ML23005A305

LLC Submittal of Topical Report Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4

ML23005A305
Person / Time
Site:	99902078, 05200050
Issue date:	01/05/2023
From:	Shaver M NuScale
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23005A304	List: ML23005A305
References
LO-133397
Download: ML23005A305 (1)

Text

LO-133397 January 5, 2023 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 Topical Report Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4

REFERENCES:

1. Letter from NuScale to NRC, NuScale Power, LLC Submittal of the Approved Version of NuScale Topical Report, Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 3, dated July 9, 2020 (ML20191A281)

NuScale Power, LLC (NuScale) hereby submits Revision 4 of the Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416. The purpose of this submittal is to request that the NRC review and approve the changes from the previously approved version in Reference 1. NuScale respectfully requests that the acceptance review be completed in 60 days from the date of transmittal. contains the proprietary version of the report entitled Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. Enclosure 2 contains the nonproprietary version of the report.

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

If you have any questions, please contact Thomas Griffith at 541-452-7813 or tgriffith@nuscalepower.com.

Sincerely, Mark W. Shaver Manager, Licensing NuScale Power, LLC Distribution: Michael Dudek, NRC Getachew Tesfaye, NRC Bruce Bavol, NRC NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-133397 Page 2 of 2 01/05/23 Enclosure 1: Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4, Proprietary Version Enclosure 2: Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4, Nonproprietary Version Enclosure 3: Affidavit of Mark W. Shaver, AF-133398 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-133397 :

Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4, Proprietary Version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-133397 :

Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416, Revision 4, 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

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Licensing Topical Report Non-Loss-of-Coolant Accident Analysis Methodology December 2022 Revision 4 Docket: 52-050 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

© Copyright 2022 by NuScale Power, LLC

© Copyright 2022 by NuScale Power, LLC i

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Licensing Topical Report COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

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

Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.

© Copyright 2022 by NuScale Power, LLC ii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Licensing Topical Report Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-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.

© Copyright 2022 by NuScale Power, LLC iii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table of Contents Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Executive Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.3 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.0 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.1 Non-LOCA Evaluation Model Roadmap . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.2 Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.0 Plant Design Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.1 Description of NuScale Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2 Plant Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.3 Decay Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.4 Emergency Core Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.5 Other Important Systems and Functions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.0 Transient and Accident Analysis Overview. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.1 Design-Basis Events and Event Classification. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.2 Design Basis Event Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.3 Non-LOCA Transient Analysis Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 4.3.1 Develop a Plant Base Model NRELAP5 Input . . . . . . . . . . . . . . . . . . . . . . . . . . 37 4.3.2 Adapt Plant Base Model NRELAP5 Input for Event Transient Analysis. . . . . . . 41 4.3.3 Perform NRELAP5 Steady State and Transient System Analysis Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 4.3.4 Evaluate Results of Transient Analysis Calculations . . . . . . . . . . . . . . . . . . . . . 42 4.3.5 Identification of Cases for Subchannel Analysis and Extraction of Boundary Condition Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 4.3.6 Identification of Cases for Accident Radiological Analysis . . . . . . . . . . . . . . . . . 47 5.0 NRELAP5 Applicability for Non-LOCA Transient Analysis . . . . . . . . . . . . . . . . . . . 50 5.1 Non-LOCA Phenomena Identification and Ranking Table and Evaluation of High-Ranked Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 5.1.1 Phenomena Identification and Ranking Table Process . . . . . . . . . . . . . . . . . . . 50 5.1.2 Non-LOCA Event Scenarios and Phases. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51

© Copyright 2022 by NuScale Power, LLC iv

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table of Contents 5.1.3 Phenomena Identification and Ranking Table Figures-of-Merit and Phenomenon Ranking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 5.1.4 Highly Ranked Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 5.2 Evaluation of Non-LOCA Phenomena Identification and Ranking Table High-Ranked Phenomena . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 115 5.3 NRELAP5 Validation and Assessments for Non-LOCA. . . . . . . . . . . . . . . . . . . . . . . . 115 5.3.1 KAIST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 116 5.3.2 NIST-1 Decay Heat Removal System Separate Effects Tests. . . . . . . . . . . . . 122 5.3.3 NIST-1 Non-LOCA Integral Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 194 5.3.4 Code to Code Benchmark for Integral Assessment of Reactivity Event Response . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 318 5.3.5 Steam Generator Modeling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 340 5.3.6 Heat Transfer Correlation Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 350 5.3.7 NIST-2 Steam Generator - Decay Heat Removal System Integral Effects Tests. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 354 5.4 Conclusions of NRELAP5 Applicability for Non-LOCA . . . . . . . . . . . . . . . . . . . . . . . . 404 6.0 NuScale NRELAP5 Plant Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 409 6.1 Thermal-Hydraulic Volumes and Heat Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . 409 6.1.1 Reactor Primary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 413 6.1.2 Core Kinetics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 423 6.1.3 Fuel Rod Design Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 423 6.1.4 Secondary System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 424 6.1.5 Decay Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 428 6.1.6 Emergency Core Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 430 6.1.7 Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 430 6.1.8 Reactor Cooling Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 433 6.2 Material Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 433 6.3 Control Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 433 6.3.1 Module Control System (Nonsafety-related) . . . . . . . . . . . . . . . . . . . . . . . . . . 434 6.3.2 Module Protection System (Safety-related) . . . . . . . . . . . . . . . . . . . . . . . . . . . 436 7.0 Non-LOCA Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 439 7.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 439 7.1.1 Achieving Steady State Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 439

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table of Contents 7.1.2 Treatment of Plant Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 442 7.1.3 Loss of Power Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 444 7.1.4 Single Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 447 7.1.5 Bounding Reactivity Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 450 7.1.6 Biasing of Other Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 453 7.1.7 Credit for Nonsafety-related Components or Operator Actions . . . . . . . . . . . . 459 7.2 Event Specific Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460 7.2.1 Decrease in Feedwater Temperature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 462 7.2.2 Increase in Feedwater Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 469 7.2.3 Increase in Steam Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 474 7.2.4 Steam System Piping Failure Inside or Outside of Containment . . . . . . . . . . . 479 7.2.5 Containment Flooding / Loss of Containment Vacuum . . . . . . . . . . . . . . . . . . 485 7.2.6 Turbine Trip / Loss of External Load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 490 7.2.7 Loss of Condenser Vacuum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 494 7.2.8 Main Steam Isolation Valve(s) Closure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 496 7.2.9 Loss of Nonemergency AC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 499 7.2.10 Loss of Normal Feedwater Flow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 504 7.2.11 Inadvertent Decay Heat Removal System Actuation . . . . . . . . . . . . . . . . . . . . 507 7.2.12 Feedwater System Pipe Break Inside or Outside Containment . . . . . . . . . . . . 512 7.2.13 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 516 7.2.14 Uncontrolled Control Rod Assembly Bank Withdrawal at Power . . . . . . . . . . . 521 7.2.15 Control Rod Misoperation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 526 7.2.16 Inadvertent Decrease in Boron Concentration . . . . . . . . . . . . . . . . . . . . . . . . . 537 7.2.17 Chemical and Volume Control System Malfunction that Increases Reactor Coolant System Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544 7.2.18 Failure of Small Lines Outside Containment . . . . . . . . . . . . . . . . . . . . . . . . . . 548 7.2.19 Steam Generator Tube Failure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 556 8.0 Representative Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 563 8.1 Cooldown and/or Depressurization of the Reactor Coolant System . . . . . . . . . . . . . . 563 8.1.1 Decrease in Feedwater Temperature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 563 8.1.2 Increase in Steam Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 571

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table of Contents 8.1.3 Main Steam Line Break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 579 8.2 Heatup and/or Pressurization of the Reactor Coolant System. . . . . . . . . . . . . . . . . . . 586 8.2.1 Loss of Normal Feedwater Flow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 586 8.2.2 Loss of Normal AC Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 600 8.2.3 Feedwater Line Break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 605 8.3 Reactivity Anomaly. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 612 8.3.1 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 612 8.3.2 Control Rod Misoperation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 619 8.4 Increase in Reactor Coolant System Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 626 8.4.1 Chemical and Volume Control System Malfunction that Increases Reactor Coolant System Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 626 8.5 Decrease in Reactor Coolant System Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 635 8.5.1 Small Line Break Outside of Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 635 8.5.2 Steam Generator Tube Failure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 643 9.0 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 652 10.0 Summary and Conclusions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 653 11.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 655

© Copyright 2022 by NuScale Power, LLC vii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 1-1 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Table 1-2 Definitions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Table 2-1 Evaluation model development and assessment process steps and associated application in the non-LOCA evaluation model . . . . . . . . . . . . . . . . 16 Table 4-1 Design basis events for which the non-LOCA system transient analysis is performed, event category, and event classification . . . . . . . . . . . . . . . . . . . . . 31 Table 4-2 Acceptance criteria for anticipated operational occurrences . . . . . . . . . . . . . . . 34 Table 4-3 Acceptance criteria for infrequent events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 Table 4-4 Acceptance criteria for postulated accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 Table 5-1 Importance rankings. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 Table 5-2 Knowledge levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 Table 5-3 High-ranked phenomena for non-LOCA events . . . . . . . . . . . . . . . . . . . . . . . . . 55 Table 5-4 Comparison between NuScale Power Module decay heat removal system and KAIST test section dimensions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 Table 5-5 Comparison between NuScale Power Module decay heat removal system and KAIST range of operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 Table 5-6 Comparison between NuScale Power Module decay heat removal system and KAIST NRELAP5 model nodalization . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 Table 5-7 NIST-1 decay heat removal system separate effects tests for NRELAP5 code validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 136 Table 5-8 Comparison between NPM and NIST-1 decay heat removal NRELAP5 nodalizations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138 Table 5-9 NIST-1 HP-03 test cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 140 Table 5-10 NIST-1 HP04 test ranges . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 Table 5-11 NIST-1 integral effects tests for NRELAP5 code validation . . . . . . . . . . . . . . . 194 Table 5-12 NLT-02a sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 202 Table 5-13 NLT-02b sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 215 Table 5-14 NLT-15p2 sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 287 Table 5-15 Non-LOCA transients helical coil steam generator operating range vs.

NRELAP5 validated range . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 342 Table 5-16 Comparison of correlation to CFD results - CFD model 2 . . . . . . . . . . . . . . . . 350 Table 5-17 NIST-2 integral effects SG/DHRS testing for NRELAP5 code validation . . . . . 376 Table 5-18 NIST-2 Run 1 sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 377 Table 6-1 Typical reactor coolant system regions and associated NRELAP5 components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 413 Table 7-1 Typical list of initial conditions considered . . . . . . . . . . . . . . . . . . . . . . . . . . . . 441

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 7-2 Example of normalized trip worth vs. time after trip . . . . . . . . . . . . . . . . . . . . . 453 Table 7-3 Examples of analytical limits and actuation delays (reactor trip system and engineered safety features actuation system) . . . . . . . . . . . . . . . . . . . . . . . . . 459 Table 7-4 Regulatory acceptance criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 460 Table 7-5 Acceptance criteria, single active failure, loss of power scenarios -

decrease in feedwater temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 463 Table 7-6 Acceptance criteria - decrease in feedwater temperature . . . . . . . . . . . . . . . . 463 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 464 Table 7-8 Representative fuel exposure study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 468 Table 7-9 Representative fuel temperature study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 468 Table 7-10 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 468 Table 7-11 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 469 Table 7-12 Acceptance criteria, single active failure, loss of power scenarios - increase in feedwater flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 470 Table 7-13 Acceptance criteria - increase in feedwater flow . . . . . . . . . . . . . . . . . . . . . . . 470 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow. . . . 471 Table 7-15 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 474 Table 7-16 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 474 Table 7-17 Acceptance criteria, single active failure, loss of power scenarios - increase in steam flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 475 Table 7-18 Acceptance criteria - increase in steam flow . . . . . . . . . . . . . . . . . . . . . . . . . . 476 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 477 Table 7-20 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 479 Table 7-21 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 479 Table 7-22 Acceptance criteria, single active failure, loss of power scenarios - steam line break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 481 Table 7-23 Acceptance criteria - steam line break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 481 Table 7-24 Initial conditions, biases, and conservatisms - steam line break . . . . . . . . . . . 482 Table 7-25 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 485 Table 7-26 Acceptance criteria, single active failure, loss of power scenarios -

containment flooding / loss of containment vacuum. . . . . . . . . . . . . . . . . . . . . 486 Table 7-27 Acceptance criteria - containment flooding / loss of containment vacuum. . . . 487 Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 487

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 7-29 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 490 Table 7-30 Acceptance criteria, single active failure, loss of power scenarios - turbine trip / loss of external load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 491 Table 7-31 Acceptance criteria - turbine trip / loss of external load . . . . . . . . . . . . . . . . . . 491 Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 492 Table 7-33 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 494 Table 7-34 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 495 Table 7-35 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 495 Table 7-36 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 495 Table 7-37 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 495 Table 7-38 Acceptance criteria, single active failure, loss of power scenarios - main steam isolation valve closure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 496 Table 7-39 Acceptance criteria - main steam isolation valve closure. . . . . . . . . . . . . . . . . 497 Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 497 Table 7-41 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 499 Table 7-42 Acceptance criteria, single active failure, loss of power scenarios - loss of normal AC power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 501 Table 7-43 Acceptance criteria - loss of normal AC power . . . . . . . . . . . . . . . . . . . . . . . . 501 Table 7-44 Initial conditions, biases, and conservatisms - loss of normal AC power . . . . . 502 Table 7-45 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 504 Table 7-46 Acceptance criteria, single active failure, loss of power scenarios - loss of normal feedwater flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 504 Table 7-47 Acceptance criteria - loss of normal feedwater flow . . . . . . . . . . . . . . . . . . . . . 505 Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 505 Table 7-49 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 507 Table 7-50 Acceptance criteria, single active failure, loss of power scenarios -

inadvertent decay heat removal system actuation . . . . . . . . . . . . . . . . . . . . . . 509 Table 7-51 Acceptance criteria - inadvertent decay heat removal system actuation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 509 Table 7-52 Initial conditions, biases, and conservatisms - inadvertent decay heat removal system actuation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 510 Table 7-53 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 512

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 7-54 Acceptance criteria, single active failure, loss of power scenarios -

feedwater line break . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 513 Table 7-55 Acceptance criteria - feedwater line break . . . . . . . . . . . . . . . . . . . . . . . . . . . . 513 Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break . . . . . . . . 514 Table 7-57 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 516 Table 7-58 Acceptance criteria, single active failure, loss of power scenarios -

uncontrolled control rod bank withdrawal from subcritical or low power startup conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 518 Table 7-59 Acceptance criteria - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 518 Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions. . . . . . . . . . . . . . . 519 Table 7-61 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 521 Table 7-62 Acceptance criteria, single active failure, loss of power scenarios -

uncontrolled control rod bank withdrawal at power . . . . . . . . . . . . . . . . . . . . . 522 Table 7-63 Acceptance criteria - uncontrolled control rod bank withdrawal at power . . . . 523 Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 524 Table 7-65 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 526 Table 7-66 Acceptance criteria, single active failure, loss of power scenarios - control rod misoperation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 529 Table 7-67 Acceptance criteria - control rod misoperation . . . . . . . . . . . . . . . . . . . . . . . . . 532 Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 533 Table 7-69 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 535 Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 535 Table 7-71 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 537 Table 7-72 Acceptance criteria, single active failure, loss of power scenarios -

inadvertent decrease in boron concentration . . . . . . . . . . . . . . . . . . . . . . . . . . 541 Table 7-73 Acceptance criteria - inadvertent decrease in boron concentration . . . . . . . . . 542 Table 7-74 Initial conditions, biases, and conservatisms - inadvertent decrease in boron concentration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 542 Table 7-75 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544 Table 7-76 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544 Table 7-77 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 7-78 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544 Table 7-79 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 544 Table 7-80 Acceptance criteria, single active failure, loss of power scenarios - reactor coolant system inventory increase . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 545 Table 7-81 Acceptance criteria - reactor coolant system inventory increase . . . . . . . . . . . 545 Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 546 Table 7-83 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 548 Table 7-84 Acceptance criteria, single active failure, loss of power scenarios - breaks in small lines carrying primary coolant outside containment . . . . . . . . . . . . . . 551 Table 7-85 Acceptance criteria - breaks in small lines carrying primary coolant outside containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 552 Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 553 Table 7-87 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 556 Table 7-88 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 556 Table 7-89 Acceptance criteria, single active failure, loss of power scenarios - steam generator tube failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 557 Table 7-90 Acceptance criteria - steam generator tube failure. . . . . . . . . . . . . . . . . . . . . . 558 Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 560 Table 7-92 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 562 Table 8-1 Decrease in feedwater temperature sequence of events. . . . . . . . . . . . . . . . . 565 Table 8-2 Increase in steam flow sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . 573 Table 8-3 Main steam line break sequence of events . . . . . . . . . . . . . . . . . . . . . . . . . . . 581 Table 8-4 Loss of normal feedwater flow sequence of events - reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 588 Table 8-5 Loss of normal feedwater flow sequence of events - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 595 Table 8-6 Sequence of events for loss of AC power . . . . . . . . . . . . . . . . . . . . . . . . . . . . 601 Table 8-7 Sequence of events for feedwater line break outside containment . . . . . . . . . 607 Table 8-8 Withdrawal of a control rod assembly bank from a low power startup condition sequence of events. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 614 Table 8-9 Single rod withdrawal sequence of events - MCHFR case . . . . . . . . . . . . . . . 620 Table 8-10 Increase in reactor coolant system inventory sequence of events. . . . . . . . . . 627

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Tables Table 8-11 Sequence of events for small line breaks carrying primary coolant outside containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 637 Table 8-12 Sequence of events for steam generator tube failure . . . . . . . . . . . . . . . . . . . 645

© Copyright 2022 by NuScale Power, LLC xiii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 2-1 Evaluation model development and assessment process . . . . . . . . . . . . . . . . . 15 Figure 3-1 NuScale Power Module schematic. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 Figure 4-1 ((2(a),(c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 Figure 4-2 ((

                                                                              }}2(a),(c) . . . . . . . . . . . . . . . . . . . . . . . . 46 Figure 4-3         ((
                                       }}2(a),(c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 Figure 5-1         Schematic of KAIST test facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 118 Figure 5-2         Measured vs predicted heat transfer coefficient. . . . . . . . . . . . . . . . . . . . . . . . 121 Figure 5-2a        Measured vs predicted heat transfer coefficient. . . . . . . . . . . . . . . . . . . . . . . . 121 Figure 5-3         Schematic of NIST-1 integral test facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 Figure 5-4         NIST-1 test facility configuration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 124 Figure 5-5         Photograph of the NIST-1 facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 126 Figure 5-6         Reactor pressure vessel thermal-hydraulic regions . . . . . . . . . . . . . . . . . . . . . 128 Figure 5-7         Lower core flow plate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130 Figure 5-8         Full-height decay heat removal condenser . . . . . . . . . . . . . . . . . . . . . . . . . . . 132 Figure 5-9         Scaled decay heat removal heat condensers . . . . . . . . . . . . . . . . . . . . . . . . . 133 Figure 5-10        NIST-1 noding diagram for full-height decay heat removal system separate effect tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 139 Figure 5-11        NIST-1 HP-03-01 decay heat removal system enthalpy flow rate code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 143 Figure 5-11a NIST-1 HP-03-01 decay heat removal system power code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144 Figure 5-12        NIST-1 HP-03-01 decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 145 Figure 5-12a NIST-1 HP-03-01 decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146 Figure 5-13        NIST-1 HP-03-01 decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 147 Figure 5-13a NIST-1 HP-03-01 decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148 Figure 5-14        NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison (1 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 149 Figure 5-14a NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison (2 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 150

© Copyright 2022 by NuScale Power, LLC xiv

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-14b NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison . 151 Figure 5-15 NIST-1 HP-03-01 cooling pool vessel level code-to-data comparison . . . . . . . 152 Figure 5-15a NIST-1 HP-03-01 cooling pool vessel level code-to-data comparison . . . . . . . 153 Figure 5-16 NIST-1 HP-03-02c decay heat removal system enthalpy flow rate code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 155 Figure 5-16a NIST-1 HP-03-02c decay heat removal system power code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 156 Figure 5-17 NIST-1 HP-03-02c decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 157 Figure 5-17a NIST-1 HP-03-02c decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 158 Figure 5-18 NIST-1 HP-03-02c decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 159 Figure 5-18a NIST-1 HP-03-02c decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 160 Figure 5-19 NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison (1 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161 Figure 5-19a NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison (2 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 162 Figure 5-19b NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 163 Figure 5-20 NIST-1 HP-03-02c cooling pool vessel level code-to-data comparison . . . . . . 164 Figure 5-20a NIST-1 HP-03-02c cooling pool vessel level code-to-data comparison . . . . . . 165 Figure 5-21 NIST-1 HP-03-03-P1 decay heat removal system enthalpy flow rate code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 167 Figure 5-21a NIST-1 HP-03-03-Part1 decay heat removal system power code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 168 Figure 5-22 NIST-1 HP-03-03-Part1 decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 169 Figure 5-22a NIST-1 HP-03-03-Part1 decay heat removal system level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 170 Figure 5-23 NIST-1 HP-03-03-Part1 decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 171 Figure 5-23a NIST-1 HP-03-03-Part1 decay heat removal system internal fluid temperature code-to-data comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 172 Figure 5-24 NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison (1 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 © Copyright 2022 by NuScale Power, LLC xv

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-24a NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison (2 of 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 174 Figure 5-24b NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 175 Figure 5-25 NIST-1 HP-03-03-Part1 cooling pool vessel level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 176 Figure 5-25a NIST-1 HP-03-03-Part1 cooling pool vessel level code-to-data comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 177 Figure 5-26 NIST-1 HP-04-02 decay heat removal system energy transfer rate . . . . . . . . 181 Figure 5-26a NIST-1 HP-04-02 decay heat removal system energy transfer rate . . . . . . . . 181 Figure 5-27 NIST-1 HP-04-02 decay heat removal system condensate temperature. . . . . 182 Figure 5-27a NIST-1 HP-04-02 decay heat removal system condensate temperature. . . . . 182 Figure 5-28 NIST-1 HP-04-02 decay heat removal system internal collapsed level . . . . . . 183 Figure 5-28a NIST-1 HP-04-02 decay heat removal system internal collapsed level . . . . . . 183 Figure 5-29 NIST-1 HP-04-02 cooling pool vessel level . . . . . . . . . . . . . . . . . . . . . . . . . . . 184 Figure 5-29a NIST-1 HP-04-02 cooling pool vessel level . . . . . . . . . . . . . . . . . . . . . . . . . . . 184 Figure 5-30 NIST-1 HP-04-02 mid cooling pool vessel fluid temperatures . . . . . . . . . . . . . 185 Figure 5-30a NIST-1 HP-04-02 mid cooling pool vessel fluid temperatures . . . . . . . . . . . . . 185 Figure 5-31 NIST-1 HP-04-02 upper cooling pool vessel fluid temperatures . . . . . . . . . . . 186 Figure 5-31a NIST-1 HP-04-02 upper cooling pool vessel fluid temperatures . . . . . . . . . . . 186 Figure 5-32 NIST-1 HP-04-03 decay heat removal system energy transfer rate . . . . . . . . 188 Figure 5-32a NIST-1 HP-04-03 decay heat removal system energy transfer rate . . . . . . . . 188 Figure 5-33 NIST-1 HP-04-03 decay heat removal system condensate temperature. . . . . 189 Figure 5-33a NIST-1 HP-04-03 decay heat removal system condensate temperature. . . . . 189 Figure 5-34 NIST-1 HP-04-03 decay heat removal system internal collapsed level comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 190 Figure 5-34a NIST-1 HP-04-03 decay heat removal system internal collapsed level comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 190 Figure 5-35 NIST-1 HP-04-03 cooling pool vessel level comparison . . . . . . . . . . . . . . . . . 191 Figure 5-35a NIST-1 HP-04-03 cooling pool vessel level comparison . . . . . . . . . . . . . . . . . 191 Figure 5-36 NIST-1 HP-04-03 mid cooling pool vessel axial temperatures. . . . . . . . . . . . . 192 Figure 5-36a NIST-1 HP-04-03 mid cooling pool vessel axial temperatures. . . . . . . . . . . . . 192 Figure 5-37 NIST-1 HP-04-03 upper cooling pool vessel axial temperatures . . . . . . . . . . . 193 Figure 5-37a NIST-1 HP-04-03 upper cooling pool vessel axial temperatures . . . . . . . . . . . 193 Figure 5-38 NRELAP5 noding diagram for the NIST-1 facility. . . . . . . . . . . . . . . . . . . . . . . 198 © Copyright 2022 by NuScale Power, LLC xvi

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-39 NRELAP5 NIST-1 model secondary system nodalization layout (NLT-2b) . . . 199 Figure 5-39a NRELAP5 NIST-1 model secondary system nodalization layout (NLT-15p2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 200 Figure 5-40 NIST-1 feedwater split headers at the steam generator tube coils connection (at the time of NLT-2b test) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 201 Figure 5-41 NLT-02a transient feedwater flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . 204 Figure 5-42 NLT-02a transient core heater rod power comparison. . . . . . . . . . . . . . . . . . . 204 Figure 5-43 NLT-02a transient pressurizer pressure comparison . . . . . . . . . . . . . . . . . . . . 205 Figure 5-43a NLT-02a transient pressurizer pressure comparison . . . . . . . . . . . . . . . . . . . . 205 Figure 5-44 NLT-02a transient riser mass flow rate comparison. . . . . . . . . . . . . . . . . . . . . 206 Figure 5-44a NLT-02a transient riser mass flow rate comparison. . . . . . . . . . . . . . . . . . . . . 206 Figure 5-45 NLT-02a transient pressurizer level comparison . . . . . . . . . . . . . . . . . . . . . . . 207 Figure 5-45a NLT-02a transient pressurizer level comparison . . . . . . . . . . . . . . . . . . . . . . . 207 Figure 5-46 NLT-02a transient core inlet temperature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 208 Figure 5-46a NLT-02a transient core inlet temperature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 208 Figure 5-47 NLT-02a transient combined middle and outer steam generator tube coil differential pressure comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 209 Figure 5-48 NLT-02a transient inner steam generator tube coil differential pressure comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 209 Figure 5-49 NLT-02a transient core exit fluid temperature comparison . . . . . . . . . . . . . . . 210 Figure 5-49a NLT-02a transient core exit fluid temperature comparison . . . . . . . . . . . . . . . 210 Figure 5-49b NLT-02a transient riser inlet fluid temperature comparison . . . . . . . . . . . . . . . 211 Figure 5-50 NLT-02a transient pressurizer heater rod power comparison . . . . . . . . . . . . . 211 Figure 5-51 NLT-02a transient steam line pressure comparison. . . . . . . . . . . . . . . . . . . . . 212 Figure 5-52 NLT-02a transient steam line mass flow rate comparison . . . . . . . . . . . . . . . . 213 Figure 5-52a NLT-02a transient steam line mass flow rate comparison . . . . . . . . . . . . . . . . 213 Figure 5-53 NLT-02b phase 1 transient core power comparison . . . . . . . . . . . . . . . . . . . . 219 Figure 5-54 NLT-02b phase 1 transient pressurizer pressure comparison . . . . . . . . . . . . . 220 Figure 5-54a NLT-02b phase 1 transient pressurizer pressure comparison . . . . . . . . . . . . . 220 Figure 5-55 NLT-02b phase 1 transient pressurizer level comparison . . . . . . . . . . . . . . . . 221 Figure 5-55a NLT-02b phase 1 transient pressurizer level comparison . . . . . . . . . . . . . . . . 221 Figure 5-56 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 221 Figure 5-57 NLT-02b phase 1 transient core inlet and outlet temperature comparison . . . 222 Figure 5-57a NLT-02b phase 1 transient core inlet and outlet temperature comparison . . . 222 © Copyright 2022 by NuScale Power, LLC xvii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-58 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 223 Figure 5-59 NLT-02b phase 1 transient steam generator steam pressure comparison . . . 223 Figure 5-59a NLT-02b phase 1 transient steam generator steam pressure comparison . . . 223 Figure 5-60 NLT-02b phase 1 transient steam generator thermal power comparison . . . . 224 Figure 5-61 NLT-02b phase 1 transient decay heat removal system heat exchanger thermal power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 224 Figure 5-62 NLT-02b phase 1 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization . . . . . . . . . . . . 225 Figure 5-62a NLT-02b phase 1 integrated compensation flow for SG and DHRS HX level equalization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 225 Figure 5-63 NLT-02b phase 1 transient steam generator level comparison . . . . . . . . . . . . 226 Figure 5-64 NLT-02b phase 1 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 227 Figure 5-64a NLT-02b phase 1 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 227 Figure 5-65 NLT-02b phase 1 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 228 Figure 5-65a NLT-02b phase 1 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 228 Figure 5-66 NLT-02b phase 1 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 229 Figure 5-66a NLT-02b phase 1 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 229 Figure 5-67 NLT-02b phase 1 transient steam generator outlet temperature comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 230 Figure 5-68 NLT-02b phase 1 transient cooling pool vessel level comparison . . . . . . . . . . 230 Figure 5-69 NLT-02b phase 1 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) . . . . . . . . . . 231 Figure 5-69a NLT-02b phase 1 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger) . . . 231 Figure 5-69b NLT-02b phase 1 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 232 Figure 5-70 NLT-02b phase 1 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 232 Figure 5-71 NLT-02b phase 2 transient core power comparison . . . . . . . . . . . . . . . . . . . . 235 Figure 5-72 NLT-02b phase 2 transient pressurizer pressure comparison . . . . . . . . . . . . . 236 © Copyright 2022 by NuScale Power, LLC xviii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-72a NLT-02b phase 2 transient pressurizer pressure comparison . . . . . . . . . . . . . 236 Figure 5-73 NLT-02b phase 2 transient pressurizer level comparison . . . . . . . . . . . . . . . . 237 Figure 5-73a NLT-02b phase 2 transient pressurizer level comparison . . . . . . . . . . . . . . . . 237 Figure 5-74 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 237 Figure 5-75 NLT-02b phase 2 transient core inlet and outlet temperature comparison . . . 238 Figure 5-75a NLT-02b phase 2 transient core inlet and outlet temperature comparison . . . 238 Figure 5-76 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 238 Figure 5-77 NLT-02b phase 2 transient steam generator steam pressure comparison . . . 239 Figure 5-77a NLT-02b phase 2 transient steam generator steam pressure comparison . . . 239 Figure 5-78 NLT-02b phase 2 transient steam generator thermal power comparison . . . . 240 Figure 5-79 NLT-02b phase 2 transient decay heat removal system heat exchanger thermal power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 240 Figure 5-80 NLT-02b phase 2 transient steam generator level comparison . . . . . . . . . . . . 241 Figure 5-81 NLT-02b phase 2 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization . . . . . . . . . . . . 242 Figure 5-81a NLT-02b phase 2 integrated compensation flow for SG and DHRS HX level equalization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 242 Figure 5-82 NLT-02b phase 2 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 243 Figure 5-82a NLT-02b phase 2 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 243 Figure 5-83 NLT-02b phase 2 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 244 Figure 5-83a NLT-02b phase 2 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 244 Figure 5-84 NLT-02b phase 2 transient steam generator outlet temperature comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 245 Figure 5-85 NLT-02b phase 2 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 246 Figure 5-85a NLT-02b phase 2 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 246 Figure 5-86 NLT-02b phase 2 transient cooling pool vessel level comparison . . . . . . . . . . 247 Figure 5-87 NLT-02b phase 2 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) . . . . . . . . . . 248 Figure 5-87a NLT-02b phase 2 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger) . . . 248 © Copyright 2022 by NuScale Power, LLC xix

