable except that restraints to prevent pipes from whipping through the pool wall as a result of a HELB in the RXB have been added to high-energy lines that penetrate the wall. 4.3 Outside the NPM bay, in the Reactor Building The following information is provided as a preliminary basis for the pipe rupture hazards analysis for the RXB area. These evaluations and conclusions are to be confirmed during a future phase in the design. Identification of Essential Structures, Systems, or Components and Other Protected Structures, Systems, or Components Essential SSC are identified in Section 3.1.1 however, the specific location of SSC is not considered for the purposes of this report and bounding analyses are performed. The RXB houses essential and other protected SSC belonging to the MPS and EDAS (includes PAM) as well as the CNTS actuator assembly skids. In a later phase in the design, the detailed location of SSC in the RXB are considered so that this hazards analysis can be verified. Identification of High- and Moderate-Energy Systems in Proximity to Essential and Other Protected Structures, Systems, or Components The RXB walls are barriers protecting SSC inside of it from ruptures in lines outside of it. A comprehensive list of high- and moderate-energy lines inside the RXB is not provided at this time; however, information regarding high-energy lines inside the RXB considered for bounding evaluations is provided in Table 3-3. © Copyright 2022 by NuScale Power, LLC 66

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Rupture Characteristics As listed in Table 3-6, circumferential breaks are postulated to occur anywhere in CVCS, MS, FW, and MHS lines routed in the RXB. These HELBs result in steam or two phase jets. Leakage cracks are postulated in high-energy MS and CVCS lines at locations that maximize long term environmental temperatures in rooms distant from the crack location. Leakage cracks are assumed to occur anywhere in seismically analyzed moderate-energy lines and circumferential MELBs are assumed to occur in non-seismically analyzed lines. Determination of Potential External Effects of Ruptures Including Dynamic and Environmental Effects Although not the focus of this report, pressure, temperature, and humidity effects and flooding due to circumferential breaks in high-energy lines and non-seismically analyzed lines, and leakage cracks in seismically analyzed moderate-energy lines are evaluated in The EQ program and flooding evaluations. The evaluations of the various external dynamic effects are summarized below: Blast waves - HELBs are postulated in the MS lines at three locations in the 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 mechanical component is less than (( }}2(a)(c), which is much less than the maximum jet impingement force. The shortness of the loading eliminates the need to consider it in load combinations. Jet reaction - Maximum jet reaction loads for the MSS breaks (pipe side and vessel side) can be found in Table 3-7. The effect of jet loads on the RXB are evaluated in flooding analyses. Jet impingement - Potential impact on a hydraulic actuator skid is acceptable because it causes the hydraulic lines to break venting off pressure or electrical lines to be damaged de-energizing the actuators, and allowing the CIVs and DHRS actuation valves to go to their safe position. For effects on concrete, MSS breaks are limiting and are assumed to occur anywhere, with no reduction in jet pressure with distance from the break. The effect of jet loads on the RXB are evaluated in flooding analyses. Because pipe rupture loads are localized, they have no effect on the overall structural integrity of the wall. In addition, the effect of erosion is negligible. Pipe whip impact - Pipe whip impact is evaluated for MS and FW lines in the pipe gallery. The MS lines, which are routed in a horizontal plane, are assumed to whip into the four foot thick steel composite concrete walls of the galleries, while the FW lines, which are routed in a vertical plane are assumed to whip into the two foot thick RC slabs (i.e., floors and ceilings). Both analysis use conservative assumptions such as assuming the full 90 degree rotation before impact. These analysis show that for the worst case HELBs in the RXB, the walls, ceilings, and floors are not perforated. © Copyright 2022 by NuScale Power, LLC 67

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Similar to jet impingement, these loads are localized and have no effect on the overall structural integrity of the wall. Dynamic subcompartment pressurization - A bounding analysis is performed in the RXB by postulating that HELBs can happen at any location in the lines and cause a secondary break in a conservatively selected smaller line. The RXB includes vents and blow-off panels to mitigate pressurization of subcompartments. The HELBs in the RXB produce dynamic pressures that are under the RXB design criteria for dynamic subcompartment pressurization, except for one small exceedance (<1 percent) which has been addressed. 4.4 Outside the RXB, in the remainder of the NPP Essential SSC are identified in Section 3.1.1; however, the specific location of SSC is not considered for the purposes of this report other than to demonstrate separation from HELB effects. Essential SSC in the Plant area is either housed in the CRB or located in a safety-related underground tunnel. A comprehensive list of high- and moderate-energy lines inside the Plant area is not provided at this time; however, based on site layout drawings, essential and other protected SSC are separated from the external dynamic effects of pipe ruptures by distance. In a later phase in the design, the detailed location of essential and other protected SSC and high- and moderate-energy lines in the Plant area will be considered so that this hazards analysis can be verified. © Copyright 2022 by NuScale Power, LLC 68

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 5.0 Results and Conclusions This report documents the methodology and results of evaluations performed to determine postulated rupture locations and the effects of those ruptures. The NRC guidance on relevant effects is identified, and how differences in the NuScale design affects application of that guidance is discussed. The 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 essential and other protected SSC. Examples of key features include: No operator action or electric power required for safe shutdown and cooldown for design basis accidents. Small volume metal containment operated at low pressure. No concern for dislodged piping insulation blocking core cooling. Greatly reduced energy of blast, pipe whip, and jet impingement effects because of smaller plant size and lower energy system conditions. Stainless steel primary and secondary piping within the containment and containment penetration area. This report concludes that external dynamic effects of HELBs do not adversely affect essential and other protected SSC required for safe shutdown or PAM. Inside the CNV and NPM bays, where the majority of essential SSC is located, the pipe rupture analysis is based on detailed design information. In the RXB, bounding evaluations are performed to demonstrate acceptability, as the detailed design is not complete. Outside the RXB, the only essential or other protected SSC are located in the CRB or in underground safety-related tunnels, where it can be shown to be separated from HELBs by distance. At a later stage in the design, detailed design information will be used to verify the above conclusions for the areas outside the NPM bays, and this report will be updated as necessary. © Copyright 2022 by NuScale Power, LLC 69

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 6.0 References 6.1 Referenced Documents 6.1.1 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) Referenced document. 6.1.2 U.S. NRC, Standard Review Plan, Plant Design for the Protection against Postulated Piping Failures in Fluid Systems Outside Containment, NUREG-0800, Chapter 3, Section 3.6.1, Rev. 3. 6.1.3 U.S. NRC, 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. 3. 6.1.4 U.S. NRC, Standard Review Plan, Leak-before-break Evaluation Procedures, NUREG-0800, Chapter 3, Section 3.6.3, Rev. 1. 6.1.5 U.S. NRC, Standard Review Plan, Protection against Postulated Piping Failures in Fluid Systems Outside Containment, NUREG-0800, Chapter 3, BTP 3-3, Rev. 3. 6.1.6 U.S. NRC, Standard Review Plan, Postulated Rupture Locations in Fluid System Piping inside and outside Containment, NUREG-0800, Chapter 3, BTP 3-4, NUREG-0800, Rev. 3. 6.1.7 U.S. NRC, Inspection of Pipe Rupture Hazards Analyses (Inside and Outside Containment) Design Acceptance Criteria (DAC)-Related ITAAC, Inspection Procedure 65001.21, November 11, 2011. 6.1.8 U.S. NRC, Two Phase Jet Loads, NUREG/CR-2913, January 1983. 6.1.9 U.S. NRC, Boiling Water Reactor ECCS Suction Strainer Performance Issue No. 7 - ZOI Adjustment for Air Jet Testing, BWROG Meeting, July 20, 2011, ML11203A432. 6.1.10 U.S. NRC, GSI-191 SER, Appendix I, ANSI/ANS Jet Model. 6.1.11 Corradini, M., Safety Evaluation for WCAP-17938-P, Revision 2, AP1000 In-Containment Cables and Non-Metallic Insulation Debris Integrated Assessment, ACRS, April 12, 2018. 6.1.12 U.S. NRC, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, Rev. 1. 6.1.13 ACRS U.S. EPR Subcommittee, February 21, 2012, ML120760106. © Copyright 2022 by NuScale Power, LLC 70