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-87b NLT-02b phase 2 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 249 Figure 5-88 NLT-02b phase 2 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 249 Figure 5-89 NLT-02b phase 3 transient core power comparison . . . . . . . . . . . . . . . . . . . . 253 Figure 5-90 NLT-02b phase 3 transient pressurizer pressure comparison . . . . . . . . . . . . . 254 Figure 5-90a NLT-02b phase 3 transient pressurizer pressure comparison . . . . . . . . . . . . . 254 Figure 5-91 NLT-02b phase 3 transient pressurizer level comparison . . . . . . . . . . . . . . . . 255 Figure 5-91a NLT-02b phase 3 transient pressurizer level comparison . . . . . . . . . . . . . . . . 255 Figure 5-92 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 256 Figure 5-93 NLT-02b phase 3 transient core inlet and outlet temperature comparison . . . 256 Figure 5-93a NLT-02b phase 3 transient core inlet and outlet temperature comparison . . . 256 Figure 5-94 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 257 Figure 5-95 NLT-02b phase 3 transient steam generator steam pressure comparison . . . 257 Figure 5-95a NLT-02b phase 3 transient steam generator steam pressure comparison . . . 257 Figure 5-96 NLT-02b phase 3 transient steam generator thermal power comparison . . . . 258 Figure 5-97 NLT-02b phase 3 transient decay heat removal system heat exchanger thermal power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 258 Figure 5-98 NLT-02b phase 3 transient steam generator level comparison . . . . . . . . . . . . 259 Figure 5-99 NLT-02b phase 3 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization . . . . . . . . . . . . 260 Figure 5-99a NLT-02b phase 3 integrated compensation flow for SG and DHRS HX level equalization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 260 Figure 5-100 NLT-02b phase 3 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 261 Figure 5-100a NLT-02b phase 3 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 261 Figure 5-101 NLT-02b phase 3 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 262 Figure 5-101a NLT-02b phase 3 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 262 Figure 5-102 NLT-02b phase 3 transient steam generator outlet temperature comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 263 Figure 5-103 NLT-02b phase 3 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 264 © Copyright 2022 by NuScale Power, LLC xx

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-103a NLT-02b phase 3 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 264 Figure 5-104 NLT-02b phase 3 transient cooling pool vessel level comparison . . . . . . . . . . 265 Figure 5-105 NLT-02b phase 3 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) . . . . . . . . . . 266 Figure 5-105a NLT-02b phase 3 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger) . . . 266 Figure 5-105b NLT-02b phase 3 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 267 Figure 5-106 NLT-02b phase 3 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 267 Figure 5-107 NLT-02b phase 4 transient core power comparison . . . . . . . . . . . . . . . . . . . . 270 Figure 5-108 NLT-02b phase 4 transient pressurizer pressure comparison . . . . . . . . . . . . . 271 Figure 5-108a NLT-02b phase 4 transient pressurizer pressure comparison . . . . . . . . . . . . . 271 Figure 5-109 NLT-02b phase 4 transient pressurizer level comparison . . . . . . . . . . . . . . . . 272 Figure 5-109a NLT-02b phase 4 transient pressurizer level comparison . . . . . . . . . . . . . . . . 272 Figure 5-110 NLT-02b phase 4 transient reactor pressure vessel level . . . . . . . . . . . . . . . . 273 Figure 5-111 NLT-02b phase 4 transient reactor pressure vessel upper plenum temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 273 Figure 5-112 NLT-02b phase 4 transient core inlet and outlet temperature comparison . . . 274 Figure 5-112a NLT-02b phase 4 transient core inlet and outlet temperature comparison . . . 274 Figure 5-113 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 275 Figure 5-114 NLT-02b phase 4 transient steam generator steam pressure comparison . . . 275 Figure 5-114a NLT-02b phase 4 transient steam generator steam pressure comparison . . . 275 Figure 5-115 NLT-02b phase 4 transient steam generator thermal power comparison . . . . 276 Figure 5-116 NLT-02b phase 4 transient decay heat removal system heat exchanger thermal power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 276 Figure 5-117 NLT-02b phase 4 transient steam generator level comparison . . . . . . . . . . . . 277 Figure 5-118 NLT-02b phase 4 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization . . . . . . . . . . . . 278 Figure 5-118a NLT-02b phase 4 integrated compensation flow for SG and DHRS HX level equalization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 278 Figure 5-119 NLT-02b phase 4 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 279 © Copyright 2022 by NuScale Power, LLC xxi

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-119a NLT-02b phase 4 transient decay heat removal system condensate temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 279 Figure 5-120 NLT-02b phase 4 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 280 Figure 5-120a NLT-02b phase 4 transient decay heat removal system condensate flow comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 280 Figure 5-121 NLT-02b phase 4 transient steam generator outlet temperature comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 281 Figure 5-122 NLT-02b phase 4 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 282 Figure 5-122a NLT-02b phase 4 transient decay heat removal system heat exchanger level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 282 Figure 5-123 NLT-02b phase 4 transient cooling pool vessel level comparison . . . . . . . . . . 283 Figure 5-124 NLT-02b phase 4 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) . . . . . . . . . . 283 Figure 5-124a NLT-02b phase 4 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger) . . . 284 Figure 5-124b NLT-02b phase 4 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 284 Figure 5-125 NLT-02b phase 4 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 285 Figure 5-126 NLT-15p2, transient core power. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 294 Figure 5-127 NLT-15p2, transient RPV pressure short term . . . . . . . . . . . . . . . . . . . . . . . . . 295 Figure 5-128 NLT-15p2, transient RPV pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 296 Figure 5-128a NLT-15p2, transient RPV pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 296 Figure 5-129 NLT-15p2, transient pressurizer level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 297 Figure 5-129a NLT-15p2, transient pressurizer level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 297 Figure 5-130 NLT-15p2, transient RPV level. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 298 Figure 5-130a NLT-15p2, transient RPV level. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 298 Figure 5-131 NLT-15p2, transient riser mass flow rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 299 Figure 5-132 NLT-15p2, transient core inlet temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . 299 Figure 5-133 NLT-15p2, transient riser inlet temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . 300 Figure 5-133a NLT-15p2, transient core inlet and riser inlet temperatures . . . . . . . . . . . . . . . 300 Figure 5-134 NLT-15p2, transient upper plenum temperature . . . . . . . . . . . . . . . . . . . . . . . 301 © Copyright 2022 by NuScale Power, LLC xxii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-134a NLT-15p2, transient upper plenum temperature . . . . . . . . . . . . . . . . . . . . . . . 301 Figure 5-135 NLT-15p2, transient secondary side pressure - 0 to 500 seconds. . . . . . . . . . 302 Figure 5-136 NLT-15p2, transient DHRS loop flow - 0 to 500 seconds. . . . . . . . . . . . . . . . . 302 Figure 5-137 NLT-15p2, transient measured steam line temperatures - 0 to 500 seconds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 303 Figure 5-138 NLT-15p2, transient DHRS HX level - 0 to 500 seconds . . . . . . . . . . . . . . . . . 303 Figure 5-139 NLT-15p2, transient secondary side pressure . . . . . . . . . . . . . . . . . . . . . . . . . 304 Figure 5-139a NLT-15p2, transient secondary side pressure . . . . . . . . . . . . . . . . . . . . . . . . . 304 Figure 5-140 NLT-15p2, transient DHRS HX inlet temperature . . . . . . . . . . . . . . . . . . . . . . 305 Figure 5-140a NLT-15p2, transient DHRS HX inlet temperature . . . . . . . . . . . . . . . . . . . . . . 305 Figure 5-141 NLT-15p2, transient DHRS HX outlet temperature . . . . . . . . . . . . . . . . . . . . . 306 Figure 5-142 NLT-15p2, transient DHRS loop flow - short term . . . . . . . . . . . . . . . . . . . . . . 306 Figure 5-143 NLT-15p2, transient DHRS loop flow rate - long term . . . . . . . . . . . . . . . . . . . 307 Figure 5-143a NLT-15p2, transient DHRS loop flow rate - long term . . . . . . . . . . . . . . . . . . . 307 Figure 5-144 NLT-15p2, transient DHRS HX level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 308 Figure 5-144a NLT-15p2, transient DHRS HX level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 308 Figure 5-145 NLT-15p2, transient steam generator tube coil level - long term . . . . . . . . . . . 309 Figure 5-146 NLT-15p2, transient steam generator tube coil level - short term . . . . . . . . . . 309 Figure 5-147 NLT-15p2, transient DHRS condensate line differential pressure . . . . . . . . . . 310 Figure 5-148 NLT-15p2, transient DHRS steam line differential pressure . . . . . . . . . . . . . . 310 Figure 5-149 NLT-15p2, transient steam generator tube coil power removal . . . . . . . . . . . . 311 Figure 5-149a NLT-15p2, transient steam generator tube coil power removal . . . . . . . . . . . . 311 Figure 5-150 NLT-15p2, transient DHRS power removal . . . . . . . . . . . . . . . . . . . . . . . . . . . 312 Figure 5-150a NLT-15p2, transient DHRS power removal . . . . . . . . . . . . . . . . . . . . . . . . . . . 312 Figure 5-151 NLT-15p2, transient cooling pool vessel level . . . . . . . . . . . . . . . . . . . . . . . . . 313 Figure 5-152 NLT-15p2, transient cooling pool vessel fluid temperature at level 3 (below decay heat removal heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 313 Figure 5-153 NLT-15p2, transient cooling pool vessel fluid temperature at level 5 (near bottom of DHRS heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 314 Figure 5-154 NLT-15p2, transient cooling pool vessel fluid temperature at level 6 (near midpoint of DHRS heat exchanger) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 314 Figure 5-155 NLT-15p2, transient cooling pool vessel fluid temperature at level 7 (top to just above DHRS heat exchanger region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 315 Figure 5-156 NLT-15p2, transient cooling pool vessel fluid temperature at level 8 (above DHRS heat exchanger region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 315 © Copyright 2022 by NuScale Power, LLC xxiii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-157 NLT-15p2, transient cooling pool vessel fluid temperature at level 9 (above DHRS heat exchanger region) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 316 Figure 5-158 Core power (fast uncontrolled rod withdrawal). . . . . . . . . . . . . . . . . . . . . . . . . 322 Figure 5-159 Total reactivity (fast uncontrolled rod withdrawal) . . . . . . . . . . . . . . . . . . . . . . 322 Figure 5-160 Pressurizer pressure (fast uncontrolled rod withdrawal) . . . . . . . . . . . . . . . . . 323 Figure 5-161 Pressurizer level (fast uncontrolled rod withdrawal). . . . . . . . . . . . . . . . . . . . . 323 Figure 5-162 Core flow (fast uncontrolled rod withdrawal) . . . . . . . . . . . . . . . . . . . . . . . . . . 324 Figure 5-163 Core inlet temperatures (fast uncontrolled rod withdrawal) . . . . . . . . . . . . . . . 324 Figure 5-164 Core outlet temperatures (fast uncontrolled rod withdrawal) . . . . . . . . . . . . . . 325 Figure 5-165 Core power (slow uncontrolled rod withdrawal) . . . . . . . . . . . . . . . . . . . . . . . . 327 Figure 5-166 Total reactivity (slow uncontrolled rod withdrawal). . . . . . . . . . . . . . . . . . . . . . 327 Figure 5-167 Pressurizer pressure (slow uncontrolled rod withdrawal). . . . . . . . . . . . . . . . . 328 Figure 5-168 Pressurizer level (slow uncontrolled rod withdrawal) . . . . . . . . . . . . . . . . . . . . 328 Figure 5-169 Core flow (slow uncontrolled rod withdrawal). . . . . . . . . . . . . . . . . . . . . . . . . . 329 Figure 5-170 Core inlet temperature (slow uncontrolled rod withdrawal) . . . . . . . . . . . . . . . 329 Figure 5-171 Core outlet temperature (slow uncontrolled rod withdrawal) . . . . . . . . . . . . . . 330 Figure 5-172 Core power (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 331 Figure 5-173 Total reactivity (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 332 Figure 5-174 Pressurizer pressure (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 332 Figure 5-175 Pressurizer level (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 333 Figure 5-176 Core flow (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 333 Figure 5-177 Core inlet temperature (power reduction). . . . . . . . . . . . . . . . . . . . . . . . . . . . . 334 Figure 5-178 Core outlet temperature (power reduction) . . . . . . . . . . . . . . . . . . . . . . . . . . . 334 Figure 5-179 Core power (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 336 Figure 5-180 Total reactivity (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 336 Figure 5-181 Pressurizer pressure (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . . 337 Figure 5-182 Pressurizer level (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 337 Figure 5-183 Core flow (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 338 Figure 5-184 Core inlet temperature (dropped control rod) . . . . . . . . . . . . . . . . . . . . . . . . . . 338 Figure 5-185 Core outlet temperature (dropped control rod). . . . . . . . . . . . . . . . . . . . . . . . . 339 Figure 5-186 Coil 1 representative pressure drop for (( }}2(a),(c) nodes (left) and 2(a),(c) (( }} nodes (right) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 344 Figure 5-187 Coil 1 representative fluid temperature for (( }}2(a),(c) nodes (left) and (( }}2(a),(c) nodes (right) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 344 © Copyright 2022 by NuScale Power, LLC xxiv

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-188 Coil 1 representative wall temperature for (( }}2(a),(c) nodes (left) and 2(a),(c) (( }} nodes (right) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 345 Figure 5-189 Decrease in feedwater temperature nodalization sensitivity steam generator secondary side inlet pressure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 345 Figure 5-190 Decrease in feedwater temperature nodalization sensitivity steam generator secondary side outlet pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 346 Figure 5-191 Decrease in feedwater temperature nodalization sensitivity reactor coolant system flow rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 347 Figure 5-192 Decrease in feedwater temperature nodalization sensitivity reactor coolant system lower plenum pressure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 348 Figure 5-193 Decrease in feedwater temperature nodalization sensitivity reactor coolant system core inlet temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 349 Figure 5-194 Decrease in feedwater temperature nodalization sensitivity reactor power . . . 349 Figure 5-195 Comparison of heat transfer correlation options for steam generator primary side heat transfer rate (linear time scale) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 352 Figure 5-196 Comparison of heat transfer correlation options for steam generator primary side heat transfer rate (logarithmic time scale) . . . . . . . . . . . . . . . . . . . . . . . . 353 Figure 5-197 NIST-2 decay heat removal system configuration . . . . . . . . . . . . . . . . . . . . . . 356 Figure 5-198 Reactor coolant system average temperature transient response for loss of feedwater break scenarios. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 360 Figure 5-199 Reactor coolant system saturation temperature transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 361 Figure 5-200 Reactor coolant system flow transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 362 Figure 5-201 Reactor coolant system pressure transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 363 Figure 5-202 Secondary side pressure transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 364 Figure 5-203 Decay heat removal system flow transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 365 Figure 5-204 Decay heat removal system power transient response for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 366 Figure 5-205 Decay heat removal system power compared to reactor coolant system temperature for loss of feedwater break scenarios . . . . . . . . . . . . . . . . . . . . . 367 Figure 5-206 NIST-2 modified separate effects main steam line part 1 model . . . . . . . . . . . 368 Figure 5-207 NIST-2 modified separate effects main steam line part 2 model . . . . . . . . . . . 369 Figure 5-208 NIST-2 modified bypass steam line to containment vessel model . . . . . . . . . . 370 © Copyright 2022 by NuScale Power, LLC xxv

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-209 NIST-2 new integral effects decay heat removal system steam line model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 371 Figure 5-210 NIST-2 modified separate effects decay heat removal system steam line model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 372 Figure 5-211 NIST-2 new decay heat removal system heat exchanger model . . . . . . . . . . . 373 Figure 5-212 NIST-2 modified decay heat removal system condensate return line model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 374 Figure 5-213 NIST-2 modified feedwater system model . . . . . . . . . . . . . . . . . . . . . . . . . . . . 375 Figure 5-214 NIST-2 Run 1 core power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 381 Figure 5-215 NIST-2 Run 1 steam generator and decay heat removal system active loop inventory comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 382 Figure 5-216 NIST-2 Run 1 steam generator and decay heat removal system heat exchanger power comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 383 Figure 5-217 NIST-2 Run 1 steam generator and decay heat removal system heat exchanger power comparison - short-term. . . . . . . . . . . . . . . . . . . . . . . . 384 Figure 5-218 NIST-2 Run 1 steam generator and decay heat removal system loop flow comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 385 Figure 5-219 NIST-2 Run 1 steam generator and decay heat removal system loop flow comparison - short-term . . . . . . . . . . . . . . . . . . . . . . . . . . . . 386 Figure 5-220 NIST-2 Run 1 primary and secondary pressure comparison . . . . . . . . . . . . . . 387 Figure 5-221 NIST-2 Run 1 steam generator and decay heat removal system level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 388 Figure 5-222 NIST-2 Run 1 steam drum level comparison . . . . . . . . . . . . . . . . . . . . . . . . . . 389 Figure 5-223 NIST-2 Run 1 decay heat removal system heat exchanger inlet and outlet temperatures comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . 390 Figure 5-224 NIST-2 Run 1 steam generator secondary side inlet and outlet temperatures comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 391 Figure 5-225 NIST-2 Run 1 steam generator secondary side saturation temperature comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 392 Figure 5-226 NIST-2 Run 1 steam generator primary side inlet and outlet temperatures comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 393 Figure 5-227 NIST-2 Run 1 pressurizer level comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . 394 Figure 5-228 NIST-2 Run 1 reactor pressure vessel level comparison . . . . . . . . . . . . . . . . . 395 Figure 5-229 NIST-2 Run 1 primary flow rate comparison . . . . . . . . . . . . . . . . . . . . . . . . . . 396 Figure 5-230 NIST-2 Run 1 cooling pool vessel level comparison . . . . . . . . . . . . . . . . . . . . 397 Figure 5-231 NIST-2 Run 1 cooling pool vessel level 5 temperature comparison. . . . . . . . . 398 Figure 5-232 NIST-2 Run 1 cooling pool vessel level 6 temperature comparison. . . . . . . . . 399 © Copyright 2022 by NuScale Power, LLC xxvi

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 5-233 NIST-2 Run 1 decay heat removal system steam line differential pressure comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 400 Figure 5-234 NIST-2 Run 1 decay heat removal system steam line orifice differential pressure comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 401 Figure 5-235 NIST-2 Run 1 decay heat removal system heat exchanger tube lower middle section fluid temperatures - data only . . . . . . . . . . . . . . . . . 402 Figure 5-236 NIST-2 Run 1 decay heat removal system heat exchanger tube bottom fluid temperatures - data only. . . . . . . . . . . . . . . . . . . . . . . . . . . . 403 Figure 6-1 NuScale Power Module (typical) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 410 Figure 6-2 Typical primary and secondary side model (heat structures and component cell details excluded) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 411 Figure 6-3 Typical NRELAP5 plant module volume regions . . . . . . . . . . . . . . . . . . . . . . . 412 Figure 6-4 Typical reactor pressure vessel downcomer model . . . . . . . . . . . . . . . . . . . . . 414 Figure 6-5 Typical core and lower plenum model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 416 Figure 6-6 Reflector / core bypass without fuel assemblies (for illustration only) . . . . . . . 417 Figure 6-7 Lower riser region, immediately above the core (for illustration only) . . . . . . . 418 Figure 6-8 Typical reactor pressure vessel core and lower riser model . . . . . . . . . . . . . . 419 Figure 6-9 Typical reactor pressure vessel upper riser model. . . . . . . . . . . . . . . . . . . . . . 420 Figure 6-10 Typical reactor pressure vessel pressurizer model . . . . . . . . . . . . . . . . . . . . . 422 Figure 6-11 Typical steam generator model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 426 Figure 6-12 Typical main steam system model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 427 Figure 6-13 Typical decay heat removal system division 1 model . . . . . . . . . . . . . . . . . . . 429 Figure 6-14 Not used. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 429 Figure 6-15 Not used. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 430 Figure 6-16 Typical containment and reactor pool model . . . . . . . . . . . . . . . . . . . . . . . . . . 432 Figure 7-1 Example of decay heat comparisons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 452 Figure 7-2 Example of setpoint relationships. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 458 Figure 7-3 Example power reduction vs. initial power and worth for dropped rod event at (( }}2(a),(c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 530 Figure 7-4 Example components of local power vs. time for dropped rod event. . . . . . . . 530 Figure 7-5 Example comparison of non-tripped rod drop events to single rod withdrawal analysis limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 531 Figure 8-1 Temperature of feedwater during the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 566 Figure 8-2 Power response for the representative decrease in feedwater temperature event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 567 © Copyright 2022 by NuScale Power, LLC xxvii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-3 Core outlet temperature for the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 567 Figure 8-4 Steam generator 2 pressure response for the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 568 Figure 8-5 Reactor pressure vessel pressure response for the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 568 Figure 8-6 Pressurizer level for the representative decrease in feedwater temperature event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 569 Figure 8-7 Reactor coolant system flow rate for the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 569 Figure 8-8 Core inlet temperature for the representative decrease in feedwater temperature event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 570 Figure 8-9 Net reactivity for the representative decrease in feedwater temperature event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 570 Figure 8-10 Main steam transient flow rate during the representative increase in steam flow event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 574 Figure 8-11 Steam generator 2 pressure response for the representative increase in steam flow event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 574 Figure 8-12 Steam generator 2 secondary side flow for the representative increase in steam flow event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 575 Figure 8-13 Core inlet temperature for the representative increase in steam flow event . . 575 Figure 8-14 Power response for the representative increase in steam flow event . . . . . . . 576 Figure 8-15 Core outlet temperature for the representative increase in steam flow event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 576 Figure 8-16 Reactor pressure vessel pressure response for the representative increase in steam flow event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 577 Figure 8-17 Pressurizer level for the representative increase in steam flow event . . . . . . . 577 Figure 8-18 Reactor coolant system flow rate for the representative increase in steam flow event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 578 Figure 8-19 Net reactivity for the representative increase in steam flow event . . . . . . . . . . 578 Figure 8-20 Steam generators 1 (unaffected) and 2 (affected) secondary flow rates for the representative main steam line break event. . . . . . . . . . . . . . . . . . . . . . . . 582 Figure 8-21 Core inlet temperature for the representative main steam line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 582 Figure 8-22 Net reactivity for the representative main steam line break event . . . . . . . . . . 583 Figure 8-23 Power response for the representative main steam line break event . . . . . . . 583 Figure 8-24 Steam generators 1 (unaffected) and 2 (affected) pressure response for the representative main steam line break event. . . . . . . . . . . . . . . . . . . . . . . . 584 © Copyright 2022 by NuScale Power, LLC xxviii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-25 Pressurizer level for the representative main steam line break event . . . . . . . 584 Figure 8-26 Reactor pressure vessel pressure response for the representative main steam line break event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 585 Figure 8-27 Reactor coolant system flow rate for the representative main steam line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 585 Figure 8-28 Core outlet temperature for the representative main steam line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 586 Figure 8-29 Reactor pressure vessel pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . 589 Figure 8-30 Pressurizer level for the representative loss of normal feedwater flow event

                  - reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 589 Figure 8-31        Steam generator 2 pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . . . . . . . 590 Figure 8-32        Power response for the representative loss of normal feedwater flow event
                  - reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 590 Figure 8-33        Reactor coolant system flow rate for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . . . . . . . 591 Figure 8-34        Core inlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . . . . . 591 Figure 8-35        Core outlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . 592 Figure 8-36        Net reactivity for the representative loss of normal feedwater flow event -

reactor coolant system pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 592 Figure 8-37 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - reactor coolant system pressure case . . . . . . . . . . . . . 593 Figure 8-38 Core inlet temperature for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 595 Figure 8-39 Core outlet temperature for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 596 Figure 8-40 Reactor pressure vessel pressure response for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . 596 Figure 8-41 Pressurizer level for the representative loss of normal feedwater flow event

                  - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 597 Figure 8-42        Steam generator 2 pressure response for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . 597 Figure 8-43        Power response for the representative loss of normal feedwater flow event
                  - secondary pressure cas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 598

© Copyright 2022 by NuScale Power, LLC xxix

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-44 Reactor coolant system flow rate for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . 598 Figure 8-45 Net reactivity for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 599 Figure 8-46 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - secondary pressure case . . . . . . . . . . . . . . . . . . . . . . . 599 Figure 8-47 Primary temperature response for the representative loss of AC power event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 602 Figure 8-48 System pressure response for the representative loss of AC power event . . . 602 Figure 8-49 Reactor pressure vessel core power response for the representative loss of AC power event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 603 Figure 8-50 Decay heat removal system response for the representative loss of AC power event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 603 Figure 8-51 RSV flow response for the representative loss of AC power event . . . . . . . . . 604 Figure 8-52 Reactor coolant system flow response for the representative loss of AC power event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 604 Figure 8-53 Pressurizer level response for the representative loss of AC power event . . . 605 Figure 8-54 Feedwater line break flow response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 608 Figure 8-55 Primary temperature response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 609 Figure 8-56 System pressure response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 609 Figure 8-57 Reactor pressure vessel core power response for the representative feedwater line break event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 610 Figure 8-58 Reactor safety valve flow response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 610 Figure 8-59 Decay heat removal system response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 611 Figure 8-60 Pressurizer level response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 611 Figure 8-61 Reactor coolant system flow response for the representative feedwater line break event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 612 Figure 8-62 Pressurizer pressure response for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 615 Figure 8-63 Reactor pressure vessel and steam generator pressure responses for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . 615 © Copyright 2022 by NuScale Power, LLC xxx

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-64 Power response for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 616 Figure 8-65 Core inlet temperature for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 616 Figure 8-66 Core inlet density for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 617 Figure 8-67 Core outlet temperature for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 617 Figure 8-68 Reactor coolant system flow rate for the bank withdrawal from a low power startup condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 618 Figure 8-69 Net reactivity for the bank withdrawal from a low power startup condition. . . . 618 Figure 8-70 Power response for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 621 Figure 8-71 Reactor pressure vessel pressure response for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 622 Figure 8-72 Pressurizer level for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 622 Figure 8-73 Core outlet temperature for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 623 Figure 8-74 Steam generator 2 pressure response for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 623 Figure 8-75 Reactor coolant system flow rate for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 624 Figure 8-76 Steam generator 2 secondary flow for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 624 Figure 8-77 Core inlet temperature for the representative single rod withdrawal MCHFR case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 625 Figure 8-78 Net reactivity for the representative single rod withdrawal MCHFR case . . . . 625 Figure 8-79 Makeup flow for increase in reactor coolant system inventory. . . . . . . . . . . . . 628 Figure 8-80 Recirculation pump flow for increase in reactor coolant system inventory. . . . 629 Figure 8-81 Letdown flow for increase in reactor coolant system inventory . . . . . . . . . . . . 629 Figure 8-82 CVCS recirculation flow rate into the reactor pressure vessel for increase in reactor coolant system inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 630 Figure 8-83 CVCS recirculation flow rate out of the reactor pressure vessel for increase in reactor coolant system inventory . . . . . . . . . . . . . . . . . . . . . . . . . . 630 Figure 8-84 Pressure at the bottom of the pressurizer for increase in reactor coolant system inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 631 Figure 8-85 Pressurizer level for increase in reactor coolant system inventory . . . . . . . . . 631 © Copyright 2022 by NuScale Power, LLC xxxi

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-86 Reactor power for increase in reactor coolant system inventory . . . . . . . . . . . 632 Figure 8-87 Reactor coolant system flow for increase in reactor coolant system inventory. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 632 Figure 8-88 Decay heat removal system flow rate for increase in reactor coolant system inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 633 Figure 8-89 Steam generator pressure for increase in reactor coolant system inventory. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 633 Figure 8-90 Core inlet and exit coolant liquid temperature for increase in reactor coolant system inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 634 Figure 8-91 Total reactivity for increase in reactor coolant system inventory . . . . . . . . . . . 634 Figure 8-92 Instantaneous break flow response (0 to 350 sec) for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 638 Figure 8-93 Pressurizer level response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 638 Figure 8-94 Reactor pressure vessel pressure response (0 to 350 sec) for the representative small break outside containment event . . . . . . . . . . . . . . . . . . 639 Figure 8-95 Steam generator pressure responses for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 639 Figure 8-96 Core power response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 640 Figure 8-97 Integrated break flow response (0 to 350 sec) for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 640 Figure 8-98 Reactor coolant system flow rate response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 641 Figure 8-99 Core outlet temperature response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 641 Figure 8-100 Reactor pressure vessel pressure response (0 to 3000 sec) for the representative small break outside containment event . . . . . . . . . . . . . . . . . . 642 Figure 8-101 Net reactivity response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 642 Figure 8-102 Level above top of core response for the representative small break outside containment event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 643 Figure 8-103 Pressurizer level response for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 646 Figure 8-104 Reactor pressure vessel and steam generator pressure responses (0 to 500 sec) for the representative steam generator tube failure event (tube failure occurs in SG1). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 646 © Copyright 2022 by NuScale Power, LLC xxxii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 List of Figures Figure 8-105 Core power response for the representative steam generator tube failure event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 647 Figure 8-106 Instantaneous break flow response for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 647 Figure 8-107 Steam generator level response for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 648 Figure 8-108 Integrated break mass release to steam generator before isolation (0 to 500 sec) for the representative steam generator tube failure event. . . . . . . . . 648 Figure 8-109 Reactor coolant system flow rate response for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 649 Figure 8-110 Core inlet and exit temperature responses for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 649 Figure 8-111 Reactor pressure vessel and steam generator responses (0 to 6000 sec) for the representative steam generator tube failure event . . . . . . . . . . . . . . . . 650 Figure 8-112 Net reactivity response for the representative steam generator tube failure event. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 650 Figure 8-113 Level above top of core response for the representative steam generator tube failure event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 651 © Copyright 2022 by NuScale Power, LLC xxxiii

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Abstract The purpose of this report is to present the NuScale evaluation model (EM) used to evaluate a NuScale Power Module (NPM) short-term system transient response to non-loss-of-coolant accident (non-LOCA) events. The non-LOCA evaluation model is developed following a graded approach to the guidance provided in Regulatory Guide (RG) 1.203 for the evaluation model development and assessment process (EMDAP). This report summarizes the NuScale plant designs, non-LOCA initiating events, classification of the events and acceptance criteria. The scope of the non-LOCA system transient analysis is described in this report, as well as the interfaces to other safety analysis methodologies. The non-LOCA events cover several different event types based on the main effect on the reactor coolant system (RCS). A comprehensive, integrated phenomena identification and ranking table (PIRT) was developed for the range of non-LOCA event types and phases of the event progression. The high-ranked phenomena of the PIRT and how they are assessed are summarized in this report. The NRELAP5 code is the system thermal-hydraulic code used for non-LOCA system transient analysis. Applicability of NRELAP5 for non-LOCA system transient analysis is assessed based on the high ranked phenomena identified in the PIRT. This report describes the selection of appropriately conservative input when applying this EM to perform non-LOCA system transient analyses. The non-LOCA methodology ensures that system transient calculations are executed for sufficient duration to demonstrate that the initiating event is mitigated and stable cooling is established. For non-LOCA initiating events that actuate the decay heat removal system, the EM is applicable for the short-term transient progression; during this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained. Representative calculations of non-LOCA events demonstrate margins to the primary and secondary pressure acceptance criteria. Other quantitative acceptance criteria, such as minimum critical heat flux ratio and radiological dose limits applicable for the non-LOCA events, are evaluated in downstream subchannel and accident radiological analyses, which are documented in separate reports. © Copyright 2022 by NuScale Power, LLC 1

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Executive Summary The purpose of this report is to present the NuScale evaluation model (EM) used to evaluate a NuScale Power Module (NPM) short-term system transient response to non-loss-of-coolant accident (LOCA) events. This report summarizes the NuScale plant designs, non-LOCA initiating events, classification of the events and acceptance criteria. The scope of the non-LOCA system transient analysis and interfaces to other safety analysis methodologies is described. Development of the non-LOCA phenomena identification and ranking table is described; the high-ranked phenomena and how they are assessed are summarized. The NRELAP5 code is the system thermal-hydraulic code used for non-LOCA system transient analysis. Applicability of NRELAP5 for non-LOCA system transient analysis is assessed. Parameters considered in the system transient analyses to specify appropriately conservative input in application of the EM are discussed. Representative transient calculations for different types of non-LOCA events are presented. These representative calculations demonstrate the application of the method only and are not intended for NRC approval of a NuScale plant design. An NPM is a small, light water integral pressurized water reactor (PWR) consisting of a nuclear core, two helical coil steam generators (SGs), and a pressurizer, all contained within a single reactor vessel. The reactor vessel is located within a small, compact steel containment vessel. An NPM is designed to operate efficiently at full power conditions using natural circulation as the means of providing core coolant flow, eliminating the need for reactor coolant pumps. The helical coil SGs, which produce superheated steam, are located in the annular space between the RCS hot leg riser and the reactor vessel inside diameter wall. The relative locations of the thermal centers in the core and the SGs promote buoyancy driven natural circulation flow. Power conversion occurs via a secondary system that includes the steam turbine-generator, the main condenser, and the plant components necessary to provide feedwater. Each NPM has dedicated chemical and volume control system (CVCS), emergency core cooling system (ECCS), and decay heat removal system (DHRS). The CVCS is used to regulate the primary side inventory via makeup and letdown to maintain pressurizer level and boron concentration during normal operation. Pressurizer heaters and spray control primary side pressure. The DHRS is a normally isolated, closed-loop, two-phase natural circulation cooling system; two trains of decay heat removal equipment are provided, one connected to each SG secondary side loop. These major features are common to NPM designs. A NuScale power plant consists of one or more NPMs, each partially immersed in its own bay of the common reactor pool. The reactor pool serves as the ultimate heat sink (UHS) and is located in a Seismic Category I building designed to withstand postulated adverse natural conditions and aircraft impact. The NuScale design instrumentation and control architecture includes the safety-related module protection system, and nonsafety-related module and plant control systems. These major features are common to NuScale power plant designs. The non-LOCA system transient evaluation model for analysis of an NPM system transient response to non-LOCA events is developed following a graded approach to the guidance provided in Regulatory Guide (RG) 1.203 for the evaluation model development and assessment process (EMDAP). The EMDAP as defined in RG 1.203 provides a structured process to establish the adequacy of a methodology for evaluating complex events that are postulated to occur in nuclear power plant systems. Six basic principles are identified in RG 1.203 as important in the process of developing and assessing an EM. Four of the principles (using 20 steps as © Copyright 2022 by NuScale Power, LLC 2

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 identified in the EMDAP process) are addressed in this report. The remaining principles related to establishing an appropriate quality assurance program and providing comprehensive, accurate, up-to-date documentation are addressed outside of this report as part of NuScale Power, LLC Quality Assurance Program Description, MN-122626 (Reference 3). As part of defining the requirements of the non-LOCA transient system analysis methodology, the events to which the methodology applies and the specific event acceptance criteria applicable to the events are identified. The NPM designs were evaluated to assure that a sufficiently broad spectrum of transients, accidents, and initiating events have been included in the scope of design basis analyses presented in Final Safety Analysis Report (FSAR) Chapter 15. An NPM is a natural circulation integral PWR. Many of the events analyzed for operating plants and in recent design certification applications are applicable to the NuScale designs. NuScale-specific events reflect unique aspects of the NuScale designs such as the DHRS and normal operation of the containment at vacuum conditions. The design-basis events for which non-LOCA system transient analysis are performed are categorized into one of 5 categories:

1. Increase in heat removal from the RCS

2. Decrease in heat removal by the secondary system

3. Reactivity and power distribution anomalies

4. Increase in reactor coolant inventory

5. Decrease in reactor coolant inventory Each event is classified as an anticipated operational occurrence (AOO), infrequent event (IE), or accident. The specific event acceptance criteria are derived from the regulatory requirements and guidance. The non-LOCA transient system analysis is part of several stages of analysis performed to confirm a plant design meets applicable acceptance criteria for a limiting set of AOOs, IEs, and accidents. The interfaces of the non-LOCA transient system analysis with downstream subchannel analysis and accident radiological analysis methodologies are identified. The subchannel analysis codes and methods, and the accident radiological source term and dose analyses are covered in separate methodologies and assessments. For non-LOCA initiating events that actuate the DHRS, the EM is applicable for the short-term transient progression; during this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained.

As part of developing the non-LOCA evaluation model, a phenomena identification and ranking table (PIRT) was developed. The non-LOCA events cover several different event types based on the main effect on the RCS. A comprehensive, integrated PIRT was performed for the range of non-LOCA event types and phases of the event progression. The PIRT panel considered an NPM design to identify systems, components, and subcomponents of the design for which phenomena were assessed. Phenomena were identified and ranked considering their level of importance relative to identified figures of merit for the different non-LOCA event types and phases of the transient progression; a knowledge ranking was established for each of the phenomena. NRELAP5 is NuScales system thermal-hydraulics code used to simulate an NPM system response during both the non-LOCA and LOCA short-term transient event progression. The © Copyright 2022 by NuScale Power, LLC 3

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 NRELAP5 code was derived from the Idaho National Laboratory (INL) RELAP5-3D© computer code. RELAP5-3D©, version 4.1.3 was used as the baseline development platform for the NRELAP5 code. RELAP5-3D© was procured and commercial grade dedication was performed by NuScale. Subsequently, features were added and changes made to address unique aspects of an NPM design and licensing methodology. NRELAP5 is a non-homogenous, non-equilibrium two-fluid thermal hydraulic systems analysis code capable of performing non-LOCA system transient analyses for an NPM. The NRELAP5 code has a heat conduction and heat transfer package that is similar in capability to other thermal-hydraulic codes in its class (such as TRAC, RETRAN or TRACE). It includes the trips and logic control systems that enable simulation of the plant protection and control system logic for analysis of a non-LOCA event in an NPM. The NRELAP5 code is described in the NuScale LOCA Evaluation Model licensing topical report. The NRELAP5 code has been assessed against several separate effects and integral effects tests as part of the code development and development of the NuScale LOCA evaluation model to demonstrate the capability to simulate an NPM response to LOCA events. Phenomena identified as high-ranked for the non-LOCA transients were evaluated with respect to the high-ranked phenomena identified and assessed as part of the NuScale LOCA evaluation model development. Additional validation of NRELAP5 against separate effects testing, integral effects testing, and code to code benchmarking, were performed as necessary to justify applicability of the NRELAP5 code for non-LOCA system thermal-hydraulic analysis. High-ranked phenomena for non-LOCA events that were not assessed as part of the NuScale LOCA evaluation model development were therefore addressed in different ways:

1. additional NRELAP5 code assessment performed against separate effects or integral effects test data

2. code-to-code benchmark performed between NRELAP5 and independent system thermal-hydraulics code

3. phenomena addressed as part of the downstream subchannel analysis

4. phenomena addressed by specifying appropriately conservative input to the system transient analysis In particular, separate and integral effects testing were performed at the NIST-1 and NIST-2 facilities to support applicability of the NRELAP5 code for non-LOCA system transient analysis.

Separate effects testing of the DHRS was performed. Integral effects testing of an NPM response to a decrease in secondary side heat transfer, and integral effects testing of DHRS operation were performed. A code-to-code benchmark was performed to compare the NRELAP5 and RETRAN-3D responses to a range of reactivity insertion conditions in an NPM. Computational fluid dynamics was used to independently assess the heat transfer correlation used in NRELAP5 for the helical SGs of the NuScale designs. The general non-LOCA transient analysis process is described in this report. The general methodology for conservatively biasing initial and boundary conditions for event analysis is presented. Each initiating event is then considered to identify the acceptance criteria that may be challenged during the event. For each non-LOCA event, a description of the event is provided including biases and conservatisms applied, sensitivity studies performed, single active failures and loss of power scenarios that challenge the event acceptance criteria. For each transient event, the acceptance criteria where margin to the limit may be challenged are identified. For © Copyright 2022 by NuScale Power, LLC 4

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 these acceptance criteria, sensitivity calculations are performed as necessary to confirm that appropriately conservative inputs are specified and to determine conditions that result in minimum margin. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed are sufficient to demonstrate that margin to the acceptance criterion is maintained. For non-LOCA initiating events that actuate the DHRS, the EM is applicable for the short-term transient progression; during this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained. For selected non-LOCA events, representative system transient results are provided to demonstrate the application of the evaluation model for an NPM. System transient calculations are executed for sufficient duration to demonstrate that the initiating event is mitigated and stable cooling is established. Results of representative calculations show that the maximum primary and secondary pressure acceptance criteria are not significantly challenged in an NPM design. Margin to other quantitative acceptance criteria for minimum critical heat flux ratio, fuel centerline temperature, and radiological dose limits applicable for the non-LOCA events are demonstrated as part of downstream subchannel or accident radiological analyses that are described in separate reports. In addition, long-term cooling analysis methodology is presented in a separate report. NuScale requests U.S. Nuclear Regulatory Commission (NRC) approval to use the EM described in this report for analyses of NPM design basis non-LOCA events that require system transient analysis. The specific scope of the non-LOCA events for which the EM applies is delineated in Section 1.2. NuScale is not seeking NRC approval of the representative calculations that are described in this report. © Copyright 2022 by NuScale Power, LLC 5

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 1.0 Introduction 1.1 Purpose The purpose of this report is to present the NuScale evaluation model (EM) used to evaluate a NuScale Power Module (NPM) system transient response to non-loss-of-coolant accident (non-LOCA) events with the NRELAP5 code. This report summarizes the NuScale plant designs and identifies the potential non-LOCA initiating events for an NPM analyzed by this EM. The classification of these non-LOCA events and relevant acceptance criteria that are prescribed in the NRC standard review plan (SRP) and the NuScale design specific review standard (DSRS) are discussed in this report. The purpose of the non-LOCA evaluation model is to model an NPM response to a non-LOCA design basis event. The non-LOCA system transient evaluation model is developed following a graded approach to the guidance provided in Regulatory Guide (RG) 1.203 (Reference 1). The non-LOCA phenomena identification and ranking table (PIRT) is described, including a summary of the high-ranked phenomena and how they are assessed. The applicability of NRELAP5 for non-LOCA system transient analysis is assessed. The scope of the non-LOCA system transient analysis is described in this report, as well as interfaces to other safety analysis methodologies. This report describes the selection of appropriately conservative input when applying this EM to perform non-LOCA system transient analyses. Representative transient calculations from application of the EM for the range of non-LOCA events are presented. 1.2 Scope NuScale requests U.S. Nuclear Regulatory Commission (NRC) approval to use the EM described in this report for analyses of NPM design basis non-LOCA events that require system transient analysis. Representative analysis results are provided in Section 8.0 of this report to illustrate results from application of the EM. These representative cases are not necessarily based on final NuScale NPM design inputs, and NRC approval of the representative results is not requested. The scope of this report includes the applicability and acceptability of this methodology to evaluate the primary and secondary system pressure acceptance criteria found in Chapter 15 of the NuScale DSRS and the SRP. This report also describes how the non-LOCA evaluation model interfaces with other analyses that evaluate acceptance criteria that are not evaluated by the non-LOCA evaluation model. The scope of the NuScale non-LOCA system transient analysis EM is summarized below: The non-LOCA evaluation model uses the NRELAP5 code to perform system transient analysis of the NPM design basis events listed in Table 4-1. The general and event-specific analysis methodologies of the EM are presented in Section 7.0. Sensitivity studies justifying the selected biasing direction are presented in Section 7.2 as part of the event-specific analysis methodology of the EM. The NRELAP5 code is described in the LOCA Evaluation Model licensing topical report (Reference 2). © Copyright 2022 by NuScale Power, LLC 6

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The non-LOCA evaluation model is applicable to a nuclear power plant that follows the general description of the NuScale plant designs in Section 3.0. The applicability of the EM is based on the non-LOCA phenomena identification and ranking table and assessment of the high-ranked phenomena that are treated as part of the system transient analysis. The non-LOCA evaluation model does not address the evaluation of specified acceptable fuel design limits (SAFDLs), which are evaluated in a downstream subchannel analysis. The subchannel analysis codes and methods are covered in separate methodologies and assessments (Reference 6, supplemented by Reference 28). However, the interface of the non-LOCA system transient analysis with the downstream subchannel analysis is part of the non-LOCA evaluation model. The non-LOCA evaluation model does not address the evaluation of the accident radiological source term and dose. The accident radiological source term and dose analyses are covered in separate methodologies and assessments (Reference 8). However, the interface of the non-LOCA system transient analysis with downstream radiological analysis is part of the non-LOCA evaluation model. The EM is applicable for the short-term non-LOCA transient progression; the non-LOCA transient analysis short-term duration and analysis process are discussed further in Section 4.2 and Section 4.3. During this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained. The reactivity control and extended passive cooling analysis methodology in the long-term, including events that transition from DHRS cooling to ECCS cooling, is addressed in separate methodologies and assessments (Reference 26). The plant design overview description in Section 3.0, plant model description in Section 6.0, input and biasing discussion in Section 7.0, and the example calculations in Section 8.0 do not include the design feature of riser holes (upper or lower). Riser holes are included in the NPM designs to mitigate potential boron dilution impacts of long term DHRS cooling and riser uncovery. The riser holes do not impact the short term DHRS cooldown during the non-LOCA phase. The PIRT and test assessments in Section 5.0 did not incorporate the riser holes; however, the PIRT and assessments documented in Section 5.0 remain valid, since the design feature of the riser holes results in negligible differences in the prediction of RCS parameters.The effect of the riser holes during extended passive cooling is addressed in separate methodologies and assessments (Reference 26). Control rod ejection accident analysis is addressed by a separate methodology (Reference 21) and is not part of the non-LOCA evaluation model. Loss of coolant accident analysis, including analysis of an inadvertent opening of one or more valves on the RPV, is addressed by a separate methodology (Reference 2) and is not part of the non-LOCA evaluation model. Analysis of the peak containment pressure and temperature response is addressed by a separate methodology (Reference 2) and is not part of the non-LOCA evaluation model. © Copyright 2022 by NuScale Power, LLC 7

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Evaluation of a return to power assuming the worst case stuck rod, if applicable to a NuScale design, is addressed in the final safety analysis report (FSAR) and is not part of the non-LOCA evaluation model. © Copyright 2022 by NuScale Power, LLC 8

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 1.3 Abbreviations Table 1-1 Abbreviations Term Definition AC alternating current ANS American Nuclear Society AOO anticipated operational occurrence ASME American Society of Mechanical Engineers BOC beginning of cycle CFR Code of Federal Regulations CHF critical heat flux CHFR critical heat flux ratio CNV containment vessel CPV cooling pool vessel CRA control rod assembly CVC chemical and volume control CVCS chemical and volume control system DACS data acquisition and control system DC direct current DCA Design Certification Application DFWT decrease in feedwater temperature DHRS decay heat removal system DNB departure from nucleate boiling DSRS Design Specific Review Standard DTC Doppler temperature coefficient ECCS emergency core cooling system EDAS augmented DC power system EDNS normal DC power system EDSS highly reliable DC power system EHVS high voltage AC electrical distribution system ELVS low voltage AC electrical distribution system EM evaluation model EMDAP evaluation model development and assessment process EMVS medium voltage AC electrical distribution system EOC end of cycle ESFAS engineered safety features actuation system FOM figure of merit FSAR Final Safety Analysis Report FWIV feedwater isolation valve FWRV feedwater regulating valve GDC General Design Criteria HTP heat transfer plate HX heat exchanger IAB inadvertent actuation block ID inside diameter IE infrequent event IET integral effects test INL Idaho National Laboratory © Copyright 2022 by NuScale Power, LLC 9

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 1-1 Abbreviations (Continued) Term Definition KAIST Korea Advanced Institute of Science and Technology L/D length to diameter ratio LOCA loss-of-coolant accident LOFW loss of feedwater LOP loss of power MASLWR multi-application small light water reactor MCHFR minimum critical heat flux ratio MCS module control system MPS module protection system MSIBV main steam isolation bypass valve MSIV main steam isolation valve MSS main steam system MTC moderator temperature coefficient NIST NuScale Integral System Test Facility NPM NuScale Power Module NRC Nuclear Regulatory Commission NRELAP5 NuScale version of RELAP5-3D© OD outside diameter OSU Oregon State University P/L pressure/level PCS plant control system P/D pitch to diameter ratio PIRT phenomena identification and ranking table PWR pressurized water reactor PZR pressurizer QAP Quality Assurance Program RCS reactor coolant system RG Regulatory Guide RPV reactor pressure vessel RRV reactor recirculation valve RSV reactor safety valve RTP rated thermal power RTS reactor trip system RVV reactor vent valve SAF single active failure SAFDL specified acceptable fuel design limit SDAA Standard Design Approval Application SET separate effects test SG steam generator SGTF steam generator tube failure SMR small modular reactor SRP standard review plan SSC structures, systems, and components TMDPJUN time dependent junction TMDPVOL time dependent volume © Copyright 2022 by NuScale Power, LLC 10

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 1-1 Abbreviations (Continued) Term Definition UCRW uncontrolled rod withdrawal UCRWS uncontrolled rod withdrawal from subcritical (or low power) UHS ultimate heat sink VAC volts alternating current © Copyright 2022 by NuScale Power, LLC 11

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 1-2 Definitions Term Definition Analytical limit Limit of a measured or calculated variable established by the safety analysis to ensure that a safety limit is not exceeded. Anticipated operational Conditions of normal operation that are expected to occur one or more occurrences (AOOs) times during the life of the nuclear power unit. Design basis accidents A postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures and components necessary to ensure public health and safety. Design basis events Postulated events used in the design to establish the acceptable performance requirements for the structures, systems and components. Design bases Information that identifies the specific functions to be performed by a system, structure or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted state of the art practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effect of a postulated accident for which a structure, system or component must meet its functional goals. Excellent agreement One of the acceptance criteria defined in RG 1.203. Excellent agreement applies when the code exhibits no deficiencies in modeling a given behavior. Major and minor phenomena and trends are correctly predicted. The calculated results are judged to agree closely with the data. The calculation, with few exceptions, lies within the specified or inferred uncertainty bands of the data. The code may be used with confidence in similar applications. Infrequent events (IEs) Events that are not classified as AOOs or as accidents, and are not expected to occur during the design life of the plant. For IEs, the acceptance criteria are specified such that some fuel damage may occur but the radiological acceptance criteria are stricter than those imposed for accidents. These include events that may be historically considered as AOOs for operating plants, but due to aspects of the NPM designs, the events are not expected to occur during the design life of the plant. Insufficient agreement One of the acceptance criteria defined in RG 1.203. Insufficient agreement applies when the code exhibits major deficiencies. The code provides an unacceptable prediction of the test data because major trends are not predicted correctly. Most calculated values lie outside the specified or inferred uncertainty bands of the data. Loss-of-coolant accident Those postulated accidents that result in a loss of reactor coolant at a rate in excess of the capability of the reactor makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. © Copyright 2022 by NuScale Power, LLC 12

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 1-2 Definitions (Continued) Term Definition Minimal agreement One of the acceptance criteria defined in RG 1.203. Minimal agreement applies when the code exhibits significant deficiencies. Overall, the code provides a prediction that is only conditionally acceptable. Some major trends or phenomena are not predicted correctly, and some calculated values lie considerably outside the specified or inferred uncertainty bands of the data. Incorrect conclusions about trends and phenomena may be reached if the code were used in similar applications, and an appropriate warning needs to be issued to users. Selected code models or facility model noding need to be reviewed, modified and assessed before the code can be used with confidence in similar applications. Non-LOCA transient Reactor coolant system transients described in the standard review plan Sections 15.1, 15.2, 15.4, and 15.5, and other comparable transients that may be unique to the NuScale systems. Other sections in the standard review plan are specific to events with reactor coolant pumps, LOCA, radiological analysis, anticipated transient without scram, or boiling water reactors, and are outside of scope. Postulated accidents A postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structures and components necessary to ensure public health and safety. Reasonable agreement One of the acceptance criteria defined in RG 1.203. Reasonable agreement applies when the code exhibits minor deficiencies. Overall, the code provides an acceptable prediction. All major trends and phenomena are correctly predicted. Differences between calculation and data are greater than deemed necessary for excellent agreement. The calculation frequently lies outside but near the specified or inferred uncertainty bands of the data. However, the correct conclusions about trends and phenomena would be reached if the code were used in similar applications. Safety-related structures, Those structures, systems and components that are relied upon to system and components remain functional during and following design-basis events to assure: (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1). © Copyright 2022 by NuScale Power, LLC 13

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 2.0 Background This topical report provides a description of the NuScale non-LOCA system transient analysis EM. The non-LOCA system transient analysis EM is developed following a graded application of the guidelines in the evaluation model development and assessment process (EMDAP) of RG 1.203. Six basic principles are identified in RG 1.203 as important in the process of developing and assessing an EM. Four of the principles (using 20 steps as identified in the EMDAP process) are addressed in this report. The remaining principles related to establishing an appropriate quality assurance program and providing comprehensive, accurate, up-to-date documentation are addressed outside of this report as part of NuScale Power, LLC Quality Assurance Program Description, MN-122626 (Reference 3). This EM utilizes the NRELAP5 code that was developed from the Idaho National Laboratory (INL) RELAP5-3D© computer code. The NRELAP5 code is described in Reference 2. Applicability of the NRELAP5 code for application in non-LOCA system transient analysis is discussed in this report. 2.1 Non-LOCA Evaluation Model Roadmap Analyses are performed to demonstrate that a nuclear power plant can meet applicable NRC regulatory acceptance criteria for a limiting set of AOOs, IEs, and accidents. The EMDAP as defined in RG 1.203 provides a structured process to establish the adequacy of a methodology for evaluating complex events that are postulated to occur in nuclear power plant systems. The EM described here is developed for simulating an NPM system transient response to non-LOCA events. NRELAP5 is NuScales system thermal-hydraulics code used to simulate an NPM system response during both the non-LOCA and LOCA short-term transient event progression. The NuScale LOCA EM (Reference 2) was developed following the EMDAP guidelines of RG 1.203. As described in Section 5.0, phenomena identified as high-ranked for the non-LOCA transients were evaluated with respect to the high-ranked phenomena identified as part of the NuScale LOCA evaluation model development. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations (Section 7.0), a graded approach to the EMDAP is applied for development of the non-LOCA system transient EM. Figure 2-1 shows various elements of EMDAP as defined in RG 1.203. The elements of the EMDAP and sections of this report that relate to the elements and steps of the EMDAP are summarized in Table 2-1. © Copyright 2022 by NuScale Power, LLC 14

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 2-1 Evaluation model development and assessment process © Copyright 2022 by NuScale Power, LLC 15

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 2-1 Evaluation model development and assessment process steps and associated application in the non-LOCA evaluation model EMDAP Description EM Section Step Element 1, Establish Requirements for Evaluation Model Capability 1 Specify analysis The purpose of the non-LOCA system transient analysis methodology purpose, transient is described in Section 1.1. Section 2.0 briefly describes the class and power plant background of the process followed to develop the non-LOCA system class. transient analysis methodology and the principal software used. Section 3.0 provides an overview of an NPM and a description of the plant operation. This overview includes the safety systems, the system logic, and operational phases that could occur in a NuScale SMR design. The high level regulatory requirements that the methodology is designed to comply with are described in Section 2.2. In Section 4.1 the non-LOCA initiating events and the classification of the events for an NPM are discussed. The acceptance criteria for the events are identified in Section 4.2. As identified in Section 4.2, margin to some of these acceptance criteria are demonstrated based on the results of the non-LOCA system transient analysis; other acceptance criteria are met as part of downstream analyses such as subchannel and radiological analyses. Downstream analyses are outside the scope of this topical report as discussed in Section 1.2. The non-LOCA transient analysis process, including interfaces with other safety analysis methodologies, is described in Section 4.3. 2 Specify figures of merit Section 5.1 discusses the FOMs that are used for the development of (FOMs). the NPM non-LOCA PIRT. 3 Identify systems, Systems, components, phases and processes are identified as a part components, phases, of the non-LOCA PIRT discussed in Section 5.1. geometries, fields, and processes that should be modeled. 4 Identify and rank Section 5.1 describes the NPM non-LOCA PIRT. phenomena and processes. Element 2, Develop Assessment Base 5 Specify objectives for Section 5.2 describes the high ranked phenomena identified from the assessment base. PIRT process and how the phenomena are addressed by NRELAP5 assessment or other approach. Many of the high ranked phenomena were assessed against experimental data as part of the LOCA evaluation model development (Reference 2); additional assessments were identified as described in Section 5.2. © Copyright 2022 by NuScale Power, LLC 16

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 2-1 Evaluation model development and assessment process steps and associated application in the non-LOCA evaluation model (Continued) EMDAP Description EM Section Step 6 Perform scaling A scaling analysis of the LOCA and emergency core cooling system analysis and identify (ECCS) has been performed for the NPM centered on the NuScale similarity criteria. Integral Systems Test-1 (NIST-1) facility. The results of the scaling analysis are discussed in Reference 2. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. 7 Identify existing data Reference 2 and Section 5.3 of this report provide the results of the and perform integral NRELAP5 validation against the SETs and IETs. effects test (IETs) and separate effects tests (SETs) to complete database. 8 Evaluate effects of IET In Reference 2, a bottom-up assessment of NRELAP5 closure models distortions and SET and correlations is presented; this assessment addresses the fidelity of scaleup capability. the models and correlations to the appropriate fundamental or SET data. In Reference 2, a top-down assessment of the NRELAP5 governing equations and numerics is presented. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. 9 Determine Reference 2 and Section 5.3 of this report cover experimental experimental uncertainties for NRELAP5 assessments against the SETs and IETs. uncertainties. Element 3, Develop Evaluation Model 10 Establish EM The NRELAP5 development plan includes programming standards and development plan. procedures, quality assurance procedures, and configuration control, which are summarized in Reference 2. 11 Establish EM Reference 2 provides a summary of NRELAP5 models and structure. correlations. The non-LOCA transient analysis process, including interfaces with other safety analysis methodologies, is described in Section 4.3. For non-LOCA system transient analysis, the typical plant model is described in Section 6.0. The non-LOCA analysis methodology is described in Section 7.0. 12 Develop or incorporate Reference 2 provides a summary of NRELAP5 models and closure models. correlations. A full description of the closure models and the associated equations used in the non-LOCA evaluation model is provided in the NRELAP5 theory and users manuals. © Copyright 2022 by NuScale Power, LLC 17

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 2-1 Evaluation model development and assessment process steps and associated application in the non-LOCA evaluation model (Continued) EMDAP Description EM Section Step Element 4, Assess Evaluation Model Adequacy Closure Relations (Bottom-up) 13 Determine model Reference 2 includes a bottom-up assessment of important NRELAP5 pedigree and models and correlations essential to simulate high-ranked PIRT applicability to phenomena for LOCA events, including discussion of model pedigree simulate physical and applicability. Considering the overlap in high-ranked phenomena processes. and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. 14 Prepare input and Reference 2 and Section 5.3 summarize the results of comparison of perform calculations to NRELAP5 against the selected SETs and IETs including evaluation of assess model fidelity code fidelity and accuracy. and accuracy. 15 Assess scalability of Reference 2 includes discussion on scalability of dominant NRELAP5 models. models and correlations that are essential to simulate high-ranked PIRT phenomena for LOCA events. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. Element 4, Assess Evaluation Model Adequacy Integrated EM (Top-down) 16 Determine capability of NRELAP5 field equations and the numeric solution scheme are field equations and discussed in Reference 2 and evaluated for their applicability to NPM numeric solutions to LOCA. Considering the overlap in high-ranked phenomena and represent processes conservatism applied to input and boundary conditions in the and phenomena. non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. 17 Determine applicability The applicability of the EM to simulate the NPM systems and of EM to simulate components is demonstrated by assessment of NRELAP5 against system components. NuScale design-specific SETs and IETs. 18 Prepare input and Reference 2 and Section 5.3 summarize the results of an assessment perform calculations to of NRELAP5 against IET data. assess system interactions and global capability. 19 Assess scalability of Reference 2 provides an evaluation of scaling distortions between the integrated calculations NIST-1 LOCA IET data and the NPM design. The scalability of the EM and data for to represent NPM LOCA phenomena and processes is presented distortions. therein. Considering the overlap in high-ranked phenomena and conservatism applied to input and boundary conditions in the non-LOCA plant transient calculations, these assessments are considered adequate for the non-LOCA system transient EM. 20 Determine EM biases For the non-LOCA system transient analyses, suitably conservative and uncertainties. input is specified in the plant calculations as described in Section 7.0, considering the effects on the appropriate acceptance criteria. © Copyright 2022 by NuScale Power, LLC 18

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 2.2 Regulatory Requirements The following General Design Criteria (GDC) of 10 CFR Part 50 Appendix A (Reference 4) are relevant to the non-LOCA transient analyses: GDC 5, as it relates to demonstrating that sharing of structures, systems, and components (SSC) does not significantly impact the ability of the SSC to perform their safety function. GDC 10, as it relates to demonstrating that SAFDLs are not exceeded during AOOs. GDC 15, as it relates to demonstrating that the reactor coolant system pressure boundary is not breached during AOOs. GDC 17, as it relates to providing electric power systems to permit functioning of SSC to assure that SAFDLs and design conditions of the reactor coolant pressure boundary are not exceeded as a result of AOOs, and that the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents. The NuScale designs support an exemption from GDC 17 and therefore GDC 17 is not relevant to this methodology. GDC 20, as it relates to demonstrating that the automatic operation of systems by the reactor protection system ensures that a plant does not exceed SAFDLs during AOOs. GDC 25, as it relates to demonstrating that the protection system design assures that SAFDLs are not exceeded for any single malfunction of the reactivity control system, such as accidental withdrawal (not ejection or dropout) of control rods. GDC 26, as it relates to demonstrating that the control rods are capable of reliably controlling reactivity changes to assure that SAFDLs are not exceeded during AOOs, with appropriate margin for malfunctions such as stuck rods. GDC 27, as it relates to demonstrating that the reactivity control systems are capable of reliably controlling reactivity changes to assure capability to cool the core is maintained under postulated accident conditions, with appropriate margin for stuck rods. Some NuScale designs support an exemption from GDC 27, and instead implement a NuScale-specific Principal Design Criterion (PDC) 27, as documented in the FSAR for those designs. As relevant to this methodology, NuScale PDC 27 is equivalent to GDC 27 (i.e., differences are not within the scope of this evaluation methodology). GDC 28, as it relates to demonstrating that the effects of postulated reactivity accidents can neither result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor impair significantly the capability to cool the core. GDC 31, as it relates to demonstrating that the probability of a rapidly propagating fracture of the reactor coolant pressure boundary is minimized when the pressure boundary is stressed under postulated accident conditions. GDC 34, as it relates to demonstrating that residual heat is removed from the reactor core at a rate such that SAFDLs and the design conditions of the reactor coolant pressure boundary are not exceeded, assuming a single failure and considering © Copyright 2022 by NuScale Power, LLC 19

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 offsite power availability. The NuScale designs support an exemption from GDC 34, and instead implement a NuScale-specific Principal Design Criterion (PDC) 34, as documented in the FSAR for the designs. As relevant to this methodology, NuScale PDC 34 is equivalent to GDC 34 (i.e., differences are not within the scope of this evaluation methodology). Regulatory guidance documents relevant to the non-LOCA transient system analysis EM development include: Regulatory Guide 1.203, Transient and Accident Analysis Methods, December 2005. NuScale Design Specific Review Standard Sections:

             -    15.0, Revision 0, Introduction - Transient and Accident Analyses June 2016.
             -    15.1.1-15.1.4, Revision 0, Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of the Turbine Bypass System or Inadvertent Operation of the Decay Heat Removal System, June 2016.
             -    15.1.5, Revision 0, Steam System Piping Failures Inside and Outside of Containment, June 2016.
             -    15.1.6, Revision 0, Loss of Containment Vacuum, June 2016.
             -    15.2.1-15.2.5, Revision 0, Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed), June 2016.
             -    15.2.6, Revision 0, Loss of Nonemergency AC Power to the Station Auxiliaries, June 2016.
             -    15.2.7, Revision 0, Loss of Normal Feedwater Flow, June 2016.
             -    15.2.8, Revision 0, Feedwater System Pipe Breaks Inside and Outside Containment (PWR), June 2016.
             -    15.5.1-15.5.2, Revision 0, Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory, June 2016.