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 6.1.14 U.S. NRC, Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR Sump Pump Performance. 6.1.15 U.S. NRC, Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants, NUREG-1339, June 1990. 6.1.16 U.S. NRC, Design Specific Review Standard for NuScale SMR Design, Introduction - Transient and Accident Response, Section 15.0, ML15355A302. 6.1.17 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.) 6.1.18 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 Edition, Section III, Division 1, Rules for Construction of Nuclear Facility Components, New York, NY. 6.1.19 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 Edition, Section XI Division 1, Rules for In-service Inspection of Nuclear Power Plant Components, New York, NY. 6.1.20 American Society of Mechanical Engineers, Power Piping, B31.1, 2018 Edition, New York, NY. 6.1.21 Westinghouse Electric Corporation, AP1000 Design Certification Document, Rev. 19. 6.1.22 Liu, J., et al., Investigation on Energetics of Ex-vessel Vapor Explosion Based on Spontaneous Nucleation Fragmentation, Journal of Nuclear Science and Technology, 2002. 6.1.23 Ho, C.M. and N.S. Nosseir, Dynamics of an Impinging Jet, Part I, The Feedback Phenomenon, Journal of Fluid Mechanics, Vol. 105, pp 119-142. 6.1.24 South Texas Project Nuclear Operating Company, Re: RAI No. 364, U7-C-STP-NRC-100233, October 14, 2010, ML102910232. 6.1.25 Federal Emergency Management Agency, Building Design for Homeland Security, Course: IS-156; Lesson: 6 - Explosive Blast;https://emilms.fema.gov/is_0156/groups/1558.html. 6.1.26 Los Alamos National Laboratory, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, NUREG/CR-6808, 2/2003, ML030780733. 6.1.27 Department of Defense, Structures to Resist the Effects of Accidental Explosions, UFC 3-340-02, December 2008. © Copyright 2022 by NuScale Power, LLC 71

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 6.1.28 American Institute of Steel Construction, Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC N690-12, Chicago, Illinois. 6.1.29 U.S. NRC, Jet Impingement in High-Energy Piping Systems, NUREG/CR-7275, March 2021. 6.1.30 Alam, M. M. A., Matsuo, S., Mamun, M., Setoguchi, T and Kim, H. D., Effect of Moisture on Supersonic Impinging Jet Flows, Proceedings of the 4th BSME-ASME International Conference on Thermal Engineering, Dhaka, Bangladesh, 27-29 December 2008. 6.1.31 Alam, M. M. A., Matsuo, S., Setoguchi, T., Supersonic Moist Air Jet Impingements on Flat Surfaces, Journal of Thermal Sciences, Vol. 19, No. 1, pp. 51-59, 2010. 6.1.32 Marklund, Jan-Erik., Evaluation of Free Jet and Jet Impingement Tests with Hot Water and Steam, Studsvik Report, Studsvik/NR-85/54, 21 May 1985. 6.1.33 Siemens, Basic Dynamic Analysis Users Guide. 6.1.34 R. Gojon et al., Large-eddy simulation of supersonic planar jets impinging on a flat plate at an angle of 60 to 90 degrees, 21st AIAA/CEAS Aeroacoustics Conference, 22-26 Jun 2015. 6.1.35 ASME V&V 20-2009, Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer, 2009. 6.1.36 H.B. Hopkins, W. Konopka, J. Leng, Validation of scramjet exhaust simulation technique at Mach 6, NASA Contractor Report 3003, 1979. 6.1.37 ANSYS CFX Release 18.0 Documentation. 6.1.38 ANSYS Fluid Dynamics Verification Manual, Release 18.0, January 2017. 6.1.39 Pal, et.al, Verification and Validation of CFD Model to Predict Jet Loads and Blast Wave Pressures from High Pressure Superheated Steam Line Break, POWER2016-59675, Proceedings of ASME 2016 Power Conference, 2016. 6.1.40 Moore, M. J., Walters, P. T., Crane, R.I., Davidson, B.J., 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, 1973. 6.1.41 K. Kitade; T. Nagatogawa; H. Nishikawa; K. Kawanishi; C. Tsuruto, 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. 6.1.42 Moody, F.J., ASME-69-HT-31, Prediction of Blowdown Thrust and Jet Forces. © Copyright 2022 by NuScale Power, LLC 72

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 6.1.43 Guidelines for Vapor Cloud Explosion, Pressure Vessel Burst, BLEVE and Flash Fire Hazards, Center for Chemical Process Safety, 2nd Edition. 6.1.44 ANSYS CFX Modeling Guide, Release 18.0. 6.1.45 Micheli, I. and Zanaboni, P., 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. 6.1.46 Pipe Flanges and Flanged fittings, ASME B16.5 6.1.47 International Journal of Impact Engineering, Bruhl, J. C. et al. "Design of Composite SC Walls to Prevent Perforation from Missile Impact," 2015, pp 75-87. 6.1.48 Piersal, A. G. and Paez, T. L., Harris Shock and Vibration Handbook, 6th Edition, McGraw-Hill Book Co., New York, NY, 2010. 6.1.49 D.O.E. Standard, Accident Analysis for Aircraft Crash into Hazardous Facilities, DOE-STD-3014-2006 (1996 Reaffirmed) 6.1.50 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2015 Edition, Section III, Division 1, Subsection NE Class MC Components, New York, NY 6.1.51 Young, C. W., Penetration Equations, Sandia National Laboratories, SAND97-2426, Printed October 1997. 6.1.52 Lindeburg, M. R., Mechanical Engineering Reference Manual for the PE Exam, 12th Edition, Professional Publications, Inc., Belmont, CA, 2006. 6.1.53 Tennessee Valley Authority, NP-1320 Research Project 1324-2, Study of the State of Design for Pipe Whip, Final Report, January 1980. 6.1.54 NuScale Power, LLC, Pipe Rupture Hazards Analysis Technical Report, TR-0816-61384-P, Revision 2. © Copyright 2022 by NuScale Power, LLC 73

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Appendix A Break Exclusion - Compliance with Regulatory Criteria A.1 Application of BTP 3-4 B.1.(ii) Criteria The application of BTP 3-4 B.1.(ii) (Reference 6.1.6) for containment penetration areas, also known as break exclusion areas, is discussed in Section 3.2.2.4 and shown in Figure A-1 through Figure A-3. Criteria similar to BTP 3-4 B.1.(ii) is also applied to the ECCS main valve flanged connections to the RPV, which is discussed in Section A.2. High-energy piping within the break exclusion area is shown in Figure A-1 for lines inside containment, Figure A-2 for CNTS lines and Figure A-3 for DHRS lines outside containment. In these figures, break exclusion areas are shown in red. Where the portion of piping being identified is limited to a single weld, a red dot is shown rather than a line. BTP 3-4 B.1.(ii) contains seven design criteria that must be addressed in order to qualify as a break exclusion area. The criteria are briefly summarized below and the following sections discuss compliance to the criteria.

1. Conservative stress and fatigue limits are met for these areas.

a. BTP 3-4 B.1.(ii)1.(a), (b), and (c) for ASME Class 1 piping.

b. BTP 3-4 B.1.(ii).1.(d) and (e) for ASME Class 2 and 3, and B31.1 piping.

c. Although not explicitly considered in BTP 3-4 B.1.(ii), where piping is analyzed using the methods of NB-3200 rather than NB, NC, or ND-3600, the cumulative usage factor requirements of BTP 3-4 B.1.(ii)1.(b) are applied, while BTP 3-4 B.1.(ii).1.(a), (c), (d), and (e), which are based on stress equations in NB, NC, or ND-3600, are not applicable.

2. Welded attachments in these areas are to be avoided, except where detailed analysis or tests are performed to demonstrate compliance with 1 above.

3. The number of welds in these areas should be minimized, with no welds in portions of piping enclosed in guard pipes unless access provisions for in service volumetric examinations have been made.

4. The length of piping in these areas should be minimized.

5. Pipe anchors and restraints in these areas should not be welded directly to the surface of the piping, except where weld designs permit 100 percent volumetric examination and where detailed analysis is performed to demonstrate compliance with 1 above.

6. Guard pipes should be constructed in accordance with the criteria of ASME Code, Section III, Subsection NE, Class MC, and should:

a. have a minimum design pressure and temperature that bounds the maximum normal operating conditions of the enclosed pipe

b. be designed to conservative stress criteria (BTP 3-4 B.1(ii)(6)(b) describes specific criteria)

c. be subjected to a single pressure test at a pressure not less than the design pressure of the enclosed pipe

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d. not prevent access for in service inspection

7. An essentially 100 percent volumetric inservice examination of all pipe welds should be conducted during each inspection interval as defined in ASME Code, Section XI, IWA-2400. A distance equal to the thickness of the pipe wall plus 1.5 inches (measured from each edge of the weld) is the minimum required length of straight unobstructed piping needed for complete examination volume coverage (using a 45° ultrasonic transducer). However, where possible, a distance equal to two times the thickness of the pipe wall plus 2.0 inches is used for design. The IWA-2200 (Reference 6.1.19) definition of essentially 100 percent coverage is used as the criteria for achieving a 100 percent volumetric examination. In paragraph IWA-2200(c) it states: Essentially 100 percent coverage is achieved when the applicable examination coverage is greater than 90 percent; however, in no case shall the examination be terminated when greater than 90 percent coverage is achieved, if additional coverage of the required examination surface or volume is practical. These inspection requirements are applied to piping larger than NPS 1.

Additionally, BTP 3-4 states that piping between the containment and the isolation valves should meet the design criteria of ASME BPVC, Section III, Subarticle NE-1120. It should be noted that the Subarticle NE-1120 was revised significantly between the 2015 and 2017 Editions (applicable requirements of NE-1120 were relocated to NE-1100); therefore, the 2015 Edition of the ASME Code (Reference 6.1.50) is used for the purposes discussing BTP 3-4 (Reference 6.1.6) compliance. NE-1120 requires that only CNVs and their appurtenances be classified as Class MC with acceptable examples given in Figure NE-1120-1. The MC classification is not used for piping in the in the break exclusion area, and the CNV, although classified as MC, is constructed and stamped as a Class 1 vessel, as allowed by NCA-2134(c). Therefore, as MC is not used for the CNV (except for NE-7000) or piping, the requirement of NE-1120(a) is satisfied. Where valves, vessel safe-ends, and vessel nozzles are located between the containment wall and break exclusion areas shown in Figure A-1 through Figure A-3, the in-service inspection criteria of BTP 3-4 B.1.(ii)(7) are also applied to welds between those components. A.1.1 Steam Generator System Main Steam and Feedwater Piping The SGS main steam and FW piping classified as a break exclusion is shown in Figure A-1. Compliance with BTP 3-4 B.1.(ii) criteria for these lines is discussed below.

1. These lines are required to meet the applicable stress criteria of BTP 3-4 B.1.(ii).

2. There are no welded attachments to these lines except that the SGS feedwater lines include branches to the DHRS and thermal relief valves. Stress analysis of the SGS feedwater lines shall include these branch connections to ensure compliance with 1 above.

3. The number of welds is minimized with the use of piping bends rather than fittings.

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4. The length of piping is minimized to the extent practical. The pipe is routed to occupy the minimum amount of space inside to the CNV while still including enough flexibility to accommodate thermal loads.

5. Pipe anchors and restraints are not welded directly to the surface of this piping.

Vessel nozzles and safe-ends are not considered anchors in the context of this requirement.

6. Guard pipes are not used.

7. Welds greater than NPS 1 satisfy the minimum weld accessibility requirements for straight unobstructed piping.

A.1.2 Containment System Piping The CNTS-MS, CNTS-FW, CNTS-injection, CNTS-discharge, CNTS-PZR spray piping, and CNTS degasification piping classified as a break exclusion is shown in Figure A-2. Compliance with BTP 3-4 B.1.(ii) criteria for these lines is discussed below.

1. These lines are required to meet the applicable stress criteria of BTP 3-4 B.1.(ii).

2. These areas only include single welds with no physical piping.

3. These areas only include single welds with no physical piping.

4. These areas only include single welds with no physical piping.

5. These areas only include single welds with no physical piping.

6. Guard pipes are not used.

7. Welds greater than NPS 1 satisfy the minimum weld accessibility requirements for straight unobstructed piping.

A.1.3 Decay Heat Removal System Piping The DHRS piping classified as a break exclusion is shown in Figure A-1 and Figure A-3. Compliance with BTP 3-4 B.1.(ii) criteria for these lines is discussed below.

1. These lines are required to meet the applicable stress criteria of BTP 3-4 B.1.(ii).

2. There are no welded attachments to these lines except those needed for instrumentation. Stress analysis of the DHRS lines shall include these connections including applicable stress indices and stress intensification factors to ensure compliance with 1 above.

3. The number of welds is minimized with the use of piping bends rather than fittings.

4. The length of piping is minimized to the extent practical. The pipe is routed to reduce overall length while still including enough flexibility to accommodate thermal loads.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0

5. Pipe anchors and restraints are not welded directly to the surface of this piping.

Vessel nozzles and safe-ends are not considered anchors in the context of this requirement.