NUREG-0800 Sections:

             -    15.0.2, Revision 0, Review of Transient and Accident Analysis Methods, March 2007.
             -    15.4.1, Revision 3, Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition, March 2007.
             -    15.4.2, Revision 3, Uncontrolled Control Rod Assembly Withdrawal at Power, March 2007.
             -    15.4.3, Revision 3, Control Rod Misoperation (System Malfunction or Operator Error), March 2007.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

             -    15.4.4-15.4.5, Revision 2, Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate, March 2007.
             -    15.4.6, Revision 2, Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR), March 2007.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 3.0 Plant Design Overview An overview is provided of the major features common to NPMs and NuScale Power Plants. The non-LOCA EM is generally applicable to NPM and NuScale Power Plant designs that are consistent with the major features and plant operation described in the following sections. Applicability of the non-LOCA EM to a specific NPM design is based on assessment of the design as described in Section 5.0. 3.1 Description of NuScale Plant An NPM is a small, light water cooled, pressurized water reactor (PWR) consisting of a nuclear core, two helical coil SGs, and a pressurizer, all contained within a single containment vessel (CNV) (Figure 3-1). Power conversion occurs via a standard secondary system that includes the steam turbine-generator, the main condenser, and the plant components necessary to provide feedwater. Each NPM is covered by a reinforced concrete biological shield and enclosed in a Reactor Building, and has a dedicated chemical and volume control system (CVCS), ECCS, and DHRS. An NPM is designed to operate efficiently at full power conditions using natural circulation as the means of providing core coolant flow, eliminating the need for reactor coolant pumps. Unique features of an NPM include: a reduced core size relative to operating PWRs, natural circulation reactor coolant flow (i.e., no reactor coolant pumps), integrated SGs and pressurizer inside the reactor pressure vessel (RPV) (i.e., there is no piping connecting the SGs or pressurizer with the reactor), simplified passive safety systems that do not rely on ECCS pumps, accumulators, tanks, or connected piping, a high-pressure steel containment, and containment partially immersed in a water-filled pool providing an effective passive heat sink for emergency cooling and decay heat removal. NuScale has achieved a substantial improvement in safety over existing plants through simplicity of design, reliance on passive safety systems, and small fuel inventory. A NuScale Power Plant consists of one or more NPMs, each in its own bay of the common reactor pool. The pool is designed for leakage control. Each bay has a reinforced concrete cover that serves as a biological shield. The cover also serves to prevent deposition of foreign materials onto an NPM. The reactor pool is located in a Seismic Category I building designed to withstand postulated adverse natural conditions and aircraft impact. © Copyright 2022 by NuScale Power, LLC 22

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 3-1 NuScale Power Module schematic © Copyright 2022 by NuScale Power, LLC 23

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 3.2 Plant Operation During nominal full power operation, the control rods are retracted to within their insertion limits and borated water is used as the primary coolant, which is driven by natural circulation. The CVCS is used to regulate the primary side inventory via manually operated makeup and either automatic or manual letdown, depending on the design, to maintain pressurizer level and boron concentration to maintain criticality. Pressurizer heaters and spray control primary side pressure. The helical coil SGs transfer the heat from the primary side to the feedwater. The DHRS heat exchangers are isolated during normal operation. The containment is evacuated to provide an insulated barrier between the reactor pressure vessel and containment. Each NPM is partially immersed in the reactor pool within the Reactor Building, which serves as the ultimate heat sink (UHS) and is open to atmospheric pressure. The pool cooling equipment is designed to maintain an average bulk pool temperature such that plant personnel can work in the Reactor Building. The NuScale instrumentation and control architecture primarily consists of the following systems: module control system (MCS) plant control system (PCS) module protection system (MPS) plant protection system The MCS and PCS provide control and monitoring of the non-safety nuclear steam supply system (e.g., steam bypass to condensers, pressurizer heaters and sprays, and feedwater control), balance of plant systems (e.g., turbine control); rod control and position indication, and plant-wide, non-safety control and indication. The MPS is composed primarily of the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS). The MPS protection functions are limited to automated safety responses to off-normal conditions. The MPS functional response to an initiating event is a reactor trip, isolation (as necessary) of feedwater, MSS, CVCS (including demineralized water system isolation to mitigate boron dilution), and containment, followed by an integrated safety actuation of one or more of the passive safety systems (DHRS and ECCS). The RTS consists of four independent separation groups with independent measurement channels to monitor plant parameters that can generate a reactor trip. Each measurement channel trips when the parameter exceeds a predetermined setpoint. The RTS coincident logic is designed so that no single failure can prevent a reactor trip when required, and no failure in a single measurement channel can generate an unnecessary reactor trip. The ESFAS consists of four independent separation groups with independent measurement channels to monitor plant parameters that can activate the operation of the engineered safety features. Each measurement channel trips when the parameter © Copyright 2022 by NuScale Power, LLC 24

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 exceeds a predetermined setpoint. The ESFAS coincident logic is designed so that no single failure can prevent a safeguards actuation when required, and no failure in a single measurement channel can generate an unnecessary safeguards actuation. Transients requiring decay heat removal are addressed by the DHRS, which provides cooling through one or both of the SGs. For a steam generator tube failure (SGTF), main steam line break, and feedwater line break, the affected SG is isolated and the DHRS provides cooling through the intact SG (depending on the break location DHRS may be operational in both SGs). Manual operation of the nonsafety-related CVCS can also be used to offset decreases in RCS inventory. If the CVCS is inadequate to address the inventory decrease, containment isolation occurs and the DHRS is actuated. If RCS inventory loss to containment persists, ECCS is actuated. Module-specific systems and functions that operate to mitigate the effects of postulated non-LOCA events (and credited in the safety analysis) include the ECCS, DHRS, CVCS and demineralized water system isolation, MPS, RTS, containment isolation and PZR heater isolation. The only safety system shared between modules is the UHS. The non-LOCA safety analyses consider ranges of UHS conditions and heat transfer such that non-LOCA analysis of a single module bounds possible UHS interactions between modules. 3.3 Decay Heat Removal System The DHRS is a closed-loop, two-phase natural circulation cooling system. Two trains of decay heat removal equipment are provided, one attached to each SG loop. Each train is capable of removing 100 percent of the decay heat load and cooling the RCS. Each train has a passive condenser immersed in the reactor pool. Upon receipt of an actuation signal, the main steam isolation valves (MSIVs) and the feedwater isolation valves close, and the decay heat removal actuation valves open, allowing heat removal via the SGs. The decay heat removal actuation valves would open upon the loss of power, thus enabling reliable long term cooling. For successful operation, liquid water enters the SG through the feedwater line and is boiled by heat from the RCS. The vapor exits the SG through the steam line and is directed to the DHRS condenser where it condenses back to liquid to return to the SG. Thus, the loop transfers heat from the RCS to the DHRS fluid using the SG and then from the DHRS to the reactor pool water. 3.4 Emergency Core Cooling System The ECCS consists of two or three independent reactor vent valves (RVVs), depending on NPM design, and two independent reactor recirculation valves (RRVs). The ECCS is initiated by simultaneously actuating the RVVs on the top of the RPV in the pressurizer region and the RRVs on the side of the RPV in the downcomer region. Opening the ECCS valves allows a natural circulation path to be established - water is vaporized in the core, leaves as steam through the RVVs, condenses and collects in the containment, and returns to the downcomer region inside the RPV through the RRVs. During normal operation, each ECCS valve is held closed by the hydraulic pressure across the valve main disc. Included in the RRV design is an inadvertent actuation block (IAB) consisting of a spring loaded arming valve in the vent port path from the main disc chamber to the © Copyright 2022 by NuScale Power, LLC 25

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 vent line. If the differential pressure across this arming valve is greater than a threshold value, the arming valve closes, which prevents the main disc chamber from discharging through the vent line, blocking the RRV from opening. The RRV does not open until the arming valve differential pressure decreases below the release pressure. The IAB is also used in the RVVs for some NPM designs. Provided the IAB device setpoint is reached, if applicable, the RVV and RRV components fail to the open (safe) position upon the loss of power, thus enabling reliable long-term cooling without operator actions, alternating current (AC) or direct current (DC) power, or make-up water. Successful operation of the ECCS requires isolation of the containment, such that the coolant inventory of the RCS is preserved. 3.5 Other Important Systems and Functions Other systems and functions that are important in mitigating plant response during a postulated non-LOCA event are discussed below. Reactor Coolant System The reactor coolant system (RCS) consists of the RPV, reactor core, riser, upper plenum, SGs (shell side), downcomer, lower plenum, and pressurizer (PZR). The arrangement of the RCS and the relative locations of the thermal centers in the core and the SGs promote buoyancy driven natural circulation flow. The RPV consists of a steel cylinder with an inside diameter of approximately 10 ft and an overall height of approximately 60 ft and is designed for a normal operating pressure of approximately 1850-2000 psia. Nozzles on the upper head provide connections for reactor safety valves (RSVs) and RVVs. The core configuration for an NPM consists of 37 fuel assemblies and 16 control rod assemblies (CRAs). The fuel assembly design is modeled from a standard 17x17 PWR fuel assembly with 24 guide tube locations for control rod fingers and a central instrument tube. The assembly is nominally half the height of standard plant fuel and is supported by five spacer grids. The U-235 enrichment is below the current U.S. manufacturer limit of 4.95 weight percent. Each NPM uses two once-through helical coil SGs for steam production. The SGs, which produce superheated steam, are located in the annular space between the RCS hot leg riser and the reactor vessel inside diameter wall. Each SG is designed to remove 50 percent of the rated core thermal power. The PZR provides the primary means for controlling RCS pressure. PZR heaters and spray maintain a constant reactor coolant pressure during operation. A steel PZR baffle plate, integral with the SG tube sheets and the RPV, acts as a thermal barrier and allows for surge flow between the PZR and the RCS. © Copyright 2022 by NuScale Power, LLC 26

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Feedwater System Feedwater from the condenser is pumped by condensate pumps to the condensate polishing equipment, where impurities are removed. Downstream of the polishing equipment, variable speed feedwater pumps supply flow to the feedwater heaters before the feedwater regulating valves control feed to the SGs. In unit operation, preheated feedwater is pumped into the tube side of the SGs where it boils. Upon receipt of a DHRS actuation signal the feedwater isolation valves close. Main Steam System Superheated steam produced in the SGs flows to a dedicated steam turbine. A generator, driven by the turbine, generates electric power that is delivered to the utility grid through a step-up transformer. A turbine steam bypass valve is provided that allows the reactor to remain in operation in the event of a turbine trip. Upon receipt of a DHRS actuation signal the MSIVs close (the steam system contains backup isolation valves in the event that an MSIV fails to isolate). Chemical and Volume Control System The primary functions of the CVCS are to purify reactor coolant, adjust the boron concentration in the reactor coolant, and supply spray flow to the pressurizer. Makeup and letdown operation can also be used to adjust the RCS inventory as needed. Equipment within the CVCS also allows for chemical addition to the reactor coolant, and heats the reactor coolant during startup. The CVCS includes demineralized water system isolation valves to mitigate boron dilution. When used for reactor coolant heating, the CVCS heats the RCS to the hot standby startup temperature, and develops natural circulation through the core sufficient to maintain the required RCS flow prior to nuclear heat addition. Containment Vessel The major safety functions of the CNV are to contain the release of radioactivity following postulated accidents, protect the RPV and its contents from external hazards, and to provide heat rejection to the reactor pool following ECCS actuation. Following an actuation of the ECCS, heat removal through the CNV rapidly reduces the containment pressure and temperature and maintains them at less than design conditions for extended periods of time. Steam is condensed on the inside surface of the CNV, which is passively cooled by conduction and convection heat transfer to the reactor pool water. © Copyright 2022 by NuScale Power, LLC 27

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Reactor Pool The reactor pool is a large pool located below the plant ground level in the Reactor Building. During normal plant operations, heat is removed from the pool through a cooling system and ultimately rejected into the atmosphere through a cooling tower or other external heat sink. In an event where AC power is lost, heat is removed from an NPM by allowing the pool to heat up and boil. Water inventory in the reactor pool is maintained at a level that is sufficient to provide at least three days of DHRS operation. © Copyright 2022 by NuScale Power, LLC 28

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 4.0 Transient and Accident Analysis Overview As part of defining the requirements of the non-LOCA transient system analysis methodology, the events to which the methodology applies and the specific event acceptance criteria applicable to the events are identified. The specific event acceptance criteria are derived from the regulatory requirements and guidance discussed in Section 2.2. The non-LOCA transient system analysis is part of several stages of analyses performed to confirm a plant design meets applicable acceptance criteria for a limiting set of AOOs, IEs, and accidents. The methodology includes the interfaces of the non-LOCA transient system analysis with other analysis methodologies, and identifies where margin to the event acceptance criteria is demonstrated. 4.1 Design-Basis Events and Event Classification The NuScale design-basis events for which the non-LOCA system transient analysis is performed, the event category, and the event classification are listed in Table 4-1. A broad spectrum of transients, accidents, and initiating events are considered in the scope of design basis analyses presented in Chapter 15 of a NuScale plant FSAR. The design basis events are identified based on: review of the NuScale plant systems to identify failures that would result in a design basis initiating event review of initiating events considered in the NuScale probabilistic risk assessment analyses to identify design basis initiating events. The probabilistic risk assessment initiating events included consideration of:

             -    the master logic diagram of failure mechanisms that may result in core damage
             -    industry generic data sources reviewed to identify initiating events
             -    advanced reactor probabilistic risk assessments review of similarities and differences of the plant design to previously evaluated NuScale plant designs As described in Section 3.0, an NPM is a natural circulation PWR with SGs that are integral to the reactor vessel. Many of the events analyzed for operating plants and in recent design certification applications are applicable to a NuScale design.

NuScale-specific events reflect unique aspects of a NuScale design such as the DHRS and vacuum conditions of containment during normal operation. The design basis events are categorized by type and expected frequency of occurrence. Limiting cases in each group are quantitatively analyzed and specific acceptance criteria for each postulated initiating event are applied. The design basis events that require non-LOCA system transient analysis are categorized into one of five categories:

1. increase in heat removal from the RCS In a NuScale design, this increase may be due to increased heat removal by the secondary system or due to increased heat removal to the containment.

© Copyright 2022 by NuScale Power, LLC 29

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

2. decrease in heat removal by the secondary system

3. reactivity and power distribution anomalies

4. increase in reactor coolant inventory

5. decrease in reactor coolant inventory The design-basis events are classified into one of three event categories:

1. AOOs - These are events that are expected to occur one or more times in the design life of the plant, conservatively quantified as events with a frequency of occurrence of 1x10-2 per module year or greater. An event that is not expected to occur in the design life of the plant may also be categorized as an AOO for conservatism.

2. Infrequent event - These are events that are not expected to occur during the design life of the plant. For IEs, the acceptance criteria are specified such that some fuel damage may occur but the radiological acceptance criteria are stricter than those imposed for accidents.

3. Postulated Accidents - These are design-basis events that are not expected to occur during the design life of the plant.

Historical precedent is used for event classification where the event is initiated by abnormal system conditions that are similar to those experienced in currently operating plants and certified designs. Event frequencies from the probabilistic risk assessment are considered for events that are unique to an NPM design or where an NPM design is such that the event frequency is expected to differ from currently operating plants and certified designs. The non-LOCA EM is applicable for the initiating events listed in Table 4-1. Because the control rod ejection accident analysis, LOCA analysis, and analysis of an inadvertent opening of one or more valves on the RPV initiating event are addressed by different methodologies, these initiating events are not included in Table 4-1. The non-LOCA system transient analyses are performed for a single module. In a NuScale design, the only shared safety-related system relied upon for event mitigation in the design basis system transient event analysis is the reactor pool portion of the UHS. Some initiating events may affect only a single module; others may affect multiple modules. In the non-LOCA system transient analysis calculations, the initial temperature of the reactor pool is bounded as described in Section 7.0. This approach bounds various module responses that may occur due to an initiating event that affects more than one module in a plant. © Copyright 2022 by NuScale Power, LLC 30

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 4-1 Design basis events for which the non-LOCA system transient analysis is performed, event category, and event classification Initiating Event Event Classification Increase in Heat Removal from the Reactor Coolant System Decrease in feedwater temperature AOO Increase in feedwater flow AOO Increase in steam flow AOO Inadvertent opening of SG relief or safety valve AOO Steam system piping failure inside or outside of containment Postulated accident Containment flooding/loss of containment vacuum AOO Inadvertent DHRS actuation(1) AOO Decrease in Heat Removal by the Secondary System Loss of external load AOO Turbine trip AOO Loss of condenser vacuum AOO Main steam isolation valve closure AOO Loss of nonemergency AC power to station auxiliaries AOO Loss of normal feedwater flow AOO Inadvertent DHRS actuation (1) AOO Feedwater system pipe break inside or outside of containment Postulated accident Reactivity and Power Distribution Anomalies Uncontrolled control rod assembly bank withdrawal from a subcritical or low AOO power startup condition Uncontrolled control rod assembly bank withdrawal at power AOO Control rod misoperation(2) AOO Single control rod assembly drop Control rod bank drop Single control rod assembly withdrawal Inadvertent decrease in boron concentration in the reactor coolant AOO Inadvertent System Operation that Increases Reactor Coolant Inventory Chemical and volume control system malfunction that increases reactor AOO coolant system inventory Decrease in Reactor Coolant System Inventory Failure of small lines carrying primary coolant outside containment Infrequent event Steam generator tube failure Postulated accident

1. Depending on the operating power at which an inadvertent DHRS actuation occurs, it may initially result in either an increase or decrease in heat removal from the RCS.

2. Control rod misoperation includes several types of events; those that require system transient analysis are identified.

© Copyright 2022 by NuScale Power, LLC 31

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 4.2 Design Basis Event Acceptance Criteria Safety analyses are performed to demonstrate that a nuclear power plant can meet applicable acceptance criteria for a limiting set of AOOs, IEs, and accidents. If the risk of an event is defined as the product of the events frequency of occurrence and its consequences, then the design of a plant should be such that events produce about the same level of risk. The acceptance criteria indicated by the GDC for nuclear power plants (Reference 4) reflect the risk of an event. Relatively frequent events such as AOOs are prohibited from resulting in serious consequences, but relatively rare events (postulated accidents) are allowed to produce more severe consequences. Design basis events for an NPM are categorized as AOOs, IEs, or postulated accidents. Table 4-2, Table 4-3, and Table 4-4 summarize the acceptance criteria applied for AOOs, IEs, and postulated accidents, respectively. The applicable acceptance criteria identified for each event are based on the event classification as identified in Table 4-1. For a limited number of events, a more conservative acceptance criterion may be applied than required based on the event classification. For many non-LOCA transient events, the specific acceptance criterion is not challenged during the event progression. For example, events that result in an increase in heat removal from the RCS may have a maximum RCS pressure higher than the initial operating pressure, but does not challenge the margin to the maximum RCS pressure acceptance criterion. In contrast, events that result in a decrease in heat removal from the RCS may result in an RCS pressurization that could challenge the maximum RCS pressure acceptance criterion. In Section 7.2, the acceptance criteria of interest for each non-LOCA event are identified. The acceptance criteria of interest are those where margin to the limit may be challenged during the event progression. In the event-specific transient analysis, sensitivity calculations are performed as necessary to ensure that the event meets acceptance criteria that may be challenged. These sensitivity calculations are performed to confirm that appropriately conservative inputs are specified to identify the case that results in minimum margin to the acceptance criterion of interest. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed are sufficient to demonstrate that margin to the acceptance criterion is maintained. A prime example of an acceptance criterion where an NPM design has significant margin is the maximum secondary system pressure. Unlike in typical PWR designs, in a NuScale design, the design pressure of the SG secondary side up to the second containment isolation valves is equal to the RCS design pressure. This feature supports the design and operation of the SG and DHRS. In a non-LOCA event that results in DHRS actuation, typically the maximum secondary side pressure occurs in the first minutes of the transient progression, following DHRS actuation. After DHRS is actuated, the fluid in the DHRS flows into the SG. Heat is transferred from the RCS primary system to the SG, where the DHRS loop inventory boils in the SG tubes. The steam flow is then condensed in the © Copyright 2022 by NuScale Power, LLC 32

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 DHRS condensers and the energy is transferred to the reactor pool UHS. The maximum pressure in the SG secondary side is limited to the saturation pressure at the temperature of the RCS fluid on the SG primary side. Therefore, the maximum secondary pressure is affected by the secondary side inventory and the primary side conditions at the time of DHRS actuation, and is less sensitive to a specific initiating event. The SG design pressure is significantly higher than pressures expected during DHRS operation. The margin to the SG design pressure is physically limited, based on the primary side conditions. The representative transient results in Section 8.0 demonstrate that significant margin to the maximum SG pressure acceptance criterion is maintained for all types of events. Therefore, extensive sensitivity calculations to maximize secondary side pressure are not necessary for the non-LOCA transient analysis calculations. © Copyright 2022 by NuScale Power, LLC 33

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 4-2 Acceptance criteria for anticipated operational occurrences How Acceptance Criterion is Parameter Acceptance Criterion Satisfied Maximum reactor coolant primary 110% of design pressure Margin to this acceptance system pressure criterion is demonstrated by the non-LOCA system transient analysis results. Maximum main steam secondary 110% of design pressure Margin to this acceptance system pressure criterion for the steam system piping up to the second containment isolation valve is demonstrated by the non-LOCA system transient analysis results. Minimum critical heat flux ratio > 95/95 critical heat flux ratio Margin to this acceptance (CHFR) Limit criterion is demonstrated by the subchannel analysis results. Subchannel analysis is outside scope of this topical report. Maximum fuel centerline melting temperature (adjusted for Margin to this acceptance temperature burnup effects) criterion is demonstrated by the subchannel analysis results. Subchannel analysis is outside scope of this topical report. An AOO should not result in a Margins to containment pressure Margin to this acceptance significant loss of reactor and temperature limits are criterion is demonstrated by the containment barrier maintained. peak containment pressure/temperature analysis results performed according to a separate analysis methodology. Maximum containment pressure and temperature analysis is outside scope of this topical report. An AOO should not generate a A postulated accident is not This acceptance criterion is postulated accident without other generated by the AOO. satisfied by demonstrating that faults occurring independently the other acceptance criteria are met. © Copyright 2022 by NuScale Power, LLC 34

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 4-3 Acceptance criteria for infrequent events How Acceptance Criterion is Parameter Acceptance Criterion Satisfied Maximum reactor coolant primary 120% of design pressure Margin to this acceptance system pressure criterion is demonstrated by the non-LOCA system transient analysis results. Maximum main steam secondary 120% of design pressure Margin to this acceptance system pressure criterion for the steam system piping up to the second containment isolation valve is demonstrated by the non-LOCA system transient analysis results. Fuel cladding integrity If the minimum CHFR is less than Margin to this acceptance or equal to the 95/95 CHFR limit, criterion is demonstrated by the the fuel rod is assumed to be failed. subchannel analysis results. If the maximum fuel centerline Subchannel analysis is outside temperature exceeds the melting scope of this topical report. temperature, the fuel rod is assumed to be failed. Containment integrity Margins to containment pressure Margin to this acceptance and temperature limits are criterion is demonstrated by the maintained. peak containment pressure/temperature analysis results performed according to a separate analysis methodology. Maximum containment pressure and temperature analysis is outside scope of this topical report. Release of radioactive material Calculated offsite doses are less Margin to this acceptance than 10% of the 10 CFR criterion is demonstrated by the 52.47(a)(2)(iv) (Reference 5) or accident radiological analysis 52.137(a)(2)(iv) (Reference 27) results. reference values. Accident radiological analysis is outside scope of this topical report. © Copyright 2022 by NuScale Power, LLC 35

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 4-4 Acceptance criteria for postulated accidents How Acceptance Criterion is Parameter Acceptance Criterion Satisfied Maximum reactor coolant primary 120% of design pressure Margin to this acceptance system pressure criterion is demonstrated by the non-LOCA system transient analysis results. Maximum main steam secondary 120% of design pressure Margin to this acceptance system pressure criterion for the steam system piping up to the second containment isolation valve is demonstrated by the non-LOCA system transient analysis results. Fuel cladding integrity If the minimum CHFR is less than Margin to this acceptance or equal to the 95/95 CHFR limit, criterion is demonstrated by the the fuel rod is assumed to be failed. subchannel analysis results. If the maximum fuel centerline Subchannel analysis is outside temperature exceeds the melting scope of this topical report. temperature, the fuel rod is assumed to be failed. Containment integrity Margins to containment pressure Margin to this acceptance and temperature limits are criterion is demonstrated by the maintained. peak containment pressure/temperature analysis results performed according to a separate analysis methodology. Maximum containment pressure and temperature analysis is outside scope of this topical report. Release of radioactive material Release does not result in offsite Margin to this acceptance doses in excess of the guidelines of criterion is demonstrated by the 10 CFR 52.47(a)(2)(iv) accident radiological analysis (Reference 5) or 52.137(a)(2)(iv) results. (Reference 27). Accident radiological analysis is outside scope of this topical report. © Copyright 2022 by NuScale Power, LLC 36

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 4.3 Non-LOCA Transient Analysis Process The main steps of the non-LOCA system transient analysis process for a specific transient event are:

1. Develop a plant base model NRELAP5 input.

2. Adapt NRELAP5 base model as necessary for specific event analysis, and desired initial conditions.

3. Perform steady state and transient system analysis calculations with NRELAP5.

4. Evaluate results of transient analysis calculations:

a. Confirm margin to maximum RCS pressure acceptance criterion

b. Confirm margin to maximum SG pressure acceptance criterion

c. Confirm appropriate transient run execution time

5. Identify which cases provide input for downstream subchannel analysis and extract boundary condition data.

6. Identify which cases provide input for downstream accident radiological analysis and extract boundary condition data.

The non-LOCA system transient analysis is performed and documented in accordance with NuScales QAP (Reference 3). The main steps of the non-LOCA system transient analysis are discussed in the following subsections. 4.3.1 Develop a Plant Base Model NRELAP5 Input NRELAP5 is NuScales system thermal-hydraulics code used to simulate an NPM system response during non-LOCA short-term transient event progression, for events that require system transient analysis. The NRELAP5 code was developed based on the Idaho National Laboratory (INL) RELAP5-3D© computer code. RELAP5-3D©, version 4.1.3 was used as the baseline development platform for the NRELAP5 code. RELAP5-3D© was procured and commercial grade dedication was performed by NuScale, and subsequently features were added and changes made to address unique aspects of an NPM design and licensing methodology. The NRELAP5 code includes models for characterization of hydrodynamics, heat transfer between structures and fluids, modeling of fuel, point reactor kinetics models, and control systems. NRELAP5 utilizes a two-fluid, non-equilibrium, non-homogenous fluid model to simulate system thermal-hydraulic responses. The NRELAP5 code is described in Reference 2. As discussed in Reference 2 and in Section 9.0, the NRELAP5 code has been developed and is maintained within NuScales QAP. © Copyright 2022 by NuScale Power, LLC 37

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The applicability of NRELAP5 for non-LOCA transient analysis is discussed in Section 5.0. As discussed in Section 4.2, the non-LOCA figures of merit evaluated with NRELAP5 are as follows: Primary side pressure Secondary side pressure (up to the second containment isolation valve) In addition, the RCS level response is evaluated with NRELAP5 for non-LOCA events that result in decrease in RCS inventory (SGTF and small line breaks outside of containment) to demonstrate that the inventory decrease is isolated in a timely manner and the DHRS provides effective decay heat removal for these events. The basis of an NRELAP5 plant base model is described in Section 6.0. An NRELAP5 plant base model is developed for a specific NPM design. Interfaces with the core design and fuel rod performance design that provide input to the transient analyses are described in the following subsections. 4.3.1.1 Interface with Core Design (Input to the Transient Analysis) Core design analysis performed in accordance with a methodology approved for a NuScale design provides input to the system transient analysis. The NuScale transient analysis methodology using NRELAP5 can be applied to a typical light water reactor fuel assembly design, and does not require that a specific code or suite of codes be used for the steady state core design analysis. The non-LOCA evaluation model methodology for specifying the input for the reactor kinetics model, axial power shape, and energy deposition factor are described in the following subsections. 4.3.1.1.1 Reactor Kinetics Model The system transient analyses are performed assuming either beginning of cycle (BOC) or end of cycle (EOC) conditions based on direction of conservatism for reactivity feedback for the transient. The total core power during a non-LOCA transient is the combination of the fission power and the decay heat. For the non-LOCA transients the fission power response is modeled using the separable point reactor kinetics model in NRELAP5. The NRELAP5 point reactor kinetics model computes both the immediate (prompt and delayed) fission power and the power from decay of fission products. In NRELAP5, input needed for the separable point kinetics model includes: Effective delayed neutron precursor yield of group i ( i ) © Copyright 2022 by NuScale Power, LLC 38

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Effective delayed neutron decay constant of group i ( i ) Prompt neutron generation time ( ) Reactivity feedback Reactivity changes from control rod movement (normal controls), or scram Decay heat model input In the non-LOCA evaluation model, the input to the point kinetics model is specified to give a conservatively high power response prior to actuation of reactor scram. The power response is biased by the input accounting for the reactivity feedback effects of moderator temperature, fuel temperature, and the normal control rod movement. After a reactor scram signal, the negative reactivity associated with insertion of the control and safety banks is conservatively modeled as described in Section 7.1.5. The moderator temperature coefficient (MTC) is a measure of the relative change in reactivity associated with a change in moderator (coolant) temperature. The Doppler temperature coefficient (DTC) is a measure of the relative change in the reactivity as the fuel temperature changes. In the non-LOCA evaluation model, reactivity feedback effects from moderator temperature changes and fuel temperature changes are conservatively bounded as described in Section 7.1.5. Negative reactivity insertion due to void generation impacts on moderator density is conservatively neglected. As described in Section 6.0 and Section 7.1.2, the rod control system and associated control logic are incorporated into the NRELAP5 model to allow simulation of reactivity changes (negative or positive) associated with normal control rod movement in response to postulated transients. In cooldown events, the normal rod control function attempts to increase average temperature by withdrawing control rods and therefore adding positive reactivity. For transients where operation of the normal rod control function provides more adverse consequences of the transient, the control rod movement is modeled to increase the power response during the transient. The decay heat power contribution is conservatively bounded high or low, as appropriate for the specific transient, as described in Section 7.1.5. Appropriate input based on the core design is used for other parameters needed as input to the point reactor kinetics model. 4.3.1.1.2 Axial Power Shape For the system transient analysis, a single channel core model is used, as described in Section 6.0. A nominal center-peaked average axial power shape is input for the single core channel for consistency with development of the reactivity feedback coefficients determined in the core design analyses. © Copyright 2022 by NuScale Power, LLC 39

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Uncertainties associated with the axial power shape and axial and radial power peaking factors that can affect the minimum critical heat flux ratio (MCHFR) and peak centerline fuel temperature are accounted for in the downstream subchannel analyses as described in Reference 6, supplemented by Reference 28. The design features of an NPM preclude challenge to the primary and secondary pressure acceptance criteria as discussed in Section 7.0 and as shown in the representative results in Section 8.0. Sensitivity studies on the axial power shape confirm that the primary and secondary system pressure, flow and fluid temperature responses are not significantly affected by the axial power shape. Therefore, use of a nominal center-peaked average axial power shape input is appropriate for the system transient analyses. 4.3.1.1.3 Energy Deposition Factor The energy deposition factor is the portion of the energy generated in the core that is directly deposited in the fuel. A bounding high energy deposition factor that results in all energy being deposited in the fuel is used in the non-LOCA analyses. For cooldown or reactivity insertion events that cause total core power changes, increasing the core energy deposited in the fuel slows the thermal-hydraulic response of the system to changing power levels. The effect of the energy deposition factor on the primary system pressure response is generally insignificant except in very fast reactivity events such as control rod ejection where Doppler reactivity feedback is important; control rod ejection analysis is outside the scope of this EM. Sensitivity studies on this parameter confirm that its effect on the system response is not significant with respect to demonstrating margin to acceptance criteria for NPM events addressed by the non-LOCA evaluation model due to the MPS design and selection of the analytical limits to actuate reactor trip. 4.3.1.2 Interface with Fuel Rod Performance Design (Input to the Transient Analysis) Fuel rod design analysis performed using a fuel performance code approved for a NuScale design provides input to the system transient analysis. The NuScale transient analysis methodology using NRELAP5 can be applied to a typical light water reactor fuel assembly design, and does not require that a specific code or suite of codes be used for the fuel performance analysis. The fuel assembly geometry, required material properties, and the fuel performance data needed for appropriate steady state initialization of the model are specified as input to the transient system analysis. The NuScale transient analysis requires input for the fuel rod and assembly geometry, fuel rod thermo-mechanical properties, and fuel rod performance information to assure adequate modeling of the initial fuel rod stored energy and transient response. © Copyright 2022 by NuScale Power, LLC 40