6. Guard pipes are not used.

7. Welds greater than NPS 1 satisfy the minimum weld accessibility requirements for straight unobstructed piping.

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure A-3 Containment penetration areas - decay heat removal system A.2 Connection of Reactor Vent Valves and Reactor Recirculation Valves to the Reactor Vessel In the design, each of two RVVs and two RRVs bolt directly to the reactor vessel. These four bolted-flange connections are classified containment penetration areas (i.e., break exclusion areas). Because this configuration does not include physical piping, a majority of the BTP 3-4 B.1.(ii) criteria do not apply. However, these BTP 3-4 B.1.(ii) criteria generically involve design stress and fatigue limits and in-service inspection (ISI) guidelines, which are addressed for these bolted connections below. Additionally, discussion is provided regarding threaded fastener design and leakage detection, in order to demonstrate that the probability of gross rupture is extremely low. The leakage detection systems along with in-service inspections provide assurance that potential failure mechanisms are detected before the onset of a catastrophic failure involving the fasteners of the bolted flange connections for the RRVs and RVVs, and therefore, that a break at this location need not be postulated. Design Stress and Fatigue Limits BTP 3-4 B.1(ii)(1) specifies more conservative stress and fatigue limits for ASME Class 1 piping in containment penetration areas than those required for piping by ASME Code, Section III, NB-3653. The bases for these more conservative limits include a desire to limit the stresses resulting from service loads (excluding those due to peak stresses) to within the material yield strength (i.e., elastic strains), and a desire for the cumulative usage factor calculation to account for the possibility of a faulty design, improperly controlled fabrication, installation errors, unexpected modes of operation including vibration, and other structural degradation mechanisms. © Copyright 2022 by NuScale Power, LLC A-6

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The RVV and RRV bolted connections are not classified as piping by their design specifications, and instead are classified as components designed to the rules of NB-3200. For the RVV and RRV bolting material, the design criteria given in NB-3230 for bolting provides greater margin against yielding due to service loads than do the rules of NB-3653 for typical piping system materials, even when considering the more restrictive limits of BTP 3-4 B.1(ii)(1). Therefore, the imposition of more conservative stress limits is not justified. Additional limits on cumulative usage factor are also not justified because the risk of a faulty design and fabrication and installation errors for a flanged connection is low compared to that of a piping system. The possible degradation mechanisms applicable to Class 1 piping systems do not apply to the ECCS valve bolting. These considerations are addressed further below. Faulty design is not a concern for the RVV and RRV flanges as the design features for these flanged connections that affect the stresses in the studs are primarily the number and size of the studs used, which are selected based on industry standards (ASME B16.5). The RVV and RRV flanged connections consist of Class 2500 NPS 5 and NPS 2 B16.5 flange configurations, respectively. ASME B16.5, "Pipe Flanges and Flanged Fittings," (Reference 6.1.46) has a history of reliability. In addition to conforming to an industry standard design, detailed analysis is required to validate the design per ASME BPVC Section III, NB-3230, including a fatigue evaluation. The fatigue evaluation for these studs utilizes the fatigue curve from ASME Section III, Division I, Mandatory Appendix I, Figure I-9.7. Figure I-9.7 is generated specifically for small diameter bolting made of SB-637 UNS N07718. Also, as required by NB-3230.3(c) for high strength bolting, a fatigue strength reduction factor of no less than 4.0 is applied to the studs. To address fabrication concerns, additional surface and ultrasonic testing examinations, beyond the ASME code requirements for these components, have been specified to properly control fabrication. Studs analyzed using NB-3232.3(b) have further requirements as stated in NB-3232.3(b)(2) and (3) that place controls on fabrication, by specifying both a minimum thread root radius and minimum radius between the head and shank, thus ensuring that the specified fatigue strength reduction factor used in the calculation of cumulative usage factor is sufficiently conservative. Unexpected modes of operation for piping systems in the nuclear industry generally involve thermal stratification, cycling, and striping. These situations do not apply to these valves. Unexpected vibration is another common concern, however, the RVVs and RRVs are within the scope of the Comprehensive Vibration Assessment Program (CVAP). The CVAP ensures that the structural components of the NPM exposed to fluid flow are precluded from the detrimental effects of flow induced vibration. Other degradation mechanisms that have contributed to past piping failures and not already discussed are addressed below. Included is an explanation as to why these mechanisms are less likely to occur in the RVV and RRV valves than in a typical piping system. © Copyright 2022 by NuScale Power, LLC A-7

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Corrosion - Not applicable as suitable materials have been selected and the bolting is not exposed to fluid during normal operation. Erosion/ Flow Assisted Corrosion - Not applicable as there is no flow through these valves during normal operation and the bolting is not exposed to fluid. Stress Corrosion Cracking (SCC) - Not applicable as suitable materials have been selected and the bolting is not exposed to fluid during normal operation. Water Hammer - Water hammer is not credible because there is no downstream piping and the valves discharge into a vacuum. Additionally, functional testing is performed for these valves including the dynamic effects of blowdown. Blowdown is classified as a service level B load in the ASME loading combinations for the valves, and therefore is included in the fatigue evaluations of the studs. In-Service Inspection BTP 3-4 B.1(ii)(1) states that a 100 percent volumetric in-service examination of all pipe welds should be conducted during each inspection interval as defined in ASME Code, Section XI, IWA-2400. This requirement is addressed for the RVV and RRV bolting by providing augmented ISI requirements for these studs that exceed the Code requirements. For in-service inspection, if the connection is disassembled during the interval, an ultrasonic testing inspection is performed on the studs. If the connection is not disassembled during the inspection interval, a volumetric inspection of the connection is performed in-place. Additionally, exceptions in the ASME code for flanged connections that allow only a sample of bolting to be inspected are not followed, and instead all flange studs for all RVVs and RRVs are inspected during each inspection interval. Threaded Fastener Design The applicable guidelines and recommendations in NUREG-1339 (Reference 6.1.15) have been adopted. Lubricants containing molybdenum sulfide are prohibited for pressure-retaining bolted joints including the RVV and RRV joints. Of the degradation mechanisms listed in NUREG-1339, only SCC could potentially affect RVV and RRV bolted joints. Alloy 718 is highly resistant to SCC in borated water. To further improve Alloy 718 SCC resistance, the solution treatment temperature range before precipitation hardening treatment is restricted to 1800 degrees F to 1850 degrees F. Additionally, the RRV bolting is submerged in borated water only during refueling, at a much lower temperature than RCS operating temperature, further reducing SCC susceptibility. The RVV bolting materials are not submerged in borated water as part of normal operating conditions. Based on these considerations, SCC is unlikely for Alloy 718 studs for RVVs and RRVs. Leakage Detection Leakage monitoring is provided by two means, the change in pressure within the CNV and collected condensate from the containment evacuation system. Even under a scenario where leakage occurs due to one or more postulated bolt breaks, containment leakage monitoring systems are sensitive to a leak rate as low as 0.05 gallons per © Copyright 2022 by NuScale Power, LLC A-8

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 minute. Containment is a relatively small closed volume and is maintained at a pressure of less than 1 psia during normal operation. High containment pressure is also a safety actuation signal that initiates a reactor trip. © Copyright 2022 by NuScale Power, LLC A-9

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Appendix B Dynamic Amplification and Potential for Resonance B.1 Background Based upon concerns raised by the ACRS in 2004, the NRC has identified in Reference 6.1.3 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 lead to backward-propagating transient shock and expansion waves that cause further unsteadiness in downstream shear layers. The concern is that 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 (Reference 6.1.17). 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 investigators 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 might result1. The basis for this concern is research into such amplification of loads that occur in the interaction of the jet issuing from vertical and short take-off and landing (V/STOL) aircraft and certain industrial gas jet applications. It causes vibration and fatigue damage to aircraft parts, jet deflectors, and parts cleaned with gas jets. This phenomenon has been studied extensively, with considerable work performed to mitigate its effects. Although SRP Section 3.6.2 (Reference 6.1.3) identifies the problem, it does not give guidance on acceptable means for resolution. Eventually, AREVA obtained NRC acceptance by a highly conservative approach in which they noted that research showed that dynamic amplification occurs only for pressure ratios (i.e., system pressure divided by ambient pressure) < 3.8. nozzle-to-target separation distance of < 5 L/D (i.e., separation distance divided by nozzle inner diameter). where the minimum jet frequency is close to resonant frequencies of impacted SSC. Despite these justifications, AREVA performed structural analysis using much more conservative assumptions of susceptibility at higher pressure ratios out to 10 L/D, setting an arbitrary match of jet and target SSC frequencies, resulting in target SSC loading 1.The system being described has three different areas which can oscillate: 1) the fluid of the jet, 2) the impinged target, and 3) the jet nozzle. Discussion of "resonance" refers to oscillation of the jet itself. This, in turn, can excite sympathetic vibrations in the target or the nozzle, either of which can cause damage as a result of amplified loads or fatigue. The NRC concern is that a resonant jet condition will cause a higher loading on an essential SSC. The target SSC does not actually have to resonate to be adversely affected. © Copyright 2022 by NuScale Power, LLC B-1

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 exceeding that predicted by experiment or analysis. During the ACRS review, the ACRS subcommittee chairman commented on the extreme conservatism of the approach (Reference 6.1.13). The other applicants all focused their original, unsuccessful arguments for not being susceptible to dynamic amplification upon a mismatch in jet frequency to target SSC natural frequency. This Appendix provides multiple justifications that resonances do not occur in HELB jets. B.2 Necessary but not Sufficient Conditions for Resonance High-energy line breaks involve steam, two phase, and liquid jets. Experiments performed with steam jets have noted oscillations on many occasions but the amplitude of the oscillations never exceeds the force at the exit plane. Occurrence of dynamic amplification leading to a resonance is not observed (Figure B-1, as an example). To determine the reason for the apparent lack of susceptibility to resonance, NuScale evaluated the criteria in Reference 6.1.23 that were identified as necessary for a resonance. Flat surface within 7.5 diameters of nozzle.

a. Although not explicitly stated in the paper, another necessary condition is that the surface be large enough to intercept most, if not all, of the expanding jet.

b. Also not stated is that the jet cannot be disturbed by intervening obstacles that partially intercept the jet, distorting the coherent structures required for resonance.

Subsonic jet with Mach number greater than 0.7. The most important observation is that phase difference at nozzle exit is an integer multiple of 2. This condition is necessary to set up a standing wave and avoid destructive interference between outgoing and return waves. The period of the wave is determined by the convection speed of coherent structures within the jet, the speed of the upstream propagating waves, which is needed to obtain a persistent phase match. The distance between the nozzle and the plate must be fixed in order to maintain the phase lock. Downstream-convected coherent structures and upstream-propagating pressure waves excite the thin shear layer near the nozzle lip. Large coherent structures play main role in the feedback mechanism. Dynamic loading as much as 50 percent higher loading than non-resonant jet. This magnitude is sufficient for its presence to be readily detectable in steam jet impingement experiments. Reference 6.1.30 identified two other necessary conditions for resonance: The jet must be axisymmetric, which allows the phase lock to be matched all the way around the break exit circumference. © Copyright 2022 by NuScale Power, LLC B-2

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Jet is normal to plate. If the jet/plate intersection is not normal, then waves returning to the break exit are out of phase and not aligned to direct return waves back to the break exit. Figure B-1 Normalized pressure of steam jet vs. distance to impingement plate (Reference 6.1.32) B.3 Susceptibility to Dynamic Amplification and Resonance NuScale has evaluated Reference 6.1.23 and Reference 6.1.30 and identified key factors that preclude occurrence of a resonant condition for a HELB in the NPP. These factors are emphasized in italics in the above list and discussed below. This approach is necessary because NRC guidance in Reference 6.1.3 is to refer to Reference 6.1.23. B.3.1 Flat Surface Within 7.5 Diameters The design is compact and has small diameter, high-energy lines. The small diameter limits the distance within which Reference 6.1.23 would claim a resonance could © Copyright 2022 by NuScale Power, LLC B-3

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 occur (i.e., 7.5 pipe diameters (i.e., L/D), although the NRC rounds up to 10 L/D in Reference 6.1.3) as shown in Table B-1. For jets from the CVCS/RCS lines, no flat surfaces exist inside the CNV. For lines in the RXB, walls, ceilings, and floors are the only flat surfaces within range. Note that these surfaces are not completely flat and that there are other interferences (e.g., neighboring pipes) that interrupt portions of the jet. Therefore, resonance cannot occur inside the CNV and is unlikely to occur in the RXB because of the absence of a suitable impingement surface. B.3.2 Mach number > 0.7 Table B-1 Range of potential resonance region in the Reactor Building Pipe Inner Diameter Distance for 7.5 L/D Flat Surfaces in Range (in) (in) (( MSS Walls, ceiling, floor FWS Walls, ceiling, floor CVCS Walls, ceiling, floor

                                                                                     }}2(a),(c)

Notes:

1. Assumed for the purposes of this table to be NPS 12 Sch. 80.

2. Assumed for the purposes of this table to be NPS 6 Sch. 80.

3. Assumed for the purposes of this table to be NPS 3 Sch. 160.

Mach number is not assessed. However, the jet is initially under expanded coming out of the break and accelerates. Therefore, it is assumed this criterion is met. B.3.3 Phase Difference Integer Multiple of 2 Another way to describe this factor is that an even number of wavelengths must fit within the nozzle lip to impingement surface distance. This factor is necessary for positive reinforcement by the upstream return waves at the nozzle lip. If the return waves are out of phase, they partially cancel the downstream propagating waves. The frequency staging with distance phenomenon described in Reference 6.1.23 is indicative of a standing wave. Assuming a speed of sound of 1130 ft/sec (13,560 in/sec)2 for normal atmospheric pressure and applying the criterion that the Mach number is at least 0.7, then the wave speed is at least 791 ft/sec (9492 in/sec). Table B-2 gives the wavelength of periodic pressure oscillations, if they occurr, for various frequencies and shows whether the wavelength can fit in the available separation of 7.5 L/D. At frequencies of less than 100 Hz for the MSS and 500 Hz for the CVCS, a whole wavelength cannot fit. Large plant components (vessels, walls, and others) generally have lower natural frequencies and because of their robust design for other loads, already include significant margin for jet loads. Smaller structures that might have higher natural frequencies are not flat and do not have sufficient surface area to intercept the 2.A higher speed of sound results in larger wavelengths for a given frequency. © Copyright 2022 by NuScale Power, LLC B-4

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 whole jet. Therefore, it is not possible to meet the 2 criterion within 7.5 L/D at a frequency capable of exciting an essential SSC. Table B-2 Wavelengths of downstream propagating waves Frequency Wavelength No. of wavelengths No. of wavelengths (Hz) (in) for CVCS in 7.5 L/D for MSS in 7.5 L/D (( 100 150 300 500 1000 3000 10,000

                                                                                                 }}2(a),(c)

B.3.4 Speed of Upstream Propagating Waves If pressure is varying in the jet, the density of the compressible fluid varies, which in turn affects the speed of wave propagation. To maintain a 2 phase difference, the speed of the downstream and upstream propagating waves must remain constant or exactly matched. In a phase-locked, non-condensable gas jet, the standing wave ensures that an upstream wave encounters the same density gradient as the downstream wave, even when it is oscillating. A steam or 2-phase HELB jet differs. As a saturated steam jet emanates from the break, it undergoes partial phase change from the high pressure and temperature at saturated conditions to atmospheric conditions at ambient pressure. The phase change occurs at differing rates depending on the localized conditions within the jet. For example, a drop in pressure in a local region of the jet promotes flashing leading to expansion of that region of the jet, causing pressure to rise and quench further flashing. This cycle is a form of negative feedback loop that suppresses the pressure oscillations needed to establish the resonant condition. Furthermore, Alam et al. (Reference 6.1.30 and Reference 6.1.31) produced numerical calculations and confirmed through experimentation that non-equilibrium condensation may occur in the jet between the nozzle and target, and the surrounding gas (i.e., steam or vapor) is heated by the release of latent heat of condensation. The overall effect is to dramatically alter the jet flow behavior, such that energy of turbulent fluctuations is reduced because of the relaxation process of evaporation and condensation. That is, the occurrence of non-equilibrium condensation reduces or dampens the energy variation of impinging jets, and the amplitudes of surface pressure oscillations become smaller as compared with dry air. Importantly, Alam et al. note: © Copyright 2022 by NuScale Power, LLC B-5