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 4.3.1.2.1 Fuel Rod and Fuel Assembly Geometry Nominal fuel rod and fuel assembly geometry information forms the basis for the NRELAP5 inputs required to describe the core heat structures, core coolant channels, and bypass channels. Pressure losses due to fuel assembly grid spacers and wall friction are accounted for. The NRELAP5 model is described in Section 6.0. 4.3.1.2.2 Fuel Rod Material Properties Fuel rod material properties appropriate for the fuel assembly design are specified by user input in the NRELAP5 model. Fuel rod material properties needed include: fuel pellet thermal conductivity fuel pellet specific heat fuel pellet density cladding thermal conductivity cladding specific heat cladding density Nominal fuel rod material properties as a function of temperature are specified. For the fuel pellet properties, UO2 properties are used to reflect the core average response. Since the fuel pellet thermal conductivity is a function of burnup and fuel temperature, and the fuel pellet thermal conductivity degrades with burnup, a representative time-in-cycle core average burnup is used to calculate the fuel thermal conductivity as a function of temperature. 4.3.1.2.3 Fuel Rod Performance Data The fuel rod gap conductance, specific heat and density in the NRELAP5 fuel rod heat structure are used to set the initial core average fuel temperature. Bounding values for fuel rod gap conductance are selected to provide conservatively high or low core average fuel temperature for the time-in-life of interest for the calculation. The conservative core average fuel temperatures are confirmed on a cycle-specific basis. 4.3.2 Adapt Plant Base Model NRELAP5 Input for Event Transient Analysis The basis of an NRELAP5 plant base model is described in Section 6.0. An NRELAP5 plant base model is adapted as necessary to perform the specific event analysis. Section 7.0 describes the non-LOCA analysis methodology, including © Copyright 2022 by NuScale Power, LLC 41

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 conservative biasing of initial and boundary conditions, single failures, and loss of power scenarios for the event analyses. 4.3.3 Perform NRELAP5 Steady State and Transient System Analysis Calculations For each analysis, one or more steady state initialization calculations are developed. After acceptable steady state conditions are obtained, the transient calculations are performed. Section 7.0 describes the non-LOCA analysis methodology including the steady state initialization and performance of null transients prior to the transient calculation. 4.3.4 Evaluate Results of Transient Analysis Calculations The results of the transient analysis calculations are assessed to confirm that the system transient response is acceptable. 4.3.4.1 Maximum Pressure Acceptance Criteria Based on the results of the transient analysis calculations performed for an event, margin to the maximum RCS pressure acceptance criterion and margin to the maximum SG secondary pressure acceptance criterion are confirmed. 4.3.4.2 Short-Term Transient Duration For each transient calculation, the following parameters are reviewed to assure that the transient calculation is executed for an appropriate duration to confirm that design basis event acceptance criteria are met. MPS actuations expected in direct response to the initiating event, for mitigation of the design basis event, have occurred if reactor trip occurs during the transient, the nuclear heat source is reduced to decay heat levels and decreases with time core average temperature is stable, or decreasing following reactor trip RCS pressure is stable or decreasing RCS fluid inventory is stable containment pressure is stable or decreasing These conditions are demonstrated for a reasonable time, typically a few hundred seconds, following the last safety system actuation expected to occur in the short term transient progression, to demonstrate that core cooling is established and conditions that result in minimum margin to the acceptance criteria have occurred. © Copyright 2022 by NuScale Power, LLC 42

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 4.3.5 Identification of Cases for Subchannel Analysis and Extraction of Boundary Condition Data For NPM non-LOCA events, VIPRE-01 is used for performance of subchannel analysis calculations. Reference 6, supplemented by Reference 28, describes the subchannel analysis methodology. In the transient subchannel analysis, the following acceptance criteria are assessed: MCHFR maximum fuel centerline temperature For non-LOCA events that require subchannel analysis, the NRELAP5 transient analyses provide the following input from the system transient calculation to the downstream subchannel analysis: reactor power as a function of time core exit pressure as a function of time core inlet temperature as a function of time total system flow rate as a function of time In the NRELAP5 system transient analysis, cases for downstream subchannel analysis are identified based on the conservative bias directions for the boundary condition input and considering an NPM natural circulation design. The conservative bias directions are discussed below, followed by a description of the methodology for identifying cases for downstream subchannel analysis. As identified in Reference 6, supplemented by Reference 28, for the system transient parameters provided by NRELAP5, the conservative bias directions to minimize the CHFR are as follows: maximum reactor power (higher power increases the actual heat flux) maximum core inlet temperature (higher temperature reduces energy addition needed to raise coolant to saturated conditions) minimum system flow rate (minimum flow is conservative as there is less coolant flow in the reactor core available for heat transfer) The effect of pressure on critical heat flux (CHF) is established on a case by case basis to determine the appropriate direction for biasing. In a NuScale design, for a given reactor module operating condition, reactor power, core inlet temperature and system flow rate are tightly coupled. As described in Section 7.1, ranges in these parameters are considered as part of biasing the system transient analysis steady state initial conditions. The NRELAP5 system analysis methodology for determining the limiting CHF cases for downstream subchannel analysis is primarily dependent on the limiting initialization. The CHF cases are evaluated at the minimum flow initialization. Other initial conditions are forced to the © Copyright 2022 by NuScale Power, LLC 43

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 limiting initialization for a given transient progression to ensure the maximum power and core inlet fluid temperature are reached prior to reactor trip system actuation. For example, in the case of a heatup event, the RCS increases in temperature, causing a pressurizer insurge and subsequent increase in pressure. The limiting CHF scenario is the transient progression that results in the highest core outlet temperature at the time of reactor trip on high pressure, which is generally the faster heatups where the pressurizer initialization is biased to delay the high pressure trip. For some transients, a spectrum of cases is analyzed from the limiting initialization. ((

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After the system transient analysis calculations are performed and assessed, for events that require subchannel analysis, a number of cases are identified as limiting for MCHFR. For the limiting cases selected, the required system transient parameters are tabulated as a function of time for input to the downstream subchannel analysis calculations, to calculate margin to CHF. The system transient parameters are provided for subchannel analysis for sufficient time for the subchannel analyses to demonstrate that the MCHFR has occurred, typically 10-15 seconds following reactor trip. © Copyright 2022 by NuScale Power, LLC 44

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 4-1 ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 4-2 ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 4-3 ((

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                                                                                                       }}2(a),(c) 4.3.6         Identification of Cases for Accident Radiological Analysis Radiological acceptance criteria are assessed in the accident radiological analyses.

Reference 8 describes the NuScale accident radiological source term analysis methodology. This section only describes radiological analyses cases that involve nuclear steam supply system transients (i.e., primary and secondary coolant system transients) and does not encompass any radiological analyses involving spent fuel movement or postulated failures in the radioactive waste system. The transient analyses can be used to provide input to the accident radiological analyses for events that result in RCS fluid loss outside of containment such as a break of small RCS piping outside of containment or SGTF. For these events, one or more transient analysis cases are identified to provide conservative input to the © Copyright 2022 by NuScale Power, LLC 47

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 accident radiological analysis. The conservative bias directions for the transient analysis input to the accident radiological analysis are: Maximum integrated mass release outside of containment prior to isolation of the RCS mass release. Basis: For a constant radionuclide concentration in the RCS, the greater the mass released outside of containment, the more severe the radiological consequences. Maximum integrated mass release between time of reactor trip and time of isolation of the RCS mass release. Basis: Reference 8 describes how iodine spiking is accounted for in the accident radiological analyses. During a transient progression, changes in reactor power, RCS average temperature, or RCS pressure could result in iodine spiking, changing the radionuclide concentration in the RCS. Consistent with Standard Review Plan Section 15.6.2 (Reference 9), for accident radiological calculations assuming a coincident iodine spike (iodine spike occurring during the event), the iodine spiking is assumed to begin at the time of reactor trip as the result of the reactor shutdown or depressurization of the primary system. In some cases, particularly for smaller breaks in RCS piping, the time between reactor trip and isolation of the break flow could be extended compared to larger break sizes. The increased time between reactor trip and isolation increases the time of mass release when the RCS radionuclide inventory reflects iodine spiking. For each transient analysis case identified, the input provided for accident radiological analysis includes: Time of reactor trip if it is calculated to occur Isolation time, at which point release of RCS fluid outside of containment is stopped. The isolation may be due to MPS response or due to operator action (as identified in Section 7.1.7, operator actions may be taken to prevent abnormal operating events from resulting in more severe events). RCS fluid mass release outside of containment as a function of time System transient response parameters such as

                  -   Reactor power as a function of time
                  -   RCS average temperature as a function of time
                  -   RCS pressure as a function of time
                  -   Secondary side feedwater and steam flow rates as a function of time As an alternative to transient analysis, the accident radiological analysis can use bounding values for both mass release and isolation times. An example bounding approach is provided as follows. If the MPS of an NPM design includes reactor trip and isolation setpoints based on pressurizer level, these setpoints effectively limit the RCS inventory that can be released to the difference between a high pressurizer level and the pressurizer level associated with the isolation, and limit the time between reactor trip actuation and completion of isolation valve closure. A bounding mass

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 release based on the MPS pressurizer level setpoints should also include the additional mass released between the time the setpoint is reached and the time isolation is complete, as well as account for pressurizer level uncertainty. Using a bounding mass release and isolation time, the accident radiological analysis can be completed independent of system transient analyses; however, the system transient analyses can still be used to confirm the assumptions are bounding. This alternative approach results in more limiting accident radiological analysis results and is therefore typically used when the accident radiological analysis results are expected to have margin. © Copyright 2022 by NuScale Power, LLC 49

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.0 NRELAP5 Applicability for Non-LOCA Transient Analysis 5.1 Non-LOCA Phenomena Identification and Ranking Table and Evaluation of High-Ranked Phenomena 5.1.1 Phenomena Identification and Ranking Table Process The PIRT process is a systematic way of gathering information from experts on a specific subject and ranking the importance of the information to meet some decision-making objective. It has been applied to many nuclear technology issues to help guide research and develop activities to satisfy regulatory requirements. The purpose of the NuScale non-LOCA PIRT is to provide an assessment of the relative importance of phenomena and processes that may occur in a NuScale Power Module (NPM) during non-LOCA events in relation to specified figures of merit (FOMs). This assessment is part of the process prescribed by Regulatory Guide 1.203 (Reference 1). The non-LOCA PIRT was developed by a panel of experts for an NPM and was built upon the state-of-knowledge at the time of its development. Non-LOCA events can be divided into several different event types based on the main effect on the reactor coolant system, as described in Section 5.1.2. A comprehensive, integrated PIRT was performed for the range of non-LOCA event types and phases of the event progression. The PIRT panel considered an NPM design to identify systems, components, and subcomponents of its design for which phenomena were assessed. The panel then followed the PIRT process to identify and rank phenomena considering the level of importance for each phenomena relative to identified FOMs for the different non-LOCA event types and phases of the transient progression. The panel also established a knowledge ranking for each of the phenomena. Following the development of the PIRT, additional insights from testing, code validation, plant calculations and analysis were established. The following discussions of the PIRT phenomena, importance, knowledge rankings, and how the phenomena are addressed reflect these developments as appropriate. The PIRT was originally developed based on the specific NPM design in development at that time. Subsequent design work has resulted in other NPM designs within the confines of the general NPM description in Section 3.0. An assessment of the applicability of the PIRT to other NPM designs was performed. The assessment included comparisons between NPM designs of: the normal operating conditions at full power; design limits; and the geometric parameters. The results of the assessment identified no new phenomena that are applicable during the non-LOCA event phases. The assessment concluded that the non-LOCA PIRT remains applicable to NPM designs that are consistent with the description in Section 3.0. Further insights from the assessment are included in the discussions of the PIRT in the following sections. © Copyright 2022 by NuScale Power, LLC 50

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.2 Non-LOCA Event Scenarios and Phases As described in Section 4.1, the non-LOCA events for an NPM are divided into five event categories based on the type of effect on the RCS. These categories are summarized below.

1. increase in heat removal from the reactor coolant system

2. decrease in heat removal by the secondary system

3. reactivity and power distribution anomalies

4. increase in reactor coolant inventory

5. decrease in reactor coolant inventory For the non-LOCA PIRT, the panel evaluated five design-basis non-LOCA events, one representative of each category:

Main steam line break inside containment: Representative of events that result in an increase in heat removal from the RCS (cooldown/depressurization events) Feedwater line break inside containment: Representative of events that result in a decrease in heat removal from the RCS (heatup/pressurization events) CRA withdrawal: Representative of events that result in a reactivity increase CVCS malfunction: Representative of events that result in an increase in RCS inventory SGTF: Representative of events that result in a decrease in RCS inventory The PIRT panel divided the short-term, non-LOCA event progression into three phases, generic for each transient type: Phase 1 - Pre-trip transient Phase 1 begins with the event initiation. The RCS power, pressure and flow rates increase or decrease from the normal power conditions, depending on the event type. The DHRS is inactive. This phase ends with actuation of the MPS response to the off-normal NPM conditions. Phase 2 - Post-trip transition Phase 2 begins with MPS actuation of reactor trip and, for most non-LOCA events, actuation of the DHRS in response to faulted secondary heat removal conditions. In cases where DHRS is actuated, the DHRS actuation valves open and the normal secondary side flow paths from the feedwater system and to the main steam system (MSS) are isolated. The secondary side DHRS loop pressurizes as decay and residual heat are transferred from the primary, causing the DHRS loop inventory to boil off in the steam generator and condense in the DHRS condenser. The reactor coolant system power and flow rates transition towards decay heat levels. © Copyright 2022 by NuScale Power, LLC 51

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Phase 3 - Stable natural circulation During phase 3, stable primary side natural circulation conditions exist with reactor coolant system power and flow rates reflecting decay power levels. In cases where DHRS is actuated, stable natural circulation in the DHRS loop is established. Secondary side flow rates and pressures in the DHRS loop decrease as the primary side pressure and temperature decrease. Section 4.3 describes criteria for the short-term non-LOCA transient end conditions. 5.1.3 Phenomena Identification and Ranking Table Figures-of-Merit and Phenomenon Ranking The PIRT panel identified figures-of-merit for each phase of the non-LOCA transient. The FOMs reflect the non-LOCA event acceptance criteria (Section 4.2) and important factors relative to an NPM design. Phase 1 - Pre-trip transient CHF - Demonstrating that margin to CHF is maintained for AOOs, or identifying the number of fuel rods that exceed CHF for IEs or postulated accidents is a primary non-LOCA acceptance criterion. The non-LOCA event analyses confirm that the plant system design is such that margin to CHF is maintained until the module protection system actuates to mitigate the event. In an NPM design, margin to CHF may be challenged during cooldown events or reactivity insertion events. Primary pressure - The maximum primary system pressure is one of the non-LOCA acceptance criteria to demonstrate acceptable RCS performance. The non-LOCA event analyses confirm that the plant system design is such that margin to the maximum primary system pressure limits is maintained until the module protection system actuates to mitigate the event. In an NPM design, margin to the maximum primary system pressure limits may be challenged during heatup events or events that increase RCS inventory. Phase 2 - Post-trip transition CHF - See above. Primary pressure - See above. Secondary pressure - The maximum secondary system pressure is one of the non-LOCA acceptance criteria to demonstrate acceptable RCS performance. In an NPM design, as the DHRS is actuated in Phase 2, the maximum secondary side pressure in the SG can occur. Containment pressure - The maximum containment pressure is an indicator of containment integrity. © Copyright 2022 by NuScale Power, LLC 52

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 In an NPM design, vapor released into containment is condensed on the containment wall and heat is transferred through the containment wall directly to the reactor pool portion of the UHS. Non-LOCA events could result in mass and energy release into containment. The containment design is intended to effectively transfer heat to the UHS so that containment integrity is maintained. Phase 3 - Stable natural circulation CHF - See above. Coolant mixture level - The RCS primary mixture level represents the boundary between a two-phase or single-phase liquid region below the level and a single-phase vapor region above the level. This level indicates whether the primary side natural circulation flow path is maintained. In an NPM design, depending on the initial and boundary condition assumptions such as initial RCS level and temperature, decay heat, and number of DHRS trains operating, the DHRS heat removal may provide sufficient cooling of the RCS that the increased primary side liquid density results in coolant volume shrinkage sufficient to decrease the RCS water level to below the top of the riser, resulting in interrupted natural circulation. If interruption of natural circulation occurs, it is the end of Phase 3, and it occurs well after reactor trip and engineered safety features have responded to the initiating event. Subcriticality - Maintaining subcriticality following reactor trip limits the nuclear fuel heat source to decay heat levels. If the soluble boron concentration in the core is significantly reduced following reactor trip and the MPS response to the initiating event, then maintaining subcriticality could be adversely affected. While RCS cooldown following reactor trip also affects the net reactivity, the non-LOCA analysis methodology is applicable to DHRS cooling shutdown evaluations only when the mixture level is above the top of the riser and primary side natural circulation flow is maintained. In an NPM design, the boron in the primary system during Phase 3 is limited to the soluble boron at the RCS critical boron concentration from normal operating conditions. During Phase 3 of the non-LOCA event progression, primary side natural circulation flow is maintained. The addition of supplemental boron by the ECCS, if applicable to the design, is outside the non-LOCA Phase 3 scope and is assessed separately (Reference 26). Each phenomenon identified in the PIRT was assigned an importance ranking and knowledge level ranking. Table 5-1 and Table 5-2 describe the importance rankings and the knowledge level rankings considered by the PIRT panel. Table 5-1 Importance rankings Importance Ranking Definition High (H) Significant influence on FOM Medium (M) Moderate influence on FOM Low (L) Small influence on FOM Inactive (I) Phenomenon not present or negligible © Copyright 2022 by NuScale Power, LLC 53

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-2 Knowledge levels Knowledge Level Definition 4 Well-known/small uncertainty 3 Known/moderate uncertainty 2 Partially known/large uncertainty 1 Very limited knowledge/uncertainty cannot be characterized 5.1.4 Highly Ranked Phenomena The following subsections summarize the phenomena listed in Table 5-3 that were ranked high importance by the PIRT panel in at least one of the three phases of the non-LOCA short-term transient response scenarios. The knowledge level assigned by the PIRT panel, and the systems and components where the phenomenon was ranked as high importance are also included. The non-LOCA PIRT is a comprehensive, integrated PIRT covering the range of non-LOCA event types and the phases of event progression. The NPM systems, components, and relevant phenomena are considered in detail. As discussed in Section 1.1, NRELAP5 is the system thermal-hydraulics code used to simulate an NPM system response during the non-LOCA short-term transient event progression. The NRELAP5 assessments performed as part of the development of Reference 2 demonstrate the capability of the code to simulate an NPM response to LOCA events. The high-ranked phenomena identified by the PIRT process for the non-LOCA transients were evaluated with respect to the high-ranked phenomena identified by the PIRT process for the LOCA scenarios, and the code assessments performed as part of the development of Reference 2. A gap analysis was performed to identify high-ranked phenomena for non-LOCA transients that were not assessed as part of the development of Reference 2. High-ranked phenomena for non-LOCA events which were not assessed as part of the development of Reference 2 were addressed in a variety of ways that included:

1. Additional NRELAP5 code assessments performed against separate effects or integral effects test data.

2. Code-to-code benchmark performed between NRELAP5 and an independent system thermal-hydraulics code.

3. The phenomenon was addressed as part of the downstream subchannel analysis.

4. The phenomenon was addressed by specifying appropriately conservative input to the system transient analysis.

The following subsections also describe the means by which each high-ranked phenomenon for non-LOCA events was assessed. In this instance, the NRELAP5 assessments performed as part of the development of Reference 2 were combined © Copyright 2022 by NuScale Power, LLC 54

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 with one or more of the means described above to qualify NRELAP5 for the high rank non-LOCA PIRT phenomena. Specific details related to the qualification of NRELAP5 include:

1. The NRELAP5 assessment against Korea Advanced Institute of Science and Technology (KAIST) high pressure condensation data, as presented in Reference 2, is described in Section 5.3.1 for convenience.

2. Separate effects testing of the full-length DHRS was performed at the NIST-1 facility and the NRELAP5 assessment is summarized in Section 5.3.2.

3. An integral effects test of an NPM response to a decrease in heat transfer from the secondary side, and integral effects test of DHRS operation was performed at the NIST-1 facility, and is summarized in Section 5.3.3.

4. A code-to-code benchmark was performed to assess the NRELAP5 prediction of an NPM response to reactivity insertion events as described in Section 5.3.4.

5. The NRELAP5 assessments against the SIET data (TF-1 and TF-2) and other legacy experiments are summarized in Reference 2. Due to the importance of heat transfer through the SG as part of the DHRS loop for non-LOCA transients, further review of these assessments is documented in Section 5.3.5.

6. A computational fluid dynamics model was used to evaluate heat transfer between primary reactor coolant and the SG helical coil tubes. Results of the simulation were compared to the ESDU model implemented in NRELAP5, as described in Section 5.3.6.

7. An integral effects test of an NPM response to a decrease in heat transfer from the secondary side, including DHRS operation, was performed at the NIST-2 facility, as is summarized in Section 5.3.7.

Table 5-3 High-ranked phenomena for non-LOCA events ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-3 High-ranked phenomena for non-LOCA events (Continued) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.1 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.2 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.3 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.4 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.5 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.6 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.7 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.1.4.8 (( }}2(a),(c) ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.2 Evaluation of Non-LOCA Phenomena Identification and Ranking Table High-Ranked Phenomena The information in this section is moved to Section 5.1.4. Evaluation results of high ranked phenomena and text discussing how they are addressed are incorporated into the discussion of each of the phenomenon listed in Section 5.1.4. This section is retained to maintain accuracy in other documents that may have referenced material from this document in sections located after this section. 5.3 NRELAP5 Validation and Assessments for Non-LOCA In Section 5.1 and Section 5.2, the high-ranked phenomena from the non-LOCA PIRT are summarized and how they are assessed in the non-LOCA evaluation model is identified. As identified in Table 5-3, additional assessments were performed to support the qualification of the NRELAP5 code for some of the non-LOCA high-ranked phenomena. This section summarizes the additional assessments performed. This section also discusses a limited number of assessments performed as part of the LOCA evaluation model development that demonstrate qualification of NRELAP5 for prediction of heat transfer from the RCS to the SG to the DHRS. Heat transfer from the RCS to reactor pool via the SG and the DHRS is important for prediction of non-LOCA transient progression after DHRS actuation, to demonstrate heat removal from the core. The NRELAP5 prediction of heat transfer in the DHRS is assessed with the KAIST high-pressure condensation experiments (Section 5.3.1), and with separate effects testing of the full-length DHRS at the NIST-1 facility (Section 5.3.2). The KAIST data was assessed as part of the LOCA evaluation model development (Reference 2) and key results are summarized in this report. The NRELAP5 code validation for the helical coil SG was assessed as part of the LOCA evaluation model (Reference 2), with testing performed at the SIET facility and other legacy experiments. The operating ranges expected during the non-LOCA transients are assessed to demonstrate the adequacy of the validation for the range of conditions expected in the non-LOCA transients (Section 5.3.5). Integral effects testing was performed at the NIST-1 test facility to support the NRELAP5 validation for non-LOCA events (Section 5.3.3). The testing assessed primary side pressurization and heat-up due to a decrease in heat transfer from the secondary side, and primary and secondary side conditions during operation of the scaled integral DHRS. In addition to the validation against test data, a code-to-code benchmark against the RETRAN-3D code (Reference 12) was performed to assess the NRELAP5 prediction of an NPM power and integral primary system response to reactivity insertion events (Section 5.3.4). © Copyright 2022 by NuScale Power, LLC 115

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Computational fluid dynamics was used to independently assess the heat transfer correlation incorporated into NRELAP5 for the helical SGs of NuScale designs (Section 5.3.6). Integral effects testing was performed at the NIST-2 test facility to support the NRELAP5 validation for non-LOCA events (Section 5.3.7). The testing assessed SG and DHRS performance during a decrease in heat transfer from the secondary side. Analysis of NPM transient response extends the applicability of the DHRS performance to small break LOCA sequences before ECCS actuation. Agreement between code predictions and data or for the code to code comparison is assessed according to the criteria described in RG 1.203 (Reference 1): Validation Requirements: Excellent Agreement applies when the code exhibits no deficiencies in modeling a given behavior. Major and minor phenomena and trends are correctly predicted. The calculated results are judged to agree closely with data. Reasonable Agreement applies when the code exhibits minor deficiencies. Overall, the code provides an acceptable prediction. All major trends and phenomena are predicted correctly. Differences between calculated values and data are greater than are deemed necessary for excellent agreement. Minimal Agreement applies when the code exhibits significant deficiencies. Overall, the code provides a prediction that is not acceptable. Some major trends or phenomena are not predicted correctly, and some calculated values lie considerably outside the specified or inferred uncertainty bands of the data. Insufficient Agreement applies when the code exhibits major deficiencies. The code provides an unacceptable prediction of the test data because major trends are not predicted correctly. Most calculated values lie outside the specified or inferred uncertainty bands of the data. Note that the results of assessments previously performed using the version of NRELAP5 associated with Reference 25 (or earlier) are retained throughout Section 5.3. The assessments were re-performed using the Reference 29 version of NRELAP5. A limited selection of the results of the updated assessments are provided. The NIST-2 assessments provided in Section 5.3.7 were performed with the Reference 29 version of NRELAP5. 5.3.1 KAIST The NRELAP5 prediction of high pressure condensation experimental data produced at the KAIST experimental facility was assessed as part of the LOCA EM development (Reference 2). The phenomenon addressed with the KAIST experimental cases provides test data for condensation inside tubes and heat transfer across tubes for the DHRS. As part of demonstrating qualification of NRELAP5 for © Copyright 2022 by NuScale Power, LLC 116

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 calculation of DHRS heat transfer rates, the assessment of NRELAP5 against the KAIST tests is summarized. The data from the KAIST experiment are qualified for use by applying non-mandatory guidance provided by NQA-1 2008/2009 Addendum (Reference 20). 5.3.1.1 Facility Description and Range of Experimental Data Assessed A schematic of the KAIST experimental facility is shown in Figure 5-1. The maximum design pressure and temperature of the test facility were 7.5 MPa (1088psia) and 300 degrees C (572 degrees F), respectively. The major components of the test facility include: SG, which supplied steam (maximum power 200 kW, test section tube, the cooling pool (cools the test section), steam line (transports steam from SG to the test section inlet), condensate drain line, lower plenum (or condensate collection tank) and air supply system. The test section was immersed in the cooling pool and was cooled by boiling and single-phase convective heat transfer on the outside surface of the test section (Reference 24). The test section was a vertical tube with an inside diameter of 4.62 cm and an effective heat transfer length of 1.8 m. The thickness of the tube wall was 2.3 mm. To reduce the entrance effect, the top 0.5 m length of the test section was insulated. The test section was submerged in the cooling pool (1.2 m x 1.2 m x 2.5 m). A steam line with an inside diameter 2.34 cm was connected from the top of the SG to the top of the test section. The condensate from the test section was drained to the lower plenum (or condensate collection tank) by gravity and then pumped back to the SG. Table 5-4 summarizes the KAIST and NPM decay heat removal system tube geometry. Table 5-5 summarizes the operational range covered by the KAIST experiment and in the NPM decay heat removal system. Table 5-6 shows a comparison between the NRELAP5 discretization between the KAIST and NPM decay heat removal system condenser tubes. Based on Table 5-4, Table 5-5, and Table 5-6 it is concluded that the geometry and operating range comparisons are acceptable for validation. ((

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Figure 5-1 Schematic of KAIST test facility

     © Copyright 2022 by NuScale Power, LLC DP Steam Line T

4000 mm P PS SafetyValve Water Atom. CV P Pool T T T Secondary P DP T T Condensor DP T Vent T T CV Tank T DP T Air Flow Steam DP Generator T CV Air Flow T Non-Loss-of-Coolant Accident Analysis Methodology T T T Air Supply Air Heat Exchanger Tank DP Compressor Lower T Plenum DP Feed & Drain T CV 200 kW Power TR-0516-49416-NP T T Water Line Electric Heater Water Recirculation Water Heat Exchanger Pump Revision 4 118

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-4 Comparison between NuScale Power Module decay heat removal system and KAIST test section dimensions Parameter KAIST Test Section NPM DHRS Geometry Single circular cross-section tube For a single DHRS condenser tube ID, cm (in) 4.62 (1.818) 2.79 (1.097) OD, cm (in) 5.08 (1.91) 3.34 (1.315) Active length, m (ft) 1.8 (5.9) 2.648-2.874 (8.7-9.4 ft) Wall thickness, mm (in) 2.3 (0.091) 2.8 (0.109) Material Stainless steel Stainless steel Table 5-5 Comparison between NuScale Power Module decay heat removal system and KAIST range of operations Parameter KAIST NPM DHRS(1) SG pressure, psia (MPa) 155.3 to 1059.5 (( (1.071 to 7.305) }}2(a),(c) Steam flow rates lbm/hr (kg/s) 79.37 to 793.7 (( (0.01 to 0.1) }}2(a),(c) SG steam temperature F (C) 362.12 to 552.2 (( (183.4 to 289) }}2(a),(c)

1. Approximate range expected for DHRS operation during the non-LOCA short-term response

2. Flow rate per tube Table 5-6 Comparison between NuScale Power Module decay heat removal system and KAIST NRELAP5 model nodalization KAIST Condenser Tube NPM DHRS Condenser Tube Parameter NRELAP5 Nodalization NRELAP5 Nodalization L/D 3.788 to 6.494 (( }}2(a),(c)

Total number of nodes 8 (( }}2(a),(c) © Copyright 2022 by NuScale Power, LLC 119

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.1.2 Experimental Procedure Conditions into the condenser tubes and pool were specified. The experiments were started by purging all non-condensable gas (i.e., air) from the test loop. This purge was done by supplying steam to the test loop and venting it to the atmosphere through the vent valve located below the test section. After all non-condensable gas was purged, the vent valve was closed and the test section was allowed to fill with the condensate by keeping the condensate drain valve closed. After the test section was completely filled, the SG pressure was increased to the test pressure. As soon as the test pressure was reached, the condensate drain valve was opened and the condensate recirculation pump was started. A constant water level in the lower plenum was maintained by control of the recirculation pump flow rate. Data acquisition was started after the process had reached a steady state. 5.3.1.3 Assessment Results The comparison results of condensed liquid flows, heat transfer coefficients, and inner wall temperatures show reasonable to excellent agreement between the calculated NRELAP5 and the KAIST measured experimental data. This agreement is a result of implementation of the 2009 extended Shah correlation in NRELAP5, which is intended to improve the predicted high pressure condensation response. Figure 5-2 presents a summary of the measured versus calculated assessments of the KAIST steam condensation experiments. The overall majority of the assessments lie within the experimental uncertainty (28 percent for heat transfer coefficient). Figure 5-2a presents the same comparison as Figure 5-2, but with the Reference 29 version of NRELAP5. © Copyright 2022 by NuScale Power, LLC 120

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-2 Measured vs predicted heat transfer coefficient Figure 5-2a Measured vs predicted heat transfer coefficient © Copyright 2022 by NuScale Power, LLC 121

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.2 NIST-1 Decay Heat Removal System Separate Effects Tests 5.3.2.1 NIST-1 Facility An NPM design relies on natural circulation flow to remove energy produced in the core. Energy is transferred to the secondary side as the primary coolant flows down over the SG tube coils. During off-normal transients, an NPM design relies on natural circulation driven by steam condensation within the CNV or the DHRS to cool the RPV. Energy within the CNV is transferred through the CNV walls into the surrounding UHS, represented by the reactor cooling pool. An NPM containment vessel is designed to accept and promote steam condensation at pressures varying from vacuum to maximum design pressures. Due to the unique nature of an NPM design the number of IET facilities is limited. A scaled facility of an NPM was constructed at Oregon State University (OSU), referred to as the NuScale Integral System Test-1 (NIST-1) facility, to assist in validation of the NRELAP5 system thermal-hydraulic code. The NIST-1 facility is a scaled facility of an NPM and consists of the major components in an NPM. These components include: an RPV, a helical coil SG system with a DHRS, a CNV, and a cooling pool vessel (CPV). The NIST-1 facility was originally conceived at OSU in 2000 as a proof-of-concept testing platform for development of small modular reactor (SMR) technology. During this period it was referred to as the multi-application small light water reactor (MASLWR) facility (Reference 11). Although the NuScale design was based on MASLWR, the concept has evolved considerably since NuScales inception in 2008. At the time that NuScale was formed, the facility was renamed the NIST facility. The NIST facility is a scaled, non-nuclear reactor that uses electric heater rods to analogously represent the heat produced from fission. It is designed to produce experimental data in support of verification and validation of thermal-hydraulic codes. In 2014 and 2015, the original NIST facility was modified by NuScale to necessitate accurate simulation and to bring the facility in-line with the NuScale plant design configuration at that time. A scaling analysis was employed for design of the NIST test facility to ensure that the facility design is capable of capturing important plant phenomena with minimal distortions. Following the upgrade, the NIST facility became the NIST-1 facility. Updates to the NIST facility included in NIST-1 are: installation of a DHRS scaling of the RVVs and RRVs to the plant design replacement of the existing CNV heat transfer plate (HTP), ECCS, and CPV; increased containment pressure rating from about 300 psi to approximately 1,000 psi © Copyright 2022 by NuScale Power, LLC 122

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 replacement and installation of new instrumentation upgrade of the data acquisition and control system (DACS) and rewiring the instrumentation with DACS modification of portions of integral reactor vessel The updated NIST-1 facility provides a well-scaled representation of a NuScale reactor design that minimizes distortions and provides the measurements necessary for safety code and reactor design validation. A schematic of the NIST-1 facility is shown in Figure 5-3. Figure 5-3 Schematic of NIST-1 integral test facility The configuration of the NIST-1 facility is shown in Figure 5-4. © Copyright 2022 by NuScale Power, LLC 123

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-4 NIST-1 test facility configuration ((

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The NIST-1 facility models the NPM at 1:3.3 length scale, 1:227.5 volume scale, and 1:1 time scale. There are three vessels in the NIST-1 facility: the RPV, CNV, and CPV as shown in Figure 5-3. A photograph of the NIST-1 facility is shown in Figure 5-5. Unlike an actual NuScale plant, the RPV and CNV are not concentric and the CNV is not immersed in the cooling pool. Rather the RPV and CNV are connected by piping that contains valves that perform the functions of the RRVs, RVVs and breaks as shown in Figure 5-3. The CNV is connected to the CPV through an HTP that is scaled to allow energy transfer to the pool in the same proportion as in an NPM. Natural circulation flow in the primary circuit is driven by heat input in the core region and heat removal to the SG tubes. Fluid heated in the core region flows upward through the RCS hot leg riser, and then downward around the outside of the SG tubes, the cold leg and the downcomer. The flow then returns to the core through the lower plenum. The core is comprised of a (( }}2(a),(c),ECI electric heater rod bundle with a maximum power of (( }}2(a),(c),ECI a power level scaled to simulate decay heat. System pressure is controlled by the pressurizer component, which contains heater rods to bring the pressurizer fluid up to saturation temperature. © Copyright 2022 by NuScale Power, LLC 125

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-5 Photograph of the NIST-1 facility © Copyright 2022 by NuScale Power, LLC 126

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Reactor Pressure Vessel Major internal components in the RPV are the core, RCS hot leg riser, pressurizer, and SG bundle. Figure 5-6 shows a view of the RPV thermal-hydraulic regions with the pressurizer at the top, separated from the lower part of the RPV by a perforated pressurizer baffle plate. The upper plenum occupies the region below the pressurizer baffle plate and above the RCS hot leg riser that extends down to the top of the core. The upper annulus between the RCS hot leg riser and the RPV shell contains the helical coil SG tubes and is labeled as the SG region. The lower part of the annulus immediately below the SG tubes is the cold leg. The lower annulus at the core elevation is the downcomer, which is separated from the core by the core shell. The lower plenum occupies the bottom of the RPV and hydraulically connects the downcomer and the core. The RPV shells and flanges are covered by (( }}2(a),(c),ECI © Copyright 2022 by NuScale Power, LLC 127

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-6 Reactor pressure vessel thermal-hydraulic regions © Copyright 2022 by NuScale Power, LLC 128

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The RPV houses the core, which is modeled by a ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-7 Lower core flow plate ((

                                                                                                      }}2(a),(c),ECI Reactor Coolant System Hot Leg Riser After leaving the core, the flow enters the chimney of the RCS hot leg riser. The RCS hot leg riser extends from above the core shroud to the upper plenum, creating a riser and downcomer configuration to enable natural circulation. As shown in Figure 5-6, the RCS hot leg riser consists of a lower shell, a conical transition, a middle shell containing the flowmeter for the primary circuit, and an upper shell. Flow exits the riser into the upper plenum, which is the space between the RCS hot leg riser outlet and the bottom of the pressurizer baffle plate.

Upper Plenum After leaving the top of the RCS hot leg riser, the flow enters the upper plenum and is directed radially outward to flow down in the annulus between the riser and the RPV shell. The pressurizer baffle plate separates the upper plenum from the pressurizer. Hydraulic communication between the pressurizer and the RPV occurs via (( }}2(a),(c) holes in the pressurizer baffle plate that are grouped in © Copyright 2022 by NuScale Power, LLC 130

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 eight clusters of three circles radially located (( }}2(a),(c),ECI from the center of the plate. Pressurizer The pressurizer is located above the upper plenum and is in thermal hydraulic communication with the upper plenum via the pressurizer baffle plate holes. The pressurizer maintains primary system static pressure during normal steady-state and transient conditions through the use of three heater elements. Each element has (( }}2(a),(c),ECI of power and is modulated by the facility control system to maintain system static pressure. Cold Leg Downcomer After leaving the upper plenum, the flow continues downward through the SG section and into the cold leg downcomer region. The cold leg downcomer is the annular space bounded by the RPV shell inner diameter and the RCS hot leg riser outer diameter. When fluid reaches the RCS hot leg riser conical transition shell, the flow area is reduced. Flow exits the cold leg downcomer into the lower plenum before it recirculates back into the core. Steam Generators The SG is a helical coil, once-through HX consisting of three vertical, parallel banks of tubes (an inner, middle, and outer coil) that wrap counter to each other in the annular space between the RCS hot leg riser and the RPV shell inner surface. In the NIST-1 facility, the primary coolant is circulated around the outside of the SG tubes. Feedwater supplied from the feedwater storage tank is pumped through the SG coils by a regenerative turbine pump. This pump utilizes a variable speed controller, which allows for precise control of the feedwater mass flow rate. Pressure in the secondary side is regulated by a pneumatically operated variable position valve located in the steam line portion of the flow loop. In the feed line, feedwater is pre-heated in an in-line heater before going into the different inlet headers. Feedwater enters the coil bundle at the bottom of the SG and is fully boiled in the tubes, resulting in steam that is superheated at the SG outlet. The boil off length is a function of core power, core exit temperature, main feed pump flow rate, and secondary side pressure. Every coil exhausts the superheated steam into a common steam bustle (header), at which point, depending on the test, piping either directs the steam to atmosphere (i.e., the stack), the CNV, or the DHRS inlet header. Lower Plenum The lower plenum is the region bounded between the tubesheet and the lower core flow plate. The lower plenum provides the connection between the downcomer and the core, thus completing the RPV flow loop. © Copyright 2022 by NuScale Power, LLC 131

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Decay Heat Removal System The NIST-1 facility has three possible configurations with the DHRS. A full-height DHRS is used for separate effects testing. The full-height DHRS has a total of eight tubes, distributed between three rows. Figure 5-8 shows the full-height DHRS. Two decay heat removal HXs are scaled for testing integral DHRS effects in the NIST-1 facility. Because an NPM has two decay heat removal trains available for use, one scaled DHR HX has one tube for simulating one NPM decay heat removal train and the second scaled HDR HX has two tubes representing two NPM decay heat removal trains. Figure 5-9 shows the scaled decay heat removal HXs. Figure 5-8 Full-height decay heat removal condenser ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-9 Scaled decay heat removal heat condensers ((

                                                                                                  }}2(a),(c),ECI Containment and Cooling Pool Vessel The CNV, representing the cavity volume between the RPV outer surface and the containment inside surface, is conjoined to the CPV and thermally separated by a scaled HTP. For an NPM, the RPV is located inside containment. However, with the NIST-1 facility, to maintain both volume and surface area scaling similitude, as well as allow proper instrumentation, the RPV is thermal hydraulically separated from the CNV. The heat transfer plate (HTP) models the scaled condensation heat transfer surface between the CNV and CPV. Fluid in the CPV, which is at ambient pressures, models the scaled volume in which an NPM containment vessel is partially immersed.

The CPV has a set of four ports allowing for the installation of one of three decay heat removal HXs. The baseline configuration is with a full height decay heat removal, which is used for SETs. Two decay heat removal HXs are scaled for testing integral DHRS effects in the NIST-1 facility. Because an NPM has two decay heat removal trains available for use, to simulate either one or both trains, one scaled decay heat removal has one tube for simulating one NPM decay heat removal train and the second scaled decay heat removal has two tubes representing two NPM decay heat removal trains. © Copyright 2022 by NuScale Power, LLC 133

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Emergency Core Cooling System and Chemical and Volume Control System Lines Eight lines connect the RPV to the CNV. Five of these lines belong to the facility ECCS, whereas the other three are part of the CVCS. As part of the ECCS, there are two independent reactor vent lines near the top of the pressurizer section, and two reactor recirculation lines in the lower downcomer of the RPV. The fifth ECCS line is a SET line that also models reactor recirculation. For the CVCS, two lines penetrate the vessel near the bottom of the SG. One of these lines penetrates both the vessel wall and the hot leg riser, simulating the make-up line into the hot leg. The other CVCS line connects to the cold leg and thus penetrates only the RPV wall. This line represents the facility CVCS discharge break line. A third CVCS line between the RPV and CNV is located at the very top of the pressurizer and functions as an analogy for the pressurizer spray line. Each line that is installed has a pneumatic isolation valve that is actuated through the test facility control system. Any lines that are not installed use a blank flange for isolation. The facility baseline configuration consists ((

                                             }}2(a),(c),ECI NIST-1 Facility Instrumentation and Control Instrumentation is used throughout the facility to measure the thermal hydraulic behavior during steady-state and transient operations. The following information is obtained and recorded by the DACS:

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                          }}2(a),(c),ECI The ECCS, CVCS and SET lines are heavily instrumented to obtain data on break flow and ECCS blowdown and recirculation flows. The components and instruments include:

an isolation valve. ((

                                      }}2(a),(c),ECI The facility control system generates signals per the system logic, including valve and relay control signals, heater and pump control signals, etc. The following systems can be regulated by the test facility control system:

core heaters (including decay power modeling) pressurizer heaters feedwater pump coolant charging pump containment heaters (used to maintain an adiabatic boundary condition on all walls of containment except for the prescribed condensation surface of the HTP) strip heat power in the ECCS, CVCS, DHRS, and steam lines (used to minimize heat loss in the lines connecting the RPV and CNV) CNV vacuum pump various pneumatic and solenoid valves for flow regulation or transient initiation The NIST-1 facility is used to assess the operation of an NPM under normal operating conditions and to assess the passive safety system responses during transient conditions. The data generated and collected by the facility DACS is © Copyright 2022 by NuScale Power, LLC 135

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 used to assist in validation of the NRELAP5 system thermal-hydraulic code for NPM analysis. 5.3.2.2 Decay Heat Removal System Separate Effects Test Matrix The NIST-1 facility was used to perform separate effects testing with the full-height DHRS. Table 5-7 describes DHRS separate effects testing performed at the NIST-1 facility; these test assessments are presented in the following subsections. Table 5-7 NIST-1 decay heat removal system separate effects tests for NRELAP5 code validation ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.2.3 NRELAP5 Model Description For the NIST-HP-03 and NIST-HP-04 test assessments, a separate effects model of the NIST-1 facility is used. This model contains hydrodynamic components and heat structures for: ((

                                                                                             }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                              }}2(a),(c)

Table 5-8 Comparison between NPM and NIST-1 decay heat removal NRELAP5 nodalizations ((

                                                                                                  }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-10 NIST-1 noding diagram for full-height decay heat removal system separate effect tests ((

                                                                                              }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.2.4 HP-03 Test Description The NIST-1 HP-03 test (HP-03) was used to assess the capability of NRELAP5 to predict convective condensation within the DHRS and heat transfer to the reactor pool across the DHRS tubes. For HP-03, the heated NIST-1 primary system was used to produce steam within the SG tubes. The steam from the SG tubes was transferred to the steam line and routed to the full-height DHRS steam header, which was located in the upper CPV. Steam was condensed in the condenser tubes, entered the DHRS condensate header, and flowed through the condensate line. The condensate control valve in the condensate line maintained DHRS pressure by controlling the rate at which condensate discharged to the atmosphere. During HP-03, superheated steam was delivered to the DHRS steam header at a range of flow rates and pressures. The incoming steam was allowed to condense within the condenser tubes and a pseudo steady state liquid level (DHRS level) was established. Table 5-9 summarizes the initial conditions of the HP-03 cases. Table 5-9 NIST-1 HP-03 test cases ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.2.5 HP-03 Results The HP-03 data trends were well predicted by NRELAP5 with reasonable to excellent agreement ((

                              }}2(a),(c),ECI Detailed results are presented herein for three HP-03 cases, specifically HP-03-01, HP-03-02c, and HP-03-03-Part1.

For the higher DHRS inlet mass flow rate cases (i.e., HP-03-01c-Part1, Part2, and Part3), NRELAP5 code-to-data comparisons show [A.] reasonable-to-excellent or excellent agreement for DHRS power, [B.] reasonable, reasonable-to-excellent, or excellent agreement for other parameters of interest, and [C.] reasonable agreement for DHRS level. ((

                              }}2(a),(c) 5.3.2.5.1              HP-03-01 Run This section compares the NRELAP5 simulation results with the measured data for HP-03-01, which was run at a DHRS pressure of ((
                                   }}2(a),(c),ECI Figure 5-11 presents the flow enthalpy into the inlet and outlet headers of the DHRS heat exchanger (an alternate view of DHRS heat removal rate, or DHRS power). The mass flow rate, pressure, and temperature at the inlet are specified as boundary conditions. As there is a very large difference in magnitude between the inlet and outlet enthalpy flow rates, the NRELAP5 outlet enthalpy flow rate and data outlet enthalpy flow rate do not need to be perfectly matched in order for NRELAP5 DHRS power to show excellent code-to-data agreement.

Figure 5-11a presents the DHRS power. Code-to-data agreement is excellent. This agreement indicates that NRELAP5 appropriately calculates the DHRS-to-CPV heat transfer. Figure 5-12 presents the NRELAP5 calculated DHRS heat exchanger level. Code-to-data agreement is reasonable. (( }}2(a),(b),(c) © Copyright 2022 by NuScale Power, LLC 141

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                   }}2(a),(b),(c) Figure 5-12a presents the same comparison as Figure 5-12, but with the Reference 29 version of NRELAP5.

Figure 5-13 presents the DHRS condenser tube internal fluid temperature. ((

                               }}2(a),(b),(c) Code-to-data agreement is reasonable. Figure 5-13a presents the same comparison as Figure 5-13, but with the Reference 29 version of NRELAP5.

Figure 5-14 and Figure 5-14a present the CPV temperature. Code-to-data agreement is excellent. While the NRELAP5 simulation is not run for an extended period of time, the CPV temperature is well-matched at the end of the comparison period. The high temperatures in the upper CPV indicate that boiling was occurring. Figure 5-14b presents the same comparison as Figure 5-14 and the bottom elevation of Figure 5-14a, but with the Reference 29 version of NRELAP5. Figure 5-15 presents the CPV level. ((

                                             }}2(a),(c) code-to-data agreement is excellent. The CPV level is observed to drop during testing, which indicates that boiling was occurring.

Figure 5-15a presents the same comparison as Figure 5-15, but with the Reference 29 version of NRELAP5. ((

                                                                                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-11 NIST-1 HP-03-01 decay heat removal system enthalpy flow rate code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-11a NIST-1 HP-03-01 decay heat removal system power code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-12 NIST-1 HP-03-01 decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-12a NIST-1 HP-03-01 decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-13 NIST-1 HP-03-01 decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-13a NIST-1 HP-03-01 decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-14 NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison (1 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-14a NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison (2 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-14b NIST-1 HP-03-01 cooling pool vessel temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-15 NIST-1 HP-03-01 cooling pool vessel level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-15a NIST-1 HP-03-01 cooling pool vessel level code-to-data comparison ((

                                                                                                   }}2(a),(b),(c),ECI 5.3.2.5.2              HP-03-02c Run This section compares the NRELAP5 simulation results with the measured data HP-03-02c, which was run at a DHRS pressure of ((
                                   }}2(a),(c),ECI Figure 5-16 presents the DHRS inlet and outlet enthalpy flow rate. ((
                                                  }}2(a),(b),(c) As there is a very large difference in magnitude between the inlet and outlet enthalpy flow rates, the NRELAP5

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 outlet enthalpy flow rate and data outlet enthalpy flow rate do not need to be perfectly matched in order for NRELAP5 DHRS power to show excellent code-to-data agreement. Figure 5-16a presents the DHRS power. Code-to-data agreement is excellent. This agreement indicates that NRELAP5 appropriately calculates the DHRS-to-CPV heat transfer. Figure 5-17 presents the DHRS level. Code-to-data agreement is reasonable. Figure 5-17a presents the same comparison as Figure 5-17, but with the Reference 29 version of NRELAP5. Figure 5-18 presents the DHRS condenser tube internal fluid temperature. ((

                               }}2(a),(b),(c) Code-to-data agreement is reasonable. Figure 5-18a presents the same comparison as Figure 5-18, but with the Reference 29 version of NRELAP5.

Figure 5-19 and Figure 5-19a present the CPV temperature. Overall code-to-data agreement is judged to be reasonable, as most trends are captured ((

                                                                                       }}2(a),(b),(c) Figure 5-19b presents the same comparison as Figure 5-19 and the bottom elevation of Figure 5-19a, but with the Reference 29 version of NRELAP5.

Figure 5-20 presents the CPV level. ((

                                             }}2(a),(c) code-to-data agreement is excellent. The CPV level is observed to drop during testing, which indicates that boiling was occurring.

Figure 5-20a presents the same comparison as Figure 5-20, but with the Reference 29 version of NRELAP5. © Copyright 2022 by NuScale Power, LLC 154

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-16 NIST-1 HP-03-02c decay heat removal system enthalpy flow rate code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-16a NIST-1 HP-03-02c decay heat removal system power code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-17 NIST-1 HP-03-02c decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-17a NIST-1 HP-03-02c decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-18 NIST-1 HP-03-02c decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-18a NIST-1 HP-03-02c decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-19 NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison (1 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-19a NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison (2 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-19b NIST-1 HP-03-02c cooling pool vessel temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-20 NIST-1 HP-03-02c cooling pool vessel level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-20a NIST-1 HP-03-02c cooling pool vessel level code-to-data comparison ((

                                                                                                  }}2(a),(b),(c),ECI 5.3.2.5.3              HP-03-03 Part 1 Run This section compares the NRELAP5 simulation results with the measured data for the HP-03-03-Part 1 test, which was run at a DHRS pressure of

(( }}2(a),(c),ECI Figure 5-21 presents the DHRS inlet and outlet enthalpy flow rate. Note that the inlet temperature, pressure, and mass flow rate are specified as boundary conditions. ((

                                                 }}2(a),(b),(c) As there is a very large difference in magnitude between the inlet and outlet enthalpy flow rates, the NRELAP5

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 outlet enthalpy flow rate and data outlet enthalpy flow rate do not need to be perfectly matched in order for NRELAP5 DHRS power to show excellent code-to-data agreement. Figure 5-21a presents the DHRS power. Code-to-data agreement is reasonable-to-excellent. This agreement indicates that NRELAP5 appropriately calculates the DHRS-to-CPV heat transfer. ((

                                                                         }}2(a),(c)

Figure 5-22 presents the DHRS level. Code-to-data agreement is reasonable-to-excellent. Figure 5-22a presents the same comparison as Figure 5-22, but with the Reference 29 version of NRELAP5. Figure 5-23 presents the DHRS condenser tube internal fluid temperature. ((

                             . }}2(a),(b),(c) Code-to-data agreement is reasonable. Figure 5-23a presents the same comparison as Figure 5-23, but with the Reference 29 version of NRELAP5.

Figure 5-24 and Figure 5-24a present the CPV temperature. Overall code-to-data agreement is reasonable-to-excellent. Most trends that appear in the test data are captured by NRELAP5. Figure 5-24b presents the same comparison as Figure 5-24 and the bottom elevation of Figure 5-24a, but with the Reference 29 version of NRELAP5. Figure 5-25 presents the CPV level. ((

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

Figure 5-25a presents the same comparison as Figure 5-25, but with the Reference 29 version of NRELAP5. © Copyright 2022 by NuScale Power, LLC 166

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-21 NIST-1 HP-03-03-P1 decay heat removal system enthalpy flow rate code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-21a NIST-1 HP-03-03-Part1 decay heat removal system power code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-22 NIST-1 HP-03-03-Part1 decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-22a NIST-1 HP-03-03-Part1 decay heat removal system level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-23 NIST-1 HP-03-03-Part1 decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-23a NIST-1 HP-03-03-Part1 decay heat removal system internal fluid temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-24 NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison (1 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-24a NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison (2 of 2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-24b NIST-1 HP-03-03-Part1 cooling pool vessel temperature code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-25 NIST-1 HP-03-03-Part1 cooling pool vessel level code-to-data comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-25a NIST-1 HP-03-03-Part1 cooling pool vessel level code-to-data comparison ((

                                                                                              }}2(a),(b),(c),ECI 5.3.2.5.4              HP-03 Summary The comparisons between the HP-03 data and HP-03 NRELAP5 simulations show that NRELAP5 has the capability to accurately predict the DHRS-to-CPV heat transfer.

For DHRS power, code-to-data comparisons show reasonable-to-excellent or excellent agreement. For the identified additional parameters of interest, including DHRS level and CPV level, code-to-data comparisons show reasonable, reasonable-to-excellent, or excellent agreement. The re-performance of the comparisons using the Reference 29 version of NRELAP5 shows the same level of agreement as for the original comparisons. © Copyright 2022 by NuScale Power, LLC 177

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.2.6 HP-04 Test Description HP-04 was used to assess NRELAP5 capability to predict ((

                                                           }}2(a),(c)

For the HP-04 test, the heated primary system was used to provide transfer heat to the SG and create steam. The steam from the SG tubes was transferred to the steam line and routed to the full-height DHRS steam header, located in the CPV. Steam is condensed in the condenser tubes and enters the DHRS condensate plenum and condensate line. For this test the condensate line directed the DHRS discharge to the environment. ((

                                                                      }}2(a),(c) These conditions were achieved at two pressure conditions. Table 5-10 presents the test conditions considered for the assessment of NRELAP5.

Table 5-10 NIST-1 HP04 test ranges ((

                                                                                                     }}2(a),(c),ECI 5.3.2.7            HP-04 Test Results The NIST-1 HP-04 test data was compared to NRELAP5 predictions designed to simulate the test conditions and test procedures in effect during the experiment.

HP-04 test data trends were well predicted by NRELAP5 with reasonable to excellent agreement for DHRS heat exchanger ((

                                                                             }}2(a),(c),ECI based on an enthalpy balance across the tubes.

The comparison of the calculated DHRS and cooling pool water levels show reasonable to excellent agreement with the data. The NIST-1 CPV heat-up response was not fully captured in the code to data comparison. The comparisons between the experimental data and the code calculated values show that NRELAP5 has the capability to accurately predict the energy transfer across the DHRS heat exchanger tubes to the CPV fluid, although the CPV © Copyright 2022 by NuScale Power, LLC 178

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 heat-up response is not fully captured. The heat transfer across the tubes is influenced by the condensate level inside the tubes and the cooling pool level. These results are discussed in more detail in the following subsections for NIST-1 generated data at DHRS pressures of (( }}2(a),(c),ECI 5.3.2.7.1 HP-04-02 Run This section compares the NRELAP5 simulation results with the measured data for the HP-04-02 run (( }}2(a),(c),ECI In figures, the data measurement uncertainty is shown in dotted lines. Code-to-data comparisons of the DHRS heat removal rate produced reasonable agreement, as shown by the prediction of the enthalpy change over the DHRS in Figure 5-26. ((

                                                           }}2(a),(c) Figure 5-26a presents the same comparison as Figure 5-26, but with the Reference 29 version of NRELAP5.

The NIST-1 profiles for the DHRS condensate outlet temperature are not fully captured in the code-to-data comparisons. ((

                                                                         }}2(a),(c) Figure 5-27a presents the same comparison as Figure 5-27, but with the Reference 29 version of NRELAP5.

Code-to-data comparisons of DHRS heat exchange tube level produced reasonable agreement, as shown in Figure 5-28. ((

                                           }}2(a),(c) Figure 5-28a presents the same comparison as Figure 5-28, but with the Reference 29 version of NRELAP5.

Although the code-to-data comparisons of CPV level produced reasonable-to-excellent agreement, ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                         }}2(a),(b),(c),ECI Figure 5-29a, Figure 5-30a, and Figure 5-31a present the same comparisons as Figure 5-29, Figure 5-30, and Figure 5-31, respectively, but with the Reference 29 version of NRELAP5.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-26 NIST-1 HP-04-02 decay heat removal system energy transfer rate ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-26a NIST-1 HP-04-02 decay heat removal system energy transfer rate

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-27 NIST-1 HP-04-02 decay heat removal system condensate temperature ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-27a NIST-1 HP-04-02 decay heat removal system condensate temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-28 NIST-1 HP-04-02 decay heat removal system internal collapsed level ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-28a NIST-1 HP-04-02 decay heat removal system internal collapsed level

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-29 NIST-1 HP-04-02 cooling pool vessel level ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-29a NIST-1 HP-04-02 cooling pool vessel level

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-30 NIST-1 HP-04-02 mid cooling pool vessel fluid temperatures ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-30a NIST-1 HP-04-02 mid cooling pool vessel fluid temperatures

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-31 NIST-1 HP-04-02 upper cooling pool vessel fluid temperatures ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-31a NIST-1 HP-04-02 upper cooling pool vessel fluid temperatures

((

                                                                                           }}2(a),(b),(c),ECI 5.3.2.7.2              HP-04-03 Run This section compares the NRELAP5 simulation results with the measured data for the HP-04-03 run ((                }}2(a),(c),ECI In figures, the data measurement uncertainty is shown in dotted lines.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Code-to-data comparison of the DHRS heat removal rate produced reasonable to excellent agreement, as shown by the prediction of the enthalpy change over the DHRS in Figure 5-32. ((

                                                                      }}2(a),(c) Figure 5-32a presents the same comparison as Figure 5-32, but with the Reference 29 version of NRELAP5.

The NIST-1 profiles for DHRS condensate outlet temperatures are not fully captured in the code-to-data comparisons. ((

                                                   }}2(a),(c) Figure 5-33a presents the same comparison as Figure 5-33, but with the Reference 29 version of NRELAP5.

Code-to-data comparison of DHRS heat exchanger tube level shows reasonable agreement, as shown in Figure 5-34. ((

                                          }}2(a),(c) Figure 5-34a presents the same comparison as Figure 5-34, but with the Reference 29 version of NRELAP5.

Although the code-to-data comparisons of CPV level produced reasonable-to-excellent agreement, ((

                                                                                }}2(a),(b),(c),ECI Figure 5-35a, Figure 5-36a, and Figure 5-37a present the same comparisons as Figure 5-35, Figure 5-36, and Figure 5-37, respectively, but with the Reference 29 version of NRELAP5.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-32 NIST-1 HP-04-03 decay heat removal system energy transfer rate ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-32a NIST-1 HP-04-03 decay heat removal system energy transfer rate

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-33 NIST-1 HP-04-03 decay heat removal system condensate temperature ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-33a NIST-1 HP-04-03 decay heat removal system condensate temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-34 NIST-1 HP-04-03 decay heat removal system internal collapsed level comparison ((

                                                                                    }}2(a),(b),(c),ECI Figure 5-34a NIST-1 HP-04-03 decay heat removal system internal collapsed level comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-35 NIST-1 HP-04-03 cooling pool vessel level comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-35a NIST-1 HP-04-03 cooling pool vessel level comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-36 NIST-1 HP-04-03 mid cooling pool vessel axial temperatures ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-36a NIST-1 HP-04-03 mid cooling pool vessel axial temperatures

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-37 NIST-1 HP-04-03 upper cooling pool vessel axial temperatures ((

                                                                                              }}2(a),(b),(c),ECI Figure 5-37a NIST-1 HP-04-03 upper cooling pool vessel axial temperatures

((

                                                                                              }}2(a),(b),(c),ECI 5.3.2.7.3              HP-04 Summary Although the CPV fluid heat-up profiles for the NIST-1 HP-04 test data were not fully captured in the NRELAP5 simulations of HP-04, the code-to-data comparisons of DHRS heat removal rate were well-matched. The re-performance of the comparisons using the Reference 29 version of

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 NRELAP5 shows the same level of agreement as for the original comparisons. This assessment concludes that NRELAP5 simulations of an NPM decay heat removal system should be expected to adequately predict the DHRS heat removal rates related to a large envelope of reactor pool liquid conditions. 5.3.3 NIST-1 Non-LOCA Integral Test 5.3.3.1 NIST-1 Facility The NIST-1 facility is described in Section 5.3.2.1. The NLT-2 tests were conducted in two sessions; first NLT-2a was conducted and a few weeks later NLT-2b was conducted. Test NLT-02a was a loss of feed water transient with a subsequent pressurization and depressurization of the RPV. In Test NLT-02b the DHRS was activated from a low reactor power level to demonstrate long term cooling capability over several hours. 5.3.3.2 Non-LOCA Integral Effects Test Matrix The NIST-1 facility was used to perform integral effects testing with the scaled two-tube DHRS heat exchanger. Table 5-11 describes DHRS integral effects testing performed at the NIST-1 facility; these test assessments are presented in the following subsections. Table 5-11 NIST-1 integral effects tests for NRELAP5 code validation ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-11 NIST-1 integral effects tests for NRELAP5 code validation (Continued) ((

                                                                                                       }}2(a),(c),ECI 5.3.3.3            NRELAP5 Model Description The NRELAP5 model of the NIST-1 facility is constructed to include the major flow paths in the facility, including the major systems; the RPV primary and secondary, the CNV, and the CPV, as well as the ECCS. ((
                            }}2(a),(c) A nodalization diagram is shown in Figure 5-38.

((

                                                                                                          }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                             }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-38 NRELAP5 noding diagram for the NIST-1 facility ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-39 NRELAP5 NIST-1 model secondary system nodalization layout (NLT-2b) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-39a NRELAP5 NIST-1 model secondary system nodalization layout (NLT-15p2) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-40 NIST-1 feedwater split headers at the steam generator tube coils connection (at the time of NLT-2b test) 5.3.3.4 NLT-2a Test Description The objective of the NLT-02 test was to measure the scaled integral system response to a decrease in heat removal from the secondary side following a loss of feedwater event, to the point of reactor trip. In an NPM design, response to a decrease in heat removal includes a reactor trip based on PZR pressure, core outlet temperature, or PZR level, as well as the activation of the DHRS heat exchanger, which permits circulation between the DHRS and the SG coils. ((

                                      }}2(a),(c)

The NIST-1 facility had the following configuration and initial conditions at the start of the NLT-02a test:

1. ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

5. ((

                                     }}2(a),(b),(c),ECI The sequence of events for the NLT-02a test is shown in Table 5-12.

Table 5-12 NLT-02a sequence of events ((

                                                                                                        }}2(a),(b),(c),ECI 5.3.3.5            NLT-2a Test Results The transient was initiated with the termination of feedwater flow. Feedwater flow was a boundary condition in the NRELAP5 simulation. Figure 5-41 compares the calculated feedwater flow with the data. Note that in figures, the data measurement uncertainty is shown in dotted lines.

The core heater rod power was another boundary condition applied to the simulation. Figure 5-42 compares core heater rod power used in the simulation with the measured core heater rod power that is supplied to the core heater rods. Figure 5-43 compares the PZR pressure. The calculated pressure shows reasonable to excellent agreement with the measured data. ((

                                                  }}2(a),(c) Figure 5-43a presents the same comparison as Figure 5-43, but with the Reference 29 version of NRELAP5.

Figure 5-44 compares the calculated riser flow rate with the measured data. The predicted response shows reasonable agreement with the data. ((

                                              }}2(a),(c) Figure 5-44a presents the same comparison as Figure 5-44, but with the Reference 29 version of NRELAP5.

The comparison of the measured and calculated PZR level is shown in Figure 5-45. The predicted level is considered in reasonable to excellent © Copyright 2022 by NuScale Power, LLC 202

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 agreement with the data; the calculation is within the uncertainty bands of the data ((

                                   }}2(a),(c) Figure 5-45a presents the same comparison as Figure 5-45, but with the Reference 29 version of NRELAP5.

The predicted core inlet temperature is compared to data in Figure 5-46. ((

                                                            }}2(a),(c) Figure 5-46a presents the same comparison as Figure 5-46, but with the Reference 29 version of NRELAP5.