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 increment in the initial degree of supersaturation S0 results in the reduction of the amplitude of surface pressure oscillations due to the occurrence of non-equilibrium condensation in moist air jets. This phenomenon, of itself, is sufficient to explain why resonant jets are not encountered in experiments representing HELBs. Additionally, as the downstream and upstream propagating waves move through different regions of the jet, the localized quality level of the steam-liquid mixture varies sufficiently to disturb the phase lock by altering the speed of the waves moving through the jet. B.3.5 Period Set by Wave Speed and Distance Between Nozzle and Plate Jet resonance occurs in situations where the jet nozzle exit is situated optimally. In experiments, the nozzle is secured to ensure repeatable results. In V/STOL aircraft and industrial gas jets, the jet is positioned to get the desired thrust against the impingement surface. Again, HELBs are different. They issue from a randomly displaced pipe that is not affixed in its current location. The force of the blowdown thrust is not absolutely stable, resulting in flutter of both angle and of break exit to target separation. These random variations introduce distortion of the wave period and vary the jet to surface alignment, preventing a phase lock from being maintained. B.3.6 Thin Shear Layer near the Nozzle Lip In situations where a resonant gas jet has been found to occur (e.g., V/STOL aircraft, industrial cleaning applications, experiments), the jet issues from a smooth, symmetrical nozzle. For the thin shear layer to be uniformly excited in the feedback loop, its surface must be smooth and regular. A HELB jet issues from a pipe end that has ripped loose from its other side and possibly impacted hard against something because of pipe whip, leaving ragged edges that may bend inward or outward, a non-circular cross-section, and a circumference that is likely not equidistant to the impingement target. These distortions interfere with the shape and regularity of the jet and the thin shear layer necessary for the feedback loop. B.3.7 Large Coherent Structures Play Main Role in Feedback Factors discussed in other parts of this section serve to disrupt the stability necessary to maintain the coherent structures. © Copyright 2022 by NuScale Power, LLC B-6

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 B.3.8 Dynamic Loading as much as 50 Percent Higher than Non-Resonant Jet Loading higher by 50 percent is a modest effect for a resonance, but even that amount has not been seen in experiments simulating HELB conditions.3 Section B.3.3 already showed that the frequency for resonance must be above 100 Hz for an MSS jet and 500 Hz for a CVCS jet to satisfy the 2 criterion. Potential targets of postulated CVCS breaks in the CNV are large metal components (cables do not have a flat surface or sufficient surface area to reflect back to the break) with natural frequencies much less than 500 Hz. In the RXB, the targets are RXB walls, floors, and ceilings, which have natural frequencies approximately in the range of 4 to 16 Hz. Dynamic amplification factor (DAF) is the amount by which an oscillation induced in a structure is affected by the proximity of the excitation frequency to the natural frequency as adjusted by the structural damping of the structure. The seismic damping factor for some nuclear plant SSC is given in Regulatory Guide 1.61 (Reference 6.1.12); 1 percent is a lower bound. The equation for dynamic amplification factor from Reference 6.1.33 is: 2 2 -1/2 2 2 DAF = 1 - ------ + ---------- Equation B-1 2 n n Where:

                      = the excitation frequency (Hz) n = the target SSC natural frequency (Hz)
                      = the damping factor (unitless)

Input of the CVCS and MSS jet values and a damping factor of 1 percent into Equation B-1 yields low DAF values for both cases, indicating that even if a resonant jet existed it would have an insignificant effect on the loading of the target SSC. This is consistent with experimental results that show an absence of resonance even where they confirm the thrust coefficient values are appropriate and that random oscillation of the jet can have brief peaks that are far above average. B.3.9 Jet Must be Axisymmetric To meet the 2 criterion, the jet must be circumferentially uniform. Otherwise, different parts of the nozzle lip would receive backward-propagating waves at 3.This effect is unrelated to the thrust coefficient of 1.26 for steam and 2-phase jets and 2.0 for liquid jets. © Copyright 2022 by NuScale Power, LLC B-7

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 different phase angles. As discussed in Section B.3.6, resonant gas jets occur in systems engineered to produce a uniform, consistent jet. The HELBs differ by having a distorted exit geometry and fluctuating discharge conditions. Further, none of the studies of resonant gas jets or non-resonant HELB jets impose an intervening object in the path of a jet. For example, if a MSS pipe HELB occurs in the RXB pipe gallery, it is likely that other piping, hangers, or structure is between the break exit and the nearest wall, floor, or ceiling. The presence of such interference prevents formation of an axisymmetric jet. B.3.10 Jet Axis Normal to Impingement Surface Although Reference 6.1.23 assumes the jet axis is normal to the impingement surface, it does not explicitly discuss it. The need for this alignment is consistent with discussion earlier about maintaining the 2 criterion. There are a number of experimental and analytical studies (Reference 6.1.34, for example) demonstrating the off-normal impingement angles preclude resonance due to reflecting the upward-propagating waves away from the nozzle exit so they cannot establish a feedback loop. In a nuclear plant HELB, the whip of the broken pipe somewhat randomly comes to rest against some obstacle. There is a negligible chance that it happens to be aimed squarely at a large, flat surface. In the RXB, where flat surfaces predominate, piping is usually installed with its axis normal to a wall. After the HELB initiates, the pipe axis is unlikely to end up normal to a wall because the only way for it to do so is to not move away from its original location (in which case the other end of the pipe is in the way) or to redirect exactly 90 degrees. While not impossible, normal alignment to a large, flat impingement surface is virtually precluded. B.3.11 Addition of Moisture Although not listed in Section B.2, introduction of moisture from outside the jet is another means to reduce or eliminate dynamic amplification. As a jet emanates from a HELB, rapid expansion causes water droplets to be carried along or to form. If the jet envelopes any obstacles, splashing occurs. Additionally, the other end of the broken pipe is discharging a counter-current flow if the HELB is double-ended. This interjection of additional moisture accentuates the nonequilbrium condensation effect and destroys any axisymmetry the jet might otherwise have had. B.4 Summary Theoretical and computational models of jet impingement, including highly controlled laboratory experiments, such as those of Ho and Nosseir (Reference 6.1.23), are based on idealized conditions that rarely occur in practice, unless specifically engineered to do so. For example, experiments are developed with the goal of minimizing the number of variables, thereby focusing on parameters easily controlled and verified by the © Copyright 2022 by NuScale Power, LLC B-8

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 experimenter. Instances of resonances outside the laboratory occur where a design is optimized close to the onset of resonance. As described in this Appendix, a HELB scenario differs from resonant-inducing conditions in multiple ways, most of which are sufficient by themselves to preclude a resonance. These include: condensable fluid in the jet distorted exit geometry absence of a large, flat impingement surface sufficiently close and perpendicular to the jet axis instability of jet exit separation distance and angle presence of obstacles or intersecting flow that disrupt jet symmetry NuScale concludes that dynamic amplification of HELB jet pressure leading to resonance with the possibility of increased loading on impinged SSC is precluded for the design. © Copyright 2022 by NuScale Power, LLC B-9