The core exit temperature comparison is shown in Figure 5-49. The initial temperature is underpredicted in the steady state but is within the uncertainty band when the core power is shut off. The trend of the predicted temperature is slightly different than the measured value between 50-125 seconds, but it remains within the uncertainty bands of the data. Figure 5-49a presents the same comparison as Figure 5-49, but with the Reference 29 version of NRELAP5. The predicted riser inlet fluid temperature, shown in Figure 5-49b, is in reasonable agreement with the measured temperature. A comparison of the calculated and measured PZR heater rod power is shown in Figure 5-50. The PZR heater rods are used to keep the RPV primary side at a target pressure. If the pressure declines, the PZR heater rod power increases. Conversely, if the PZR pressure increases, the PZR heater rod power declines. ((

                                                                                    }}2(a),(c)

The code-to-data comparison of the steam line pressure is provided in Figure 5-51. ((

                                                            }}2(a),(c)

A comparison of the steam line mass flow rate is shown in Figure 5-52. The predicted steam line mass flow rate shows reasonable agreement with the data, considering ((

                                                               }}2(a),(c). Figure 5-52a presents the same comparison as Figure 5-52, but with the Reference 29 version of NRELAP5.

In conclusion, it is demonstrated that NRELAP5 is capable of predicting the data trends of non-LOCA events, such as a loss of feedwater, with a high degree of confidence. © Copyright 2022 by NuScale Power, LLC 203

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-41 NLT-02a transient feedwater flow comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-42 NLT-02a transient core heater rod power comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-43 NLT-02a transient pressurizer pressure comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-43a NLT-02a transient pressurizer pressure comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-44 NLT-02a transient riser mass flow rate comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-44a NLT-02a transient riser mass flow rate comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-45 NLT-02a transient pressurizer level comparison ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-45a NLT-02a transient pressurizer level comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-46 NLT-02a transient core inlet temperature ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-46a NLT-02a transient core inlet temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-47 NLT-02a transient combined middle and outer steam generator tube coil differential pressure comparison ((

                                                                                               }}2(a),(b),(c),ECI Figure 5-48 NLT-02a transient inner steam generator tube coil differential pressure comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-49 NLT-02a transient core exit fluid temperature comparison ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-49a NLT-02a transient core exit fluid temperature comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-49b NLT-02a transient riser inlet fluid temperature comparison ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-50 NLT-02a transient pressurizer heater rod power comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-51 NLT-02a transient steam line pressure comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-52 NLT-02a transient steam line mass flow rate comparison ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-52a NLT-02a transient steam line mass flow rate comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.3.6 NLT-2b Test Description The objectives of the NLT-02b test were to observe the integral response of the DHRS and RPV after initial DHRS activation, under quasi-steady conditions as the cooling pool was allowed to heat-up, and finally in a period of DHRS-driven cooling and depressurization after the core power transitioned to decay heat mode. The principal parameters of interest during initial DHRS activation are DHRS condensate mass flow rate, DHRS pressure, and temperatures throughout the SG secondary and DHRS loop. After the initial pressure peak due to DHRS activation subsides, additional parameters of interest are the primary flow rate, primary level, primary pressure and temperatures, and CPV temperatures adjacent to the DHRS heat exchanger. ((

                                                      }}2(a),(c)

The NIST-1 facility had the following configuration and initial conditions at the start of the NLT-02b test:

1. ((

                                                                       }}2(a),(c)

The sequence of events for the NLT-02b test is shown in Table 5-13. ((

                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-13 NLT-02b sequence of events ((

                                                                                                    }}2(a),(b),(c) 5.3.3.7            NLT-2b Phase 1 Test Results Code-to-data comparisons of key parameters for the first phase of the NLT-2b test are shown in Figure 5-53 through Figure 5-70. Note that in figures, the data measurement uncertainty is shown in dotted lines. This phase consisted of termination of the feedwater flow and turning off the core heater rod power prior to the opening of the DHRS steam line and condensate line valves to begin the transition to DHRS cooling. ((
                                           }}2(a),(b),(c)

In general, the NRELAP5 simulation showed reasonable agreement with the measured data for this time period. The measured core heater rod power is used as a boundary condition in the simulation. A comparison of the core heater rod power is shown in Figure 5-53. ((

                                }}2(a),(c)

Comparison of the PZR pressure response is provided in Figure 5-54. ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                          }}2(a),(b),(c),ECI Figure 5-55a presents the same comparison as Figure 5-55, but with the Reference 29 version of NRELAP5.

Code-to-data comparisons for the core inlet and exit temperatures are shown in Figure 5-57. ((

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

Comparison of the measured and calculated steam line pressure is shown in Figure 5-59. ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

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

The initial liquid mass inventory indicated by the steady state SG tube level and DHRS HX tube level showed reasonable agreement with the data in the simulation. ((

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

The feedwater mass flow rate was terminated prior to the opening of the DHRS loop steam and condensate valves. © Copyright 2022 by NuScale Power, LLC 217

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-63 and Figure 5-64 show code-to-data comparisons of the SG and DHRS heat exchanger tube liquid level, respectively. ((

                                                                        }}2(a),(c) Figure 5-64a presents the same comparison as Figure 5-64, but with the Reference 29 version of NRELAP5.

Figure 5-65 provides a comparison of the measured and calculated DHRS condensate fluid temperature. Although the trends of the data are calculated, the comparison is considered minimal, ((

                                                           }}2(a),(c) Figure 5-65a presents the same comparison as Figure 5-65, but with data from a different location and using the Reference 29 version of NRELAP5.

Figure 5-66 shows that DHRS condensate flow is slightly underpredicted. The predicted flow rate starts out slightly higher than the data, but the trend is correct, and the magnitude is reasonable compared to the data during the quasi steady state phase. The flow plot in combination with the DHRS heat exchanger level (Figure 5-64) and SG level (Figure 5-63), with inspection of the pressure drops in the steam lines and condensate lines, provides confidence that the total DHRS loop resistance is appropriately accounted for. Figure 5-66a presents the same comparison as Figure 5-66, but with the Reference 29 version of NRELAP5. Figure 5-67 shows ((

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

Figure 5-68 compares predicted CPV level during phase 1 of the transient. ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

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

Figure 5-53 NLT-02b phase 1 transient core power comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-54 NLT-02b phase 1 transient pressurizer pressure comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-54a NLT-02b phase 1 transient pressurizer pressure comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-55 NLT-02b phase 1 transient pressurizer level comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-55a NLT-02b phase 1 transient pressurizer level comparison

((

                                                                                         }}2(a),(b),(c),ECI Figure 5-56 Not Used

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-57 NLT-02b phase 1 transient core inlet and outlet temperature comparison ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-57a NLT-02b phase 1 transient core inlet and outlet temperature comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-58 Not Used Figure 5-59 NLT-02b phase 1 transient steam generator steam pressure comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-59a NLT-02b phase 1 transient steam generator steam pressure comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-60 NLT-02b phase 1 transient steam generator thermal power comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-61 NLT-02b phase 1 transient decay heat removal system heat exchanger thermal power comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-62 NLT-02b phase 1 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-62a NLT-02b phase 1 integrated compensation flow for SG and DHRS HX level equalization

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-63 NLT-02b phase 1 transient steam generator level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-64 NLT-02b phase 1 transient decay heat removal system heat exchanger level comparison ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-64a NLT-02b phase 1 transient decay heat removal system heat exchanger level comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-65 NLT-02b phase 1 transient decay heat removal system condensate temperature comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-65a NLT-02b phase 1 transient decay heat removal system condensate temperature comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-66 NLT-02b phase 1 transient decay heat removal system condensate flow comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-66a NLT-02b phase 1 transient decay heat removal system condensate flow comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-67 NLT-02b phase 1 transient steam generator outlet temperature comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-68 NLT-02b phase 1 transient cooling pool vessel level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 230

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-69 NLT-02b phase 1 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-69a NLT-02b phase 1 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger)

((

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

© Copyright 2022 by NuScale Power, LLC 231

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-69b NLT-02b phase 1 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-70 NLT-02b phase 1 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region)

((

                                                                                           }}2(a),(b),(c),ECI
 © Copyright 2022 by NuScale Power, LLC 232

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.3.8 NLT-2b Phase 2 Test Results Code-to-data comparisons of key parameters for the second phase of the NLT-2b test are shown in Figure 5-71 through Figure 5-88. Note that in figures, the data measurement uncertainty is shown in dotted lines. ((

                                                     }}2(a),(c)

In general, the NRELAP5 simulation showed reasonable to excellent agreement with the measured data for this time period, except for the CPV temperatures adjacent to and above the DHRS heat exchanger. ((

                                       }}2(a),(c)

It is noted that in the simulation, the fluid temperatures at and above the DHRS heat exchanger in the CPV warm up to saturation. ((

                                }}2(a),(c)

Figure 5-71 shows the core power comparison. The core power from the test was input directly into NRELAP5. Figure 5-72 shows that the PZR pressure ((

                                   }}2(a),(c) Figure 5-72a presents the same comparison as Figure 5-72, but with the Reference 29 version of NRELAP5.

Figure 5-73 shows that the PZR level ((

                                                                                                          }}2(a),(c)

Figure 5-73a presents the same comparison as Figure 5-73, but with the Reference 29 version of NRELAP5. Figure 5-75 shows the core inlet and outlet temperature comparisons for phase 2. The agreement is reasonable to excellent. This agreement indicates that the total energy added to the RPV fluid is reasonably calculated by the model. The © Copyright 2022 by NuScale Power, LLC 233

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 reasonably predicted temperature indicates the RPV fluid volume change due to density change should also be reasonably predicted. Figure 5-75a presents the same comparison as Figure 5-75, but with the Reference 29 version of NRELAP5. As discussed in Section 5.3.3.7 for Phase 1, the RPV loop flow rate is not compared. Figure 5-77 compares the SG pressure data to the NRELAP5 prediction. The SG pressure ((

                                                                                            . }}2(a),(c)

Figure 5-77a presents the same comparison as Figure 5-77, but with the Reference 29 version of NRELAP5. Comparisons of the SG power and the DHRS HX thermal power are shown on Figure 5-78 and Figure 5-79, respectively. The thermal powers for both SG and DHRS track the data reasonably well (( }}2(a),(c) The SG level (Figure 5-80) and DHRS heat exchanger level (Figure 5-85) compare reasonably well to the data. ((

                                       }}2(a),(c) Figure 5-85a presents the same comparison as Figure 5-85, but with the Reference 29 version of NRELAP5.

The DHRS condensate temperature (Figure 5-82) follows the same trends as the data, but similar to phase 1 the code prediction is about 50°F higher than the data. Figure 5-82a presents the same comparison as Figure 5-82, but with data from a different location and using the Reference 29 version of NRELAP5. Figure 5-83 compares the DHRS condensate flow. The DHRS flow ((

                                                                                                }}2(a),(c)

Figure 5-83a presents the same comparison as Figure 5-83, but with the Reference 29 version of NRELAP5. The SG tube exit temperatures are compared in Figure 5-84. Similar to phase 1, (( }}2(a),(c) For reference, the fluid temperature in the steam bustle is also shown in the figure. The steam temperature increases with core power ((

                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-86 compares the measured and calculated CPV liquid level. The comparison shows reasonable agreement when recognizing that the NRELAP5 level increase is due to thermal expansion, where the NRELAP5 liquid temperatures in the upper CPV reach saturation temperature, well above the data temperature (Figure 5-88). The CPV fluid temperatures were reset at the start of the restart run to match the data after stirring the CPV, hence the good level match at the start of phase 2 (Figure 5-87, Figure 5-87a, Figure 5-87b, and Figure 5-88). ((

                                                     }}2(a),(c)

Figure 5-71 NLT-02b phase 2 transient core power comparison ((

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

© Copyright 2022 by NuScale Power, LLC 235

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-72 NLT-02b phase 2 transient pressurizer pressure comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-72a NLT-02b phase 2 transient pressurizer pressure comparison

((

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

© Copyright 2022 by NuScale Power, LLC 236

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-73 NLT-02b phase 2 transient pressurizer level comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-73a NLT-02b phase 2 transient pressurizer level comparison

((

                                                                                         }}2(a),(b),(c),ECI Figure 5-74 Not Used

© Copyright 2022 by NuScale Power, LLC 237

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-75 NLT-02b phase 2 transient core inlet and outlet temperature comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-75a NLT-02b phase 2 transient core inlet and outlet temperature comparison

((

                                                                                         }}2(a),(b),(c),ECI Figure 5-76 Not Used

© Copyright 2022 by NuScale Power, LLC 238

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-77 NLT-02b phase 2 transient steam generator steam pressure comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-77a NLT-02b phase 2 transient steam generator steam pressure comparison

((

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

© Copyright 2022 by NuScale Power, LLC 239

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-78 NLT-02b phase 2 transient steam generator thermal power comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-79 NLT-02b phase 2 transient decay heat removal system heat exchanger thermal power comparison

((

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

© Copyright 2022 by NuScale Power, LLC 240

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-80 NLT-02b phase 2 transient steam generator level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-81 NLT-02b phase 2 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-81a NLT-02b phase 2 integrated compensation flow for SG and DHRS HX level equalization

((

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

© Copyright 2022 by NuScale Power, LLC 242

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-82 NLT-02b phase 2 transient decay heat removal system condensate temperature comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-82a NLT-02b phase 2 transient decay heat removal system condensate temperature comparison

((

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

© Copyright 2022 by NuScale Power, LLC 243

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-83 NLT-02b phase 2 transient decay heat removal system condensate flow comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-83a NLT-02b phase 2 transient decay heat removal system condensate flow comparison

((

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

© Copyright 2022 by NuScale Power, LLC 244

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-84 NLT-02b phase 2 transient steam generator outlet temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-85 NLT-02b phase 2 transient decay heat removal system heat exchanger level comparison ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-85a NLT-02b phase 2 transient decay heat removal system heat exchanger level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 246

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-86 NLT-02b phase 2 transient cooling pool vessel level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-87 NLT-02b phase 2 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-87a NLT-02b phase 2 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger)

((

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

© Copyright 2022 by NuScale Power, LLC 248

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-87b NLT-02b phase 2 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-88 NLT-02b phase 2 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region)

((

                                                                                           }}2(a),(b),(c),ECI
 © Copyright 2022 by NuScale Power, LLC 249

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.3.9 NLT-2b Phase 3 Test Results Code-to-data comparisons of key parameters for the third phase of the NLT-2b test are shown in Figure 5-89 through Figure 5-106. Note that in figures, the data measurement uncertainty is shown in dotted lines. ((

                                                                                      }}2(a),(c)

In general, the NRELAP5 simulation showed reasonable to excellent agreement with the measured data for this time period. A comparison of the SG and DHRS heat exchanger tube levels are shown in Figure 5-98 and Figure 5-103, respectively. The SG level calculated response exhibits excellent agreement with the measured data, and the calculated DHRS HX tube level shows reasonable agreement. During this phase, as seen in Figure 5-99, ((

                                                                              }}2(a),(c) Figure 5-99a shows that there is ((
                                                                                }}2(a),(c) Figure 5-103a presents the same comparison as Figure 5-103, but with the Reference 29 version of NRELAP5.

© Copyright 2022 by NuScale Power, LLC 250

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-96 and Figure 5-97 compare the heat removal in the SG and DHRS heat exchanger tubes, respectively. ((

                                             }}2(a),(c) Figure 5-95 shows that the SG pressure is reasonably predicted in this phase. Figure 5-95a presents the same comparison as Figure 5-95, but with the Reference 29 version of NRELAP5.

The PZR pressure is compared in Figure 5-90. ((

                                                                             }}2(a),(c) Figure 5-90a presents the same comparison as Figure 5-90, but with the Reference 29 version of NRELAP5.

The calculated PZR level is compared to data in Figure 5-91. ((

                                                                                                      }}2(a),(c)

Figure 5-91a presents the same comparison as Figure 5-91, but with the Reference 29 version of NRELAP5. The core inlet and outlet temperatures are compared in Figure 5-93. Reasonable agreement is observed in these figures, ((

                                                                                  }}2(a),(c) Figure 5-93a presents the same comparison as Figure 5-93, but with the Reference 29 version of NRELAP5.

As discussed in Section 5.3.3.7 for Phase 1, the RPV loop flow rate is not compared. Figure 5-101 compares the measured and calculated DHRS condensate flow. ((

                                                          }}2(a),(c) Figure 5-101a presents the same comparison as Figure 5-101, but with the Reference 29 version of NRELAP5.

SG tube exit temperatures are compared in Figure 5-102. The predicted temperature is ((

                                                                                           }}2(a),(c)

The condensate fluid temperature is compared in Figure 5-100. ((

                                                                                                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                                      }}2(a),(c)

Figure 5-104 compares the measured and calculated CPV liquid level. The comparison shows reasonable agreement with the data out to the time of the drain and refill of the CPV ((

                                 }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 252

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-89 NLT-02b phase 3 transient core power comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-90 NLT-02b phase 3 transient pressurizer pressure comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-90a NLT-02b phase 3 transient pressurizer pressure comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-91 NLT-02b phase 3 transient pressurizer level comparison ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-91a NLT-02b phase 3 transient pressurizer level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 255

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-92 Not Used Figure 5-93 NLT-02b phase 3 transient core inlet and outlet temperature comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-93a NLT-02b phase 3 transient core inlet and outlet temperature comparison

((

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

© Copyright 2022 by NuScale Power, LLC 256

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-94 Not Used Figure 5-95 NLT-02b phase 3 transient steam generator steam pressure comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-95a NLT-02b phase 3 transient steam generator steam pressure comparison

((

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

© Copyright 2022 by NuScale Power, LLC 257

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-96 NLT-02b phase 3 transient steam generator thermal power comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-97 NLT-02b phase 3 transient decay heat removal system heat exchanger thermal power comparison

((

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

© Copyright 2022 by NuScale Power, LLC 258

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-98 NLT-02b phase 3 transient steam generator level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-99 NLT-02b phase 3 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-99a NLT-02b phase 3 integrated compensation flow for SG and DHRS HX level equalization

((

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

© Copyright 2022 by NuScale Power, LLC 260

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-100 NLT-02b phase 3 transient decay heat removal system condensate temperature comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-100a NLT-02b phase 3 transient decay heat removal system condensate temperature comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-101 NLT-02b phase 3 transient decay heat removal system condensate flow comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-101a NLT-02b phase 3 transient decay heat removal system condensate flow comparison

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-102 NLT-02b phase 3 transient steam generator outlet temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-103 NLT-02b phase 3 transient decay heat removal system heat exchanger level comparison ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-103a NLT-02b phase 3 transient decay heat removal system heat exchanger level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 264

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-104 NLT-02b phase 3 transient cooling pool vessel level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-105 NLT-02b phase 3 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger) ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-105a NLT-02b phase 3 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger)

((

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

© Copyright 2022 by NuScale Power, LLC 266

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-105b NLT-02b phase 3 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger) ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-106 NLT-02b phase 3 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region)

((

                                                                                            }}2(a),(b),(c),ECI
  © Copyright 2022 by NuScale Power, LLC 267

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.3.10 NLT-2b Phase 4 Test Results Code-to-data comparisons of key parameters for the fourth phase of the NLT-2b test are shown in Figure 5-107 through Figure 5-125. Note that in figures, the data measurement uncertainty is shown in dotted lines. ((

                                      }}2(a),(c) the fluid volume had shrunk sufficiently to uncover the top of the riser pipe (located at the bottom of the upper plenum). Consistent with the scope of the short-term non-LOCA transient EM, the results presented here focus on the time after the decay heat mode was actuated until top of the riser uncovered.

The simulation gave reasonable results for this phase out to the point of riser uncovery. Measured and calculated core heater rod power is compared in Figure 5-107. The core heater rod power is a boundary condition in the simulation. Energy continued to be removed through the SG tube coil ((

                                                                }}2(a),(c) The energy removal resulted in a cooldown of the primary system. The cooldown caused a shrinkage of the RPV fluid volume and corresponding decrease in RPV pressure (Figure 5-108).

Figure 5-108a presents the same comparison as Figure 5-108, but with the Reference 29 version of NRELAP5. The level in the PZR declined due to the system cooldown as shown in Figure 5-109. The rate of PZR level decline in the calculation is in reasonable agreement with the measured data. ((

                                                                           }}2(a),(c) Figure 5-109a presents the same comparison as Figure 5-109, but with the Reference 29 version of NRELAP5.

As discussed in Section 5.3.3.7 for Phase 1, the RPV loop flow rate is not compared. © Copyright 2022 by NuScale Power, LLC 268

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 A comparison of the measured and calculated upper plenum fluid temperature and the core inlet and outlet temperatures are provided in Figure 5-111 and Figure 5-112, respectively. ((

                                      }}2(a),(c) Figure 5-112a presents the same comparison as Figure 5-112, but with the Reference 29 version of NRELAP5.

The SG pressure is compared in Figure 5-114. ((

                                                                                }}2(a),(c) Figure 5-114a presents the same comparison as Figure 5-114, but with the Reference 29 version of NRELAP5.

Figure 5-115 and Figure 5-116 show that both the SG and DHRS power ((

                                         }}2(a),(c)

Figure 5-118 and Figure 5-118a show ((

                                                                                                           }}2(a),(c)

Figure 5-117 and Figure 5-122 show comparisons on the SG tube level and DHRS HX tube level, respectively. ((

                                                      }}2(a),(c) Figure 5-122a presents the same comparison as Figure 5-122, but with the Reference 29 version of NRELAP5.

Figure 5-119 compares the measured and calculated DHRS condensate temperature. ((

                          }}2(a),(c) Figure 5-119a presents the same comparison as Figure 5-119, but with data from a different location and using the Reference 29 version of NRELAP5.

Figure 5-120 compares the measured and calculated DHRS condensate flow. Figure 5-120a presents the same comparison as Figure 5-120, but with the Reference 29 version of NRELAP5. A comparison of the SG outlet temperature is provided in Figure 5-121. The calculated outlet temperature shows ((

                                                                         }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 269

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The measured and calculated CPV level comparison is shown in Figure 5-123. After the drain and fill process, ((

                                }}2(a),(c), the calculated level shows reasonable agreement with the data.

Measured and calculated CPV fluid temperatures within the elevation of the DHRS heat exchanger are shown in Figure 5-124, Figure 5-124a, Figure 5-124b, and Figure 5-125. ((

                                                    }}2(a),(c)

Figure 5-107 NLT-02b phase 4 transient core power comparison ((

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

© Copyright 2022 by NuScale Power, LLC 270

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-108 NLT-02b phase 4 transient pressurizer pressure comparison ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-108a NLT-02b phase 4 transient pressurizer pressure comparison

((

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

© Copyright 2022 by NuScale Power, LLC 271

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-109 NLT-02b phase 4 transient pressurizer level comparison ((

                                                                                        }}2(a),(b),(c),ECI Figure 5-109a NLT-02b phase 4 transient pressurizer level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 272

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-110 NLT-02b phase 4 transient reactor pressure vessel level ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-111 NLT-02b phase 4 transient reactor pressure vessel upper plenum temperature comparison

((

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

© Copyright 2022 by NuScale Power, LLC 273

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-112 NLT-02b phase 4 transient core inlet and outlet temperature comparison ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-112a NLT-02b phase 4 transient core inlet and outlet temperature comparison

((

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

© Copyright 2022 by NuScale Power, LLC 274

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-113 Not Used Figure 5-114 NLT-02b phase 4 transient steam generator steam pressure comparison ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-114a NLT-02b phase 4 transient steam generator steam pressure comparison

((

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

© Copyright 2022 by NuScale Power, LLC 275

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-115 NLT-02b phase 4 transient steam generator thermal power comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-116 NLT-02b phase 4 transient decay heat removal system heat exchanger thermal power comparison

((

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

© Copyright 2022 by NuScale Power, LLC 276

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-117 NLT-02b phase 4 transient steam generator level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-118 NLT-02b phase 4 calculated compensation flow for steam generator and decay heat removal system heat exchanger level equalization ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-118a NLT-02b phase 4 integrated compensation flow for SG and DHRS HX level equalization

((

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

© Copyright 2022 by NuScale Power, LLC 278

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-119 NLT-02b phase 4 transient decay heat removal system condensate temperature comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-119a NLT-02b phase 4 transient decay heat removal system condensate temperature comparison

((

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

© Copyright 2022 by NuScale Power, LLC 279

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-120 NLT-02b phase 4 transient decay heat removal system condensate flow comparison ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-120a NLT-02b phase 4 transient decay heat removal system condensate flow comparison

((

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

© Copyright 2022 by NuScale Power, LLC 280

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-121 NLT-02b phase 4 transient steam generator outlet temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-122 NLT-02b phase 4 transient decay heat removal system heat exchanger level comparison ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-122a NLT-02b phase 4 transient decay heat removal system heat exchanger level comparison

((

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

© Copyright 2022 by NuScale Power, LLC 282

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-123 NLT-02b phase 4 transient cooling pool vessel level comparison ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-124 NLT-02b phase 4 transient cooling pool vessel region 4 temperature comparison (below decay heat removal system heat exchanger)

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-124a NLT-02b phase 4 transient cooling pool vessel region 5 temperature comparison (near bottom of decay heat removal system heat exchanger) ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-124b NLT-02b phase 4 transient cooling pool vessel region 6 temperature comparison (near mid-point of decay heat removal system heat exchanger)

((

                                                                                            }}2(a),(b),(c),ECI
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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-125 NLT-02b phase 4 transient cooling pool vessel region 7 temperature comparison (just above the decay heat removal system heat exchanger tube region) ((

                                                                                                }}2(a),(b),(c),ECI 5.3.3.11           NLT-2 Summary The comparisons between the NIST-1 NLT-2a integral experimental data and the code calculated values showed that NRELAP5 has the capability to accurately predict primary heat up and pressurization due to a loss of feedwater.

NLT-02b is a transient test with quasi-steady core heater rod power during an extended DHRS recirculation mode and with core heater rod decay power during a DHRS-driven decay power mode. During the DHRS recirculation mode, the transient response of pressure, level, and temperature in both the RPV and the DHRS were predicted by NRELAP5 with reasonable agreement. NRELAP5 assessment of the NLT-2b phases 1 through 4 demonstrates the ability of the code to predict the heat transfer from primary side to the SG and from the DHRS to CPV. After completion of core power maneuvering and after full start of the DHRS loop, NRELAP5 calculated pressurizer pressure, core inlet and outlet temperature, primary flow rate, SG pressure, energy transfer from the primary to secondary, energy transfer from DHRS to CPV, SG level, DHRS level, and CPV level results are all within reasonable to excellent agreement. The pressurizer level for NLT-2b phase 1 was ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                              . }}2(a),(b),(c) However, even with these discrepancies, the results show that the important parameters of total energy transfer from the primary side to the SG and from the DHRS to CPV are well predicted.

For NLT-2b phases 1 through 4, the calculated condensate line temperature is in minimal agreement with data, with the temperature being over predicted by NRELAP5. However, even with the discrepancy, the important parameter of heat transfer from primary side to the SG and from the DHRS to CPV are well predicted. For NLT-2b phases 1 through 4, the CPV fluid heat-up profiles were not fully captured in the NRELAP5 simulations. ((

                                  }}2(a),(c)

For NLT-2b phases 1 through 4, code-to-data comparisons show converged reasonable agreement of DHRS heat removal rate, which is considered to be the most important parameter in this assessment, demonstrating that NRELAP5 is capable of capturing total energy removal rate of the DHRS to the cooling pool. The re-performance of the comparisons using the Reference 29 version of NRELAP5 shows the same level of agreement as for the original comparisons. 5.3.3.12 NLT-15-p2 Test Description The objective of the NLT-15-p2 test was to measure the scaled integral system response to a decrease in heat removal from the secondary side following a loss of feedwater event, followed by core decay heat removal via the DHRS. During the test, ((

                                                                }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

4. ((

                                         }}2(a),(c)

The sequence of events for the NLT-15p2 test is shown in Table 5-14 Table 5-14 NLT-15p2 sequence of events ((

                                                                                                        }}2(a),(c) 5.3.3.13           NLT-15 p2 Test Results

((

                         }}2(a),(c)

Comparison of NRELAP5 calculated parameters to the measured data are presented below. Note that in the figures, the data uncertainty is shown in dot-dashed lines. A comparison of the primary side behavior is discussed first. This comparison is followed by a discussion of the startup of the DHRS (first 500 seconds). A comparison of the secondary side is then presented, and finally, a discussion of the CPV behavior concludes this section. © Copyright 2022 by NuScale Power, LLC 287

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Primary side response Figure 5-126 shows the core heater rod power. As expected, the predicted power is an overlay of the measured power since core power is a boundary condition for the simulation. Pressurizer pressure for short term is shown in Figure 5-127. ((

                                                                }}2(a),(c) The calculated pressure in the short term, compared to the measured data, is considered reasonable.

The long term pressurizer pressure comparison is shown in Figure 5-128. ((

                              }}2(a),(c) Figure 5-128a presents the same comparison as Figure 5-128, but with the Reference 29 version of NRELAP5.

((

                         }}2(a),(c) Long term, the pressurizer level response shows excellent agreement with the measured data. Figure 5-129a presents the same comparison as Figure 5-129, but with the Reference 29 version of NRELAP5.

Once DHRS loop flow is established, heat removal through the steam generator tube coil cools the RPV fluid. The liquid volume in the primary loop shrinks as the RPV system cools. After the pressurizer empties, a decline in the RPV level below the pressurizer baffle plate continues as shown in Figure 5-130. This figure compares the calculated and measured RPV level. As observed, the NRELAP5 © Copyright 2022 by NuScale Power, LLC 288

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 comparison of the RPV level shows excellent agreement with the measured data, and thus gives credence that NRELAP5 correctly calculates the change in volume due to fluid cooldown. ((

                                                                     }}2(a),(c) Figure 5-130a presents the same comparison as Figure 5-130, but with the Reference 29 version of NRELAP5.

The RPV riser flow rate ((

                                  }}2(a),(c)

Figure 5-132 through Figure 5-134 compare calculated RPV loop temperatures to the measured data for the core inlet, the riser inlet, and the upper plenum. Reasonable to excellent agreement is observed in the fluid temperatures. ((

                                                                      }}2(a),(c) Figure 5-133a presents the same comparisons as Figure 5-132 and Figure 5-133, but with the Reference 29 version of NRELAP5. Similarly, Figure 5-134a presents the same comparison as Figure 5-134, but with the Reference 29 version of NRELAP5.

DHRS Startup The sequence of events shows ((

                                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                                              }}2(a),(c)

The measured DHRS HX level is shown in Figure 5-138. ((

                                                                                                 }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                            }}2(a),(c) the response of NRELAP5 is judged to be reasonable.

DHRS and Steam Generator Long Term Behavior A comparison of the short term and long term steam line pressure response is given in Figure 5-135 and in Figure 5-139, respectively. ((

                                                   }}2(a),(c) Figure 5-139a presents the same comparison as Figure 5-139, but with the Reference 29 version of NRELAP5.

The comparison of the DHRS HX inlet and outlet temperatures are shown in Figure 5-140 and Figure 5-141, respectively. ((

                                        }}2(a),(c) The overall response of the predicted inlet temperature is deemed reasonable. Figure 5-140a presents the same comparison as Figure 5-140, but with the Reference 29 version of NRELAP5.

The DHRS HX outlet temperature shown in Figure 5-141 ((

                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-142 and Figure 5-143 compares the short term and long term DHRS loop mass flow rate, respectively. ((

                         }}2(a),(c) Overall, the predicted DHRS flow rate follows the trends of the data and is deemed reasonable. Figure 5-143a presents the same comparison as Figure 5-143, but with the Reference 29 version of NRELAP5.

Comparisons of the level in the DHRS HX and in the steam generator tube coil are shown in Figure 5-144 and Figure 5-145. In general, the predicted DHRS HX level compares reasonably with the measured data. ((

                                                                                                   }}2(a),(c)

NRELAP5 shows reasonable agreement with the data trends. Figure 5-144a presents the same comparison as Figure 5-144, but with the Reference 29 version of NRELAP5. Differential pressure comparisons across the condensate line and across the DHRS steam line are given in Figure 5-147 and Figure 5-148. ((

                                                                                                       }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                   }}2(a),(c) The calculation of the heat removal through the steam generator tube coil is deemed reasonable. Figure 5-149a presents the same comparison as Figure 5-149, but with the Reference 29 version of NRELAP5.

Figure 5-150 compares the measured heat removal rate through the DHRS HX tubes with the predicted results. ((

                                                    }}2(a),(c) Again, overall, the trend of the data is represented and is deemed reasonable. Figure 5-150a presents the same comparison as Figure 5-150, but with the Reference 29 version of NRELAP5.

CPV Behavior A comparison of the CPV liquid level and the liquid temperatures are provided here. Figure 5-151 compares the measured CPV level with the calculated level. ((

                                       }}2(a),(c) The prediction of the CPV level is deemed reasonable.

Figure 5-152 through Figure 5-157 compares the measured and calculated CPV fluid temperature throughout the CPV. ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                          }}2(a),(c)

Figure 5-126 NLT-15p2, transient core power ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-127 NLT-15p2, transient RPV pressure short term ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-128 NLT-15p2, transient RPV pressure ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-128a NLT-15p2, transient RPV pressure

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-129 NLT-15p2, transient pressurizer level ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-129a NLT-15p2, transient pressurizer level

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-130 NLT-15p2, transient RPV level ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-130a NLT-15p2, transient RPV level

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-131 NLT-15p2, transient riser mass flow rate ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-132 NLT-15p2, transient core inlet temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-133 NLT-15p2, transient riser inlet temperature ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-133a NLT-15p2, transient core inlet and riser inlet temperatures

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-134 NLT-15p2, transient upper plenum temperature ((

                                                                                          }}2(a),(b),(c),ECI Figure 5-134a NLT-15p2, transient upper plenum temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-135 NLT-15p2, transient secondary side pressure - 0 to 500 seconds ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-136 NLT-15p2, transient DHRS loop flow - 0 to 500 seconds

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-137 NLT-15p2, transient measured steam line temperatures - 0 to 500 seconds ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-138 NLT-15p2, transient DHRS HX level - 0 to 500 seconds

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-139 NLT-15p2, transient secondary side pressure ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-139a NLT-15p2, transient secondary side pressure

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-140 NLT-15p2, transient DHRS HX inlet temperature ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-140a NLT-15p2, transient DHRS HX inlet temperature

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-141 NLT-15p2, transient DHRS HX outlet temperature ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-142 NLT-15p2, transient DHRS loop flow - short term

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-143 NLT-15p2, transient DHRS loop flow rate - long term ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-143a NLT-15p2, transient DHRS loop flow rate - long term

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-144 NLT-15p2, transient DHRS HX level ((

                                                                                            }}2(a),(b),(c),ECI Figure 5-144a NLT-15p2, transient DHRS HX level

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-145 NLT-15p2, transient steam generator tube coil level - long term ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-146 NLT-15p2, transient steam generator tube coil level - short term

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-147 NLT-15p2, transient DHRS condensate line differential pressure ((

                                                                                       }}2(a),(b),(c),ECI Figure 5-148 NLT-15p2, transient DHRS steam line differential pressure

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-149 NLT-15p2, transient steam generator tube coil power removal ((

                                                                                      }}2(a),(b),(c),ECI Figure 5-149a NLT-15p2, transient steam generator tube coil power removal

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-150 NLT-15p2, transient DHRS power removal ((

                                                                                         }}2(a),(b),(c),ECI Figure 5-150a NLT-15p2, transient DHRS power removal

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-151 NLT-15p2, transient cooling pool vessel level ((

                                                                                             }}2(a),(b),(c),ECI Figure 5-152 NLT-15p2, transient cooling pool vessel fluid temperature at level 3 (below decay heat removal heat exchanger)

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-153 NLT-15p2, transient cooling pool vessel fluid temperature at level 5 (near bottom of DHRS heat exchanger) ((

                                                                                             }}2(a),(b),(c),ECI Figure 5-154 NLT-15p2, transient cooling pool vessel fluid temperature at level 6 (near midpoint of DHRS heat exchanger)

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-155 NLT-15p2, transient cooling pool vessel fluid temperature at level 7 (top to just above DHRS heat exchanger region) ((

                                                                                           }}2(a),(b),(c),ECI Figure 5-156 NLT-15p2, transient cooling pool vessel fluid temperature at level 8 (above DHRS heat exchanger region)

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-157 NLT-15p2, transient cooling pool vessel fluid temperature at level 9 (above DHRS heat exchanger region) ((

                                                                                                }}2(a),(b),(c),ECI 5.3.3.14           NLT-15p2 Summary Test NLT-15p2 is a loss of feedwater scenario followed by core decay heat removal via the DHRS system. ((
                           }}2(a),(c)

The NRELAP5 simulation of this test showed reasonable to excellent agreement with the initial pressure and loop temperature response in the RPV. After isolation of the secondary side and during the initial stages of DHRS heat removal operation, the RPV pressure rose due to heat addition from the core decay power and lack of complete removal through the steam generator tube coil. A rise in the pressurizer liquid level was also observed. Both the rise in the RPV pressure and the rise in the pressurizer level were well predicted by NRELAP5. After DHRS heat removal flow was established, the RPV pressure turned over and declined throughout the remainder of the test. NRELAP5 also predicted this pressure behavior in general. © Copyright 2022 by NuScale Power, LLC 316

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Both the pressurizer level response and the RPV level response showed excellent agreement with the measured response, thus showing that NRELAP5 can accurately predict change in fluid volume due to cooldown. The predicted RPV loop temperatures showed reasonable to excellent agreement with the measured data demonstrating that heat removal through the steam generator tube coil was reasonably predicted. Prediction of the secondary side behavior showed reasonable agreement with the measured data. All major trends of the data were modeled correctly by NRELAP5. ((

                         }}2(a),(c) However, NRELAP5 reasonably calculated the trends of the data.

((

                                                                             }}2(a),(c)

The calculated fluid temperature inside the DHRS heat exchanger tubes generally predicted the trends of the measured data. ((

                                                                                   }}2(a),(c)

The calculated differential pressure in the DHRS condensate line and steam line ((

                                            }}2(a),(c)

Overall, NRELAP5 is capable of producing reasonable results in simulating events such as those observed in the NLT-15p2 test. The re-performance of the comparisons using the Reference 29 version of NRELAP5 shows the same level of agreement as for the original comparisons. © Copyright 2022 by NuScale Power, LLC 317

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.4 Code to Code Benchmark for Integral Assessment of Reactivity Event

Response

5.3.4.1 Background The NRELAP5 code is used in the non-LOCA evaluation model to perform non-LOCA system transient analysis for an NPM. A series of code-to-code benchmarking comparison calculations was performed with the RETRAN-3D code to validate the performance of NRELAP5s point kinetics model during the reactivity transient events and to supplement the validation of the integral primary system thermal-hydraulic response to reactivity transients. RETRAN-3D was developed to perform transient thermal-hydraulic analysis of light water reactors. RETRAN-3D was developed as an evolution to the RETRAN codes that have been sponsored by EPRI since 1975 and used for licensing basis analyses of commercial light water reactors in the U.S. The RETRAN-3D code is based on the one-dimensional homogeneous equilibrium model, in comparison to the two-fluid, non-equilibrium non-homogeneous fluid model utilized in the NRELAP5 code. The NRC reviewed RETRAN-3D and issued a Safety Evaluation Report indicating that RETRAN-3D is an acceptable tool for performing PWR licensing analyses for a series of categories of anticipated transients, infrequent incidents and accident analyses for PWRs, including the reactivity transient events (Reference 12). Compared to operating PWRs, an NPM natural circulation primary system flow and helical coil SG design are unique features. For NRELAP5, a new helical coil SG component was added to the NRELAP5 code as described in Reference 2; primary and secondary side fluid flow and heat transfer over the SG were validated against testing performed at the SIET-TF1 and SIET-TF2 facilities, as described in Section 5.3.5. RETRAN-3D does not include specific models for helical coil SG heat transfer and wall friction; as described in Section 5.3.4.2, for the purposes of the benchmark calculations, a modeling simplification was made such that the RETRAN-3D and NRELAP5 primary side heat transfer coefficients were consistent under steady state conditions. Therefore, the scope of the benchmark calculation is focused on the reactivity response and the integral primary side response during a reactivity transient. The scope of the code-to-code benchmarking includes comparisons of reactor power, reactivity, primary side flow, pressure and temperatures for a set of benchmarking reactivity transients. These stylized benchmark cases are designed so as to be representative of the NuScale reactivity transients, but simplified to minimize impact of model differences in the secondary side or control system responses that are not the focus of the benchmark comparisons. © Copyright 2022 by NuScale Power, LLC 318

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.4.2 Approach for RETRAN-3D Benchmark To perform the code-to-code benchmark calculations, an NPM was modeled using both NRELAP5 and RETRAN-3D. The NRELAP5 base model used for code-to-code benchmarking was consistent with the base model described in Section 6.0 of this report. The RETRAN-3D base model was developed based on the NRELAP5 base model. The NRELAP5 model provided the geometric information, trips, control systems and reactor kinetics to convert to RETRAN-3D input cards. The NRELAP5 model volume, junction and heat conductor (heat structure) input were used as the initial basis for the RETRAN-3D model input and nodalization. ((

                                                           }}2(a),(c)

For NRELAP5, a new helical coil SG component was added to the NRELAP5 code as described in Reference 2 to model the helical coil SG heat transfer and wall friction in an NPM design. In RETRAN-3D there is no helical coil model, ((

                                                                          }}2(a),(c)

Differences in the predicted pressurizer pressure and level were observed in the benchmark calculation results due to differences between the pressurizer modeling in the NRELAP5 and RETRAN-3D calculations. ((

                                                                                                }}2(a),(c) The RETRAN-3D thermal-hydraulics computer code used in this study, the version with a safety evaluation report (SER) from the NRC, uses a basic three-equation homogeneous equilibrium model. When two-phase, the constraint on this representation is equilibrium (saturation) between the phases of water. When fluid surges into or out of the pressurizer, steam can be compressed or water can flash, which is why RETRAN-3D has a two-region non-equilibrium pressurizer model that allows non-equilibrium between the phases of water.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 RETRAN-3D also includes a volume change term in the phasic energy equations for the two-region component. The RETRAN-3D two region model allows the user limited control at the interface. Additional interfacial modeling is available in RETRAN-3D when using the sub-node option in the code. Comparisons between the NRELAP5 code and two region models (Reference 13, for example) can show larger pressures in a two-region component than in NRELAP5 for in-surge transients. This difference is due to some degree because of work terms for volume change and using enthalpy in the formulation. RETRAN-3D uses enthalpy at the junctions for the energy equations whereas NRELAP5 uses internal energy. Enthalpy contains internal energy in addition to the flow work term, which NRELAP5 does not include in its formulation, except for special application models for blow down. There are also differences in pressurizer spray, the letdown and charging models, along with insurge for the two codes. Contrary to the NRELAP5 model that uses a standard junction to model spray, the RETRAN-3D model uses a special spray junction in the pressurizer spray system that condenses sufficient steam from the vapor region to bring the spray flow to saturation. This model involves removal of mass and energy from the vapor region of the pressurizer and depositing the mass of the spray and condensed vapor directly into the liquid region of the pressurizer without a time delay, which leads to higher pressurizer pressure. After the RETRAN-3D base deck was developed, a steady-state calculation was performed where adequately consistent results between the two code calculations were obtained. Then, four transient calculations were performed. These stylized transients were selected to cover a range of reactivity insertion rates and RCS response that are observed in NPM reactivity event calculations. The calculations performed were:

1. Uncontrolled rod withdrawal from full power steady-state, using a higher reactivity insertion rate to result in a high power MPS actuation signal (fast uncontrolled rod withdrawal [UCRW]).

2. Uncontrolled rod withdrawal from full power steady-state, using a lower reactivity insertion rate to result in a high pressurizer pressure MPS actuation signal (slow UCRW).

3. Power reduction from full power to 50 percent of rated power.

4. Dropped control rod assembly from 50 percent rated power steady state conditions, using a dropped rod worth that is relatively large, but insufficient to cause reactor trip.

((

                                                 }}2(a),(c)

For the reactivity transients, the focus of the benchmark comparison is on the primary side response before scram, driven by the reactivity insertion and power © Copyright 2022 by NuScale Power, LLC 320

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 change, and the short term responses after scram, driven by the scram worth. The longer term and secondary side response is not examined in detail due to the simplified modeling of the helical coil SG in the RETRAN-3D model, and the high pressure condensation conditions after DHRS actuation that are not typical of operating pressurized water reactors. Therefore the key parameters of interest for the reactivity benchmarking transients are reactor power, reactivity, pressurizer pressure and level, core flow, and core temperatures. These parameters are plotted and compared for each transient. When performing code-to-code comparisons, agreement must be assessed in some manner. The method used herein is based on RG 1.203 (Reference 1) as described in Section 5.3. The acceptance criterion applied herein for the benchmarking calculations is that at least reasonable agreement must be observed. 5.3.4.3 Fast Uncontrolled Rod Withdrawal from Full Power In this case, the uncontrolled rod withdrawal occurs at time zero from the full power steady state condition and a high reactivity insertion rate (13.36 pcm/sec) is assumed. This reactivity insertion rate causes a reactor scram on high power. The DHRS is not actuated on high power and therefore normal secondary side cooling continues. The automatic turbine trip following reactor trip is not credited. The steam flow varies depending on the SG pressure. ((

                                                                          }}2(a),(c)

Figure 5-158 through Figure 5-164 show excellent agreement between the two models. At the end of the full power steady state initialization, the core flow, inlet temperatures and outlet temperatures are slightly different between the two models. These differences are carried over to the transient, which can be observed at time zero on Figure 5-162 through Figure 5-164. Due to these steady-state differences and their effects to the reactivity feedback, the core power increase rates are slightly different when the UCRW reactivity insertion is modeled. The RETRAN-3D model predicts a slightly earlier time to reach the high core power trip analytical limit (Figure 5-158 and Figure 5-159). The core power and reactivity curves are very close to each other between the two models during the transient. The pressurizer pressure and level match well during the short time after scram, and then the NRELAP5 model predicts a slightly higher pressurizer pressure and level (Figure 5-160 and Figure 5-161). The core flow and outlet temperature match fairly well between the RETRAN-3D model and the NRELAP5 model, after considering the differences from the steady state condition at the transient initiation (Figure 5-162 and Figure 5-164). The timing of the core inlet temperature drop is slightly off (Figure 5-163), indicating the heat transfer to the secondary side is slightly different between the two models. © Copyright 2022 by NuScale Power, LLC 321

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-158 Core power (fast uncontrolled rod withdrawal) ((

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

Figure 5-159 Total reactivity (fast uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-160 Pressurizer pressure (fast uncontrolled rod withdrawal) ((

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

Figure 5-161 Pressurizer level (fast uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-162 Core flow (fast uncontrolled rod withdrawal) ((

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

Figure 5-163 Core inlet temperatures (fast uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-164 Core outlet temperatures (fast uncontrolled rod withdrawal) ((

                                                                                                      }}2(a),(b),(c) 5.3.4.4            Slow Uncontrolled Rod Withdrawal from Full Power In this case, the uncontrolled rod withdrawal occurs at time zero from the full power steady state condition, and a lower reactivity insertion rate (5.344 pcm/sec) is assumed. This reactivity insertion results in RCS heatup and reactor scram from a high RCS pressure rather than a high power condition. During the slow UCRW transient, upon reactor trip on high RCS pressure, DHRS is actuated and therefore normal steam flow and feedwater flow are isolated.

In addition to the reduced reactivity insertion rate, pressurizer spray and CVCS were isolated on transient initiation to maximize the pressurization rate. The UCRW reactivity insertion is initiated at time zero. Figure 5-165 through Figure 5-171 show reasonable to excellent agreement between the two codes. As explained in Section 5.3.4.3, at the end of the full power steady state initialization, the core flow, inlet and outlet temperatures are slightly different between the two models. These differences are carried over to the transient, which can be observed at time zero on Figure 5-169 through Figure 5-171. Due to these steady-state differences and their effects to the reactivity feedback, the rates of core power increase are slightly different when the UCRW reactivity insertion is modeled. The RETRAN-3D model predicts a slightly faster power increase rate than NRELAP5 (Figure 5-165). Due to the faster power increase and different pressurizer models (discussion in © Copyright 2022 by NuScale Power, LLC 325

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Section 5.3.4.2), RETRAN-3D predicts higher increase rates on the pressurizer pressure and level after transient initiation (Figure 5-167 and Figure 5-168), and therefore a slightly earlier timing of the reactor trip on high pressurizer pressure. Following the reactor scram, the core power and reactivity curves are very close to each other between the two models during the transient (Figure 5-165 and Figure 5-166). Pressurizer pressure and level continue to increase and reach their peak values before decreasing. NRELAP5 predicts a slightly higher peak pressurizer pressure and level (Figure 5-167 and Figure 5-168). As the pressure and level decrease, RETRAN-3D has a faster initial depressurization resulting in a lower pressurizer pressure. Then the relative rate of decrease changes and the RETRAN-3D pressure is higher than the NRELAP5 pressure. ((

                               }}2(a),(b),(c) The pressurizer levels predicted by the two codes show similar behavior. The discrepancy is attributed to different heat removal rates from the pressurizer between the NRELAP5 and RETRAN-3D pressurizer models, due to differences in the pressurizer modeling as discussed in Section 5.3.4.2.

The core flow and outlet temperature match fairly well between the RETRAN-3D model and the NRELAP5 model, after considering the differences from the steady state condition at the transient initiation (Figure 5-169 and Figure 5-171). The core inlet temperatures shown the same trend and the biggest difference is approximately 2 percent (Figure 5-170). The core inlet temperature difference could be attributed to the core flow differences. © Copyright 2022 by NuScale Power, LLC 326

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-165 Core power (slow uncontrolled rod withdrawal) ((

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

Figure 5-166 Total reactivity (slow uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-167 Pressurizer pressure (slow uncontrolled rod withdrawal) ((

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

Figure 5-168 Pressurizer level (slow uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-169 Core flow (slow uncontrolled rod withdrawal) ((

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

Figure 5-170 Core inlet temperature (slow uncontrolled rod withdrawal) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-171 Core outlet temperature (slow uncontrolled rod withdrawal) ((

                                                                                                   }}2(a),(b),(c) 5.3.4.5            Power Reduction In this case, the benchmark calculations start at full power conditions, followed by a reduction to 50 percent power ((
                                                                                             }}2(a),(c) It is noted that the dropped rod transient benchmark calculation (Section 5.3.4.6) is then performed as a continuation from the 50 percent power steady state condition.

((

                                      }}2(a),(b)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-172 through Figure 5-178 show reasonable to excellent agreement between the two codes. The core power and reactivity curves are almost identical (Figure 5-172 and Figure 5-173). At the time of the power reduction, the RETRAN-3D pressurizer level drops below the pressurizer level of NRELAP5. ((

                                                                     }}2(a),(c),(b) During the power reduction, the RETRAN-3D predicted pressurizer pressure is higher than NRELAP5. The biggest difference of approximately 3.5 percent is attributed to different heat removal rates from the pressurizer between the NRELAP5 and RETRAN-3D pressurizer models, due to differences in the pressurizer modeling as discussed in Section 5.3.4.2.

As explained in Section 5.3.4.3, at the end of the full power steady state initialization, the core flow, core inlet and core outlet temperatures are slightly different between the two models. These differences are carried over to the transient calculation, which can be observed at time zero on Figure 5-176 through Figure 5-178. During the transient power reduction, the core flow, core inlet and core outlet temperatures have a consistent small difference between the RETRAN-3D and NRELAP5 predictions. Figure 5-172 Core power (power reduction) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-173 Total reactivity (power reduction) ((

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

Figure 5-174 Pressurizer pressure (power reduction) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-175 Pressurizer level (power reduction) ((

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

Figure 5-176 Core flow (power reduction) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-177 Core inlet temperature (power reduction) ((

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

Figure 5-178 Core outlet temperature (power reduction) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.4.6 Dropped Control Rod The dropped control rod case is initiated from the 50 percent power steady state condition at time zero. ((

                                                     }}2(a),(c)

Figure 5-179 through Figure 5-185 show reasonable to excellent agreement between the two codes. The core power drops from 80 MWth to ((

                                                            }}2(a),(b),(c) due to reactivity feedback effects (Figure 5-179 and Figure 5-180). The core power and reactivity curves are very close.

The pressurizer pressure and pressurizer level are close when they decrease following the dropped rod negative reactivity insertion. When the pressurizer pressure and level start to increase following the power recovery, NRELAP5 predicts slightly higher pressure and level compared to RETRAN-3D. ((

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

As explained in Section 5.3.4.5, when the transient reaches steady state at the end of the power reduction, the core flow, core inlet and core outlet temperatures are slightly different between the two models. These differences are carried over to the transient calculation, which can be observed at time zero on Figure 5-183 through Figure 5-185. During the transient, the core flow, core inlet and core outlet temperatures have a consistent small difference between the RETRAN-3D and NRELAP5 predictions. © Copyright 2022 by NuScale Power, LLC 335

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-179 Core power (dropped control rod) ((

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

Figure 5-180 Total reactivity (dropped control rod) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-181 Pressurizer pressure (dropped control rod) ((

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

Figure 5-182 Pressurizer level (dropped control rod) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-183 Core flow (dropped control rod) ((

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

Figure 5-184 Core inlet temperature (dropped control rod) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-185 Core outlet temperature (dropped control rod) ((

                                                                                                       }}2(a),(b),(c) 5.3.4.7            Conclusions from Benchmark Four different transients were performed for code-to-code benchmarking between NRELAP5 and RETRAN-3D: Reactivity insertion reflecting a fast UCRW from full power conditions, reactivity insertion reflecting a slow UCRW from full power conditions, negative reactivity insertion to reduce power from 100 percent to 50 percent power, and negative reactivity insertion simulating a dropped rod from 50 percent power. The results from all four of the transients showed that the comparison between the power and the total reactivity were consistently excellent, in that the calculation results of the two codes lined up nearly identically with one another. This agreement is important because the main purpose of the benchmark is to compare the point kinetics model response between NRELAP5 and RETRAN-3D.

There was also an overall pattern in the core inlet and outlet temperatures of a small temperature difference from the steady state conditions that continued throughout the calculation; however, the overall phenomena matched well. The pressurizer pressure and the pressurizer level were the two characteristics of the system that did not provide as close of a comparison as the other four characteristics, but still reasonably close. The discrepancy is attributed to differences between the NRELAP5 and RETRAN-3D pressurizer models, as discussed in Section 5.3.4.2. © Copyright 2022 by NuScale Power, LLC 339

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Overall, NRELAP5 and RETRAN-3D have reasonable to excellent agreement on the parameters of interest in the reactivity feedback and primary system response in the benchmark calculations. The power and reactivity responses predicted by the NRELAP5 point kinetics model during the transients show excellent agreement with the RETRAN-3D prediction. 5.3.5 Steam Generator Modeling 5.3.5.1 Background NuScales LOCA Topical Report (Reference 2) Section 7.4 discusses the validation of NRELAP5 for helical coil SG (HCSG) modeling. The validation was mainly against SIET TF-1 and TF-2 test data. It was concluded that NRELAP5 showed reasonable to excellent agreement with test data. The validation is further investigated in this report to ensure the unique characteristics of the non-LOCA transients (comparing to LOCA) are identified and evaluated. Specifically, this investigation ensures the operating ranges expected during the non-LOCA transients are adequately covered by the validated ranges. The evaluation was originally performed for the NPM design at the time. Subsequent design work resulted in other NPM designs within the confines of the general NPM description in Section 3.0. Additional bottom-up evaluation confirmed that the Reference 29 NRELAP5 key physical models for SG and DHRS heat removal during non-LOCA events remain applicable. 5.3.5.2 Helical Coil Steam Generator Modeling In the LOCA topical report (Reference 2), NRELAP5 shows reasonable to excellent agreement with test data on the HCSG primary and secondary side, based on comparisons against the SIET TF-2 and SIET TF-1 test data. 5.3.5.3 Helical Coil Steam Generator Operating Ranges vs. Validated Ranges In the LOCA topical report (Reference 2), NRELAP5 shows reasonable to excellent agreement with test data on the helical coil SG secondary side. ((

                               }}2(a),(c) Further evaluation is provided herein to ensure NRELAP5 is validated for DHRS operation.

Based on the typical helical coil SG secondary pressure, temperature and flow rate during DHRS operation, ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

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

Table 5-15 summarizes the helical coil SG operating range for non-LOCA transients vs. the validated range in NRELAP5. The majority of the helical coil SG secondary side operating range is covered by the validated range of NRELAP5. ((

                                                                               }}2(a),(c) Therefore, the operating range of the helical coil SG primary side is sufficiently covered by the validated range of NRELAP5.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-15 Non-LOCA transients helical coil steam generator operating range vs. NRELAP5 validated range ((

                                                                                                }}2(a),(b),(c) 5.3.5.4            Helical Coil Steam Generator Nodalization Sensitivity

((

                                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                       }}2(a),(c)

Based on these studies, modeling the helical coil SG with ((

                      }}2(a),(c) nodes is expected to produce reasonably accurate results for the non-LOCA transients. Considering these studies, steam generator modeling requirements are summarized in Section 6.1.1 and Section 6.1.4.2 for the primary and the secondary, respectively.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-186 Coil 1 representative pressure drop for (( }}2(a),(c) nodes (left) and (( }}2(a),(c) nodes (right) ((

                                                                                             }}2(a),(b),(c),ECI Figure 5-187 Coil 1 representative fluid temperature for ((             }}2(a),(c) nodes (left) and

(( }}2(a),(c) nodes (right) ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-188 Coil 1 representative wall temperature for (( }}2(a),(c) nodes (left) and (( }}2(a),(c) nodes (right) ((

                                                                                                }}2(a),(b),(c),ECI Figure 5-189 Decrease in feedwater temperature nodalization sensitivity steam generator secondary side inlet pressure

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-190 Decrease in feedwater temperature nodalization sensitivity steam generator secondary side outlet pressure ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-191 Decrease in feedwater temperature nodalization sensitivity reactor coolant system flow rate ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-192 Decrease in feedwater temperature nodalization sensitivity reactor coolant system lower plenum pressure ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-193 Decrease in feedwater temperature nodalization sensitivity reactor coolant system core inlet temperature ((

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

Figure 5-194 Decrease in feedwater temperature nodalization sensitivity reactor power ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.6 Heat Transfer Correlation Comparison 5.3.6.1 Background ((

                                  }}2(a),(c) 5.3.6.2            Computational Fluid Dynamics Calculations

((

                          }}2(a),(c)

Table 5-16 Comparison of correlation to CFD results - CFD model 2 ((

                                                                                                 }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 5-16 Comparison of correlation to CFD results - CFD model 2 (Continued) ((

                                                                                                  }}2(a),(c)

((

                                                          }}2(a),(c) 5.3.6.3            Comparison of Heat Transfer Correlations within NRELAP5

((

                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-195 Comparison of heat transfer correlation options for steam generator primary side heat transfer rate (linear time scale) ((

                                                                                                       }}2(a) (c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-196 Comparison of heat transfer correlation options for steam generator primary side heat transfer rate (logarithmic time scale) ((

                                                                                                      }}2(a),(c) 5.3.6.4            Heat Transfer Correlation Conclusion

((

                                                                               }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.3.7 NIST-2 Steam Generator - Decay Heat Removal System Integral Effects Tests 5.3.7.1 NIST-2 Facility Section 5.3.2.1 describes the NIST-1 facility. In 2018 work began on the NIST-1 facility to modify the secondary flow piping, the DHRS, and minor components of other systems. Specific modifications include: Increased main steam system pressure capacity Increased steam generator and DHRS pressure capacity Replacement of most secondary side piping and valves Installation of a steam-flow control valve with improved controllability Installed new instruments or increased existing instrument ranges The DHRS is a system of piping that interfaces with the secondary flow piping, allowing closed loop DHRS testing or once through DHRS testing. The DHRS is divided into five pipe sections; the DHRS bypass steam line, the DHRS steam line, the DHRS scaled height two tube heat exchanger, the DHRS condensate line, and the DHRS vent line. The DHRS piping configuration is shown in Figure 5-197 and a description of each pipe section follows. ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 After this upgrade, the facility name was changed to NIST-2. Except for the changes described above, the NIST-2 facility is consistent with the descriptions and figures shown in Section 5.3.2.1. © Copyright 2022 by NuScale Power, LLC 355

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-197 NIST-2 decay heat removal system configuration ((

                                                                                                 }}2(a),(c),ECI 5.3.7.2            NIST-2 Steam Generator - Decay Heat Removal System Integral Effects Test Objective Initial assessment and validation of NRELAP5 for prediction of SG-DHRS phenomena during non-LOCA event sequences was provided in Section 5.3.1, Section 5.3.2, Section 5.3.3, and Section 5.3.5. NuScale identified that additional testing and analysis was needed to reduce the uncertainty associated with heat removal by the DHRS during non-LOCA event sequences.

Section 3.0 provides general discussion of NPM designs, including the DHRS. Changes between NPM designs have resulted in changes to the DHRS geometry, such as the DHRS steam piping configuration, and detailed condenser design. The PIRT in Section 5.1 identifies several high ranked phenomena associated with DHRS operation in Phase 3. NuScale identified that additional testing and analysis was needed to understand the impact of potential DHRS design changes on the relevant high ranked phenomena. © Copyright 2022 by NuScale Power, LLC 356

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The NIST-2 SG-DHRS integral effects testing is designed to enhance the DHRS validation database to address the above considerations. The NIST-2 facility permits test conditions that are significantly closer to the prototypical NPM operation conditions than was achieved in prior testing. 5.3.7.3 Applicability of Decay Heat Removal System Operation to Small Break Loss-of-Coolant Accident Sequences The assessment and validation of NRELAP5 for prediction of SG-DHRS phenomena provided in Section 5.3.1, Section 5.3.2, and Section 5.3.3 was focused on application during non-LOCA event sequences. DHRS actuation is also expected to occur during LOCA event sequences. For smaller LOCA break sizes, there may be an extended period where DHRS is in operation but ECCS is not yet actuated. In a small break LOCA sequence where there is such an extended period, the phenomena associated with DHRS performance are applicable. ((

                                                                                                    }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                       }}2(a),(c)

Based on these assessments, the additional NIST-2 testing and analysis is applicable to justify the heat removal by the DHRS during the small break LOCA conditions prior to ECCS valve opening. 5.3.7.3.1 Effect of Break Size and Location on Decay Heat Removal System Performance As will be discussed further in Section 5.3.7.5, the NIST-2 testing is representative of an LOFW transient. Therefore, an analysis of the SG-DHRS performance in the LOFW transient and small break LOCA was performed to confirm that there were no unique high-ranked phenomena for DHRS performance in a small break LOCA as compared to an LOFW transient and that the SG-DHRS performance is similar to the performance in the small break LOCA scenario. A summary of the analysis and results is presented below. ((

                                                                                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                       }}2(a),(c) Therefore, NRELAP5 validation and applicability to simulate DHRS heat removal during non-LOCA events extends to simulate DHRS heat removal during LOCA events, particularly small LOCA events with an extended timeframe of liquid flow over the riser and DHRS heat removal prior to ECCS actuation.

Figure 5-198 Reactor coolant system average temperature transient response for loss of feedwater break scenarios ((

                                                                                                         }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-199 Reactor coolant system saturation temperature transient response for loss of feedwater break scenarios ((

                                                                                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-200 Reactor coolant system flow transient response for loss of feedwater break scenarios ((

                                                                                              }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-201 Reactor coolant system pressure transient response for loss of feedwater break scenarios ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-202 Secondary side pressure transient response for loss of feedwater break scenarios ((

                                                                                              }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-203 Decay heat removal system flow transient response for loss of feedwater break scenarios ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-204 Decay heat removal system power transient response for loss of feedwater break scenarios ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-205 Decay heat removal system power compared to reactor coolant system temperature for loss of feedwater break scenarios ((

                                                                                                        }}2(a),(c) 5.3.7.4            NIST-2 NRELAP5 Model Description The NRELAP5 model of the NIST-1 facility is described in Section 5.3.3.3. The differences between the NRELAP5 models for NIST-2 and NIST-1 model are the result of incorporating the facility changes described in Section 5.3.7.1. The main changes are in the secondary side and the DHRS. Figure 5-206 through Figure 5-213 show the NRELAP5 component types and numbers for the secondary side and DHRS for NIST-2.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-206 NIST-2 modified separate effects main steam line part 1 model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-207 NIST-2 modified separate effects main steam line part 2 model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-208 NIST-2 modified bypass steam line to containment vessel model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-209 NIST-2 new integral effects decay heat removal system steam line model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-210 NIST-2 modified separate effects decay heat removal system steam line model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-211 NIST-2 new decay heat removal system heat exchanger model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-212 NIST-2 modified decay heat removal system condensate return line model ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-213 NIST-2 modified feedwater system model ((

                                                                                                 }}2(a),(c),ECI 5.3.7.5            NIST-2 Steam Generator - Decay Heat Removal System Integral Effects Test Based