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Appendix C Pipe Whip This appendix provides the additional details of the methodology for pipe whip evaluations and also documents the evaluations and their results. The methodology includes the determination of whether a pipe has sufficient energy to whip, whether a whipping pipe can actually contact an essential or other protected SSC, 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. This methodology incorporates the guidance and considerations described in Section 3.4.4. Pipe break locations that are evaluated are listed in Table 3-6. As discussed in Section 3.2.2.3.1, SSC located inside the RPV, which acts as a barrier, are considered separated from break effects inside the CNV. The remainder of the essential SSC located in the CNV area and also the PAM instrumentation located inside the CNV listed in Table 3-5 are evaluated in Section C.1.2 for effects of pipe whip. There are no pipe break locations inside the NPM bay area (Table 3-6); therefore, no pipe evaluations are performed for that area. Section C.2 discusses pipe whip in the RXB using bounding assumptions to verify acceptability. However, detailed evaluations based on final design information are need to confirm this evaluation at a later stage in the design. C.1 Inside the Containment Vessel C.1.1 Pipe Whip Screening The thrust force is determined using the methods described in Section 3.4.2. As described in Section 3.4.4, the whip evaluation may consider the absence of energy reservoirs. Table 3-6 indicates that there is little steam in the line for the break in the RPV degas line at the RPV head, which is because the CIV in this line is normally closed. Therefore, because there is little steam in the pipe, and there is no reservoir to supply more energy, pipe whip is not considered for the RPV degasification line break at the RPV nozzle. The distance to the plastic hinge point is based on the methodology documented in Reference 6.1.45. Figure C-1 shows a piping system of two mass segments, m1 and m2, 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. © Copyright 2022 by NuScale Power, LLC C-1

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 The equation to calculate the minimum distance for a plastic hinge, Lh, from the break, from Reference 6.1.45 is: 3M P 8LF L h = ----------- 1 + 1 + ------------b- Equation C-1 2F b 3M P Where, L h = distance from hinge point to pipe as shown in Figure C-1 M P = S y Z p , plastic bending moment F b = thrust force as calculated in Section 3.4.2, Equation 3-1 (FT) L = length from break location to first elbow S y = yield strength of pipe material 4 3 3 Z p = --- r - r , plastic bending section modulus 3 o i Equation C-1 above takes into account inertia effects. Note that if L=0, Equation C-1 reduces to: 3M L h = ----------P- Equation C-2 Fb Figure 1 from Reference 6.1.45 describes the geometry of the plastic hinge. From Equation C-2, the minimum distance, L h , for the thrust force to overcome the pipe resistance to bending can be determined. Alternatively, if there is a length of pipe at the end of the whipping segment, then Equation C-1 applies. Resulting values for NuScale piping at maximum operating temperatures are shown in Table C-1 below. The results in Table C-1 regarding whether the pipe whips or not are based on the first run of pipe after the first bend from the break location. Note that the distance available for a hinge to form in the piping is the distance perpendicular to the jet force, which induces a moment in the line. Therefore, the actual length of piping may be longer if the bend is less than 90° (Figure C-2). As described in Section 3.4.4, the safe-ends and nozzles do not whip because they are short, stiff, straight, and restrained by the component. © Copyright 2022 by NuScale Power, LLC C-2

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Table C-1 Maximum hinge length Lh to avoid pipe whip for reactor coolant system lines ((

                                                                                                  }}2(a)(c)

Figure C-1 Hinge location for elbow less than 90° (i.e., non-perpendicular pipe run) The results in Table C-1, show that a plastic hinge forms in the first run of pipe after first elbow only for the RCS Injection and Discharge line break locations at the RPV nozzles. This break occurs in the annulus between the RPV and CNV and has little room to whip before impacting the CNV wall (Figure C-3 below). As discussed in Section E.6, the relocated jet ZOI during pipe whip is also shown in Figure C-2 as a green hemisphere. Note that the hinge location shown in the figure is located at the first pipe support; however, the Table C-1 results indicate the hinge happens much nearer to the break location. As discussed in Section 3.2.2.3.1, the CNV wall is a robust component, with a wall thickness greater than the outside diameter of the whipping pipe. The impact is required to be considered in the design of the CNV per its design specification; nonetheless, the impact does not affect the safety function of the CNV. © Copyright 2022 by NuScale Power, LLC C-3

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure C-2 Resulting pipe whip due to breaks in the reactor pressure vessel discharge and injection lines at the reactor pressure vessel nozzles ((

                                                                                              }}2(a),(c)

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 C.1.2 Pipe Whip Evaluation for Inside Containment Vessel Area Table C-1 presents the results of the pipe whip screening evaluation. Two potential instances of pipe whip are identified due to terminal end breaks at the RPV nozzles of the RCS injection and discharge lines. These breaks occur in the annulus between the RPV and CNV, where there is little room to whip. Nearby essential SSC include the following: RCS: RPV Shell SGS: Piping DHRS: Piping CNTS: CNV Shell MPS/PAM: RCS Flow, RCS TCold and THot, CNV Level, RRV position As shown in Figure C-2, both lines whip less than a foot before immediately contacting the CNV shell. Due to the concave cylindrical surface, the pipe comes to a rest after hitting the wall, and does not come into contact with the SSC listed above. Note that detailed cable routing for MPS/PAM instruments is not yet complete, but design requirements ensure cables are not routed in the path of the whipping pipes. The safety function of the CNV is unaffected by the impact. It should be noted that this whip analysis is not dependent on the pipe support design inside the CNV because the distance available to hinge (as listed in Table C-1) in each case where whip occurred is not limited by a support, but instead limited by the piping layout. Also, in the two cases where whip did occur, the plastic hinge occurred before the support. However, this evaluation is dependent on the pipe routing inside containment and even small changes in that routing require reevaluation of pipe whip screening and effects. C.2 Reactor Building Area C.2.1 Pipe Whip Screening The high-energy lines in the RXB area considered for pipe whip effects are listed in Table 3-3. This report does not generally consider detailed pipe routings or HELB locations in the RXB area except for the MS line (main run) in the RXB piping gallery, which is routed in a horizontal plane; and therefore, is assumed to only whip laterally for evaluations of the RXB structure (i.e., it whips into the structural composite walls rather than the RC slabs). Additionally, pipe whip is assumed to also result in one additional (limiting) break in a line that is of smaller NPS than the whipping pipe, for consideration in dynamic subcompartment pressurization evaluations, and in determining multi-module effects. However, for pipe whip impact evaluations on the RXB structure only the MS and FW lines in the piping gallery are evaluated at this stage in the design, which is expected to be bounding for the other smaller high-energy lines elsewhere in the RXB. The MS © Copyright 2022 by NuScale Power, LLC C-5

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 line pipe whip is bounding for the 4 ft thick SC walls, and FW line pipe whip is bounding for the 2 ft thick slabs. In a later phase in the design, the detailed location of essential and other protected SSC and high- and moderate-energy lines in the Plant area is considered so that this hazards analysis can be verified. C.2.1.1 Screening for Onset of Pipe Whip (Main Steam and Feedwater Only) 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., FW, MSS) has a minimum Lh that is necessary to initiate pipe whip. The distance to the plastic hinge point is based on the methodology documented in Reference 6.1.45. As previously stated in Section C.1.1, Figure C-1 shows a piping system of two mass segments, M1 and M2 (these mass segments are defined hereafter in lower case m mass terms, but remain capitalized in referenced Figure C-1), 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, d, as depicted by position B. From Equation C-2, 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 Equation C-1 applies. Resulting Lh values for postulated plastic hinge locations on the NPS 12 MSS pipe and NPS 6 FWS pipe are for schedule 80, SA-335 P11 piping material with operating temperatures shown in Table C-2. The stainless steel SA-312 TP304L portion of the MSS pipe in the RXB that is also subject to forming a postulated plastic hinge is schedule 120, but its less severe pipe whip is not evaluated because it has longer maximum hinge lengths to prevent pipe whip shown in Table C-2 below. Table C-2 Maximum hinge length Lh to avoid pipe whip ((

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Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 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 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.2.1.2 Simplified Solution of Pipe Whip Impact Velocity The thrust load applied at the break location immediately after rupture, depicted by position A, causes a plastic hinge to form allowing the pipe segment to rotate at an angular velocity, (rad/sec) and traverse a path, , as depicted by position B (Figure C-1). The following methodology leads to conservative impact velocity. After break-opening, the steady-state jet thrust force, Fb, is found (shown as FT in Section 3.4.2). As previously discussed in Section C.1.1, the plastic hinge is formed at a distance, Lh, from the break, resulting in a plastic moment defined by: M L h = ------p- Equation C-3 Fb Where L h = Distance from hinge point to pipe as shown in Figure C-1 M p = Bending moment The above form is commonly utilized in static analyses yet neglects the influence of pipe length from the break to the first elbow, as well as restraint effects. It allows for a conservative estimation of the minimum unrestrained length of pipe that causes the formation of a plastic hinge but leads to unrealistically short hinge lengths. A more realistic formulation for the hinge length often used in restraint design, which is based on energy balance (Reference 6.1.45) assumes the possibility of an additional length of piping, L, located perpendicular to rotational motion (Figure C-1, portion of piping labeled m 2 ): 3M P 8LF L h = ----------- 1 + 1 + ------------b- [Referencing Equation C-1] 2F b 3M P © Copyright 2022 by NuScale Power, LLC C-7

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Nonetheless, if there is no additional length of piping, (i.e., setting L = 0) the hinge length equation reduces to: 3M L h = ----------P- [Referencing Equation C-2] Fb Here, the plastic bending moment, assuming small deformations, may be taken as: MP = Sy Zp Equation C-4 Where S y = Yield strength of pipe Z p = Plastic bending section modulus, which is given by 4 3 3 Z p = --- r - r Equation C-5 3 o i Where r o = pipe outer radius r i = pipe inner radius While equations for Z p (plastic section modulus) above and Z e (elastic section modulus) are described in Reference 6.1.45, Z e (Equation C-7) defined below, is specifically taken from the supporting Reference 6.1.53 for accuracy. If large deformation is assumed, which include strain hardening behavior of the material, a better approximation to the plastic bending moment capacity is given as: M P = S y Z p + ( S u - S y )Z e Equation C-6 Where S u = Ultimate strength of pipe Z e = Elastic bending section modulus, which is given by © Copyright 2022 by NuScale Power, LLC C-8

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 4 4 Z e = -------- r - r Equation C-7 4r o o i In solving the pipe whip problem, work-energy principles are applied to the model of Figure C-1, while noting that position A depicts the piping system immediately after rupture and just before 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. Additionally, the positive depth values that can be determined for the estimates of how far the respective pipes may penetrate into their defined targets in Section C.2.2 and Section C.2.3 implies that there is no potential for rebound upon impact. Therefore, adding a factor for rebound force is unnecessary for this methodology of Reference 6.1.45 given to find the maximum impact velocity of the whipping pipe at position B. The effective mass of the system and associated kinetic energy are generally derived from dynamic principles (Reference 6.1.45 and Reference 6.1.48) as: m m eff = ----- + m 2 Equation C-8 3 ( KE ) A + W A - B = ( KE ) B Equation C-9 Where the work is defined as the thrust force over the distance traversed minus the plastic moment resistance: WA - B = Fb - Mp Equation C-10 Where

                       = the angle through which the pipe whips (radians)

The kinetic energy of the whipping pipe about the hinge point is based on rotational kinematics (Figure C-4): 1 2 ( KE ) B = --- I h Equation C-11 2 Where

                       = Angular velocity I h = Mass moment of inertia about hinge-point

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 Figure C-3 Mass moment of inertia about centroidal axis Figure C-4 Mass moment of inertia about hinge location © Copyright 2022 by NuScale Power, LLC C-10

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 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: 2 I h = I o + m eff d Equation C-12 Where 1 2 1 2 1 2 I o = m eff --- r + --- r + ------ L Equation C-13 4 o 4 i 12 h Where r o and r i terms above (as well as in Equation C-15) represent the pipe's outer radius, R1 and inner radius, R2, respectively. Capitalized terms R1 and R2 are from the mass moment of inertia cell for row 9 of Table 3.2 in Reference 6.1.48. L d = -----h- Equation C-14 2 Then, substituting 1 2 1 2 1 2 I h = m eff --- r + --- r + --- L Equation C-15 4 o 4 i 3 h The work-energy equation can be re-written and solved for the tangential linear velocity at the point of impact: 1 2 F b - M p = --- I h Equation C-16 2 2 1 Vt F b - M p = --- I h ------ Equation C-17 2 L h Solving for the pipe velocity at impact: 2 12 V t = L h ---- ( F b - M p ) Equation C-18 Ih © Copyright 2022 by NuScale Power, LLC C-11

Pipe Rupture Hazards Analysis TR-121507-NP Revision 0 C.2.2 SC Wall Impact by Main Steam System Pipe Whip This section proves the steel-plate composite (SC) pool wall resists perforating its rear plate because of the dynamic impact of the MS pipe whipping onto it as an ideal missile. The basic methodology is standardized but without regard to the modified equation numbers. Given a MS line break in the pipe gallery near the NPM bay wall, a hinge length, Lh, of (( }}2(a)(c) 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 based on Lh exceeding the maximum value to prevent pipe whip. The assumed rotation of the pipe extends over a 90° sector, (i.e.,

             = 90° (Figure C-1)).

Equations from step 2 onward of this evaluation are based on the paper by J. C. Bruhl et al (Reference 6.1.47). The three-step approach is presented based on the failure mechanism depicted in Figure 2 from Reference 6.1.47. Penetration depths per Equation C 6-3 or Equation C 6-4 of DOE-STD-3014-96 are not applicable to SC-walls because as stated in Reference 6.1.47, tests include steel plates that are often interpreted to provide additional (equivalent) concrete thickness, which is not accounted for in the equation for penetration depth. Instead, proof that the back plate cannot be penetrated is shown. The presented approach is summarized below:

1. Selecting an initial concrete wall thickness, Tc. For an existing design, this is the SC wall concrete thickness.

2. Computing the concrete conical plug weight (WCP) and residual velocity (Vr) of the concrete conical plug dislodged by the missile after penetrating into the SC wall.

3. Calculating the required thickness, trp, of the rear steel faceplate to prevent its tearing fracture and thus perforation of the SC wall due to the concrete plug projectile.

Figure 2 from Reference 6.1.47 describes the impact sequence of the missile on SC walls. ((

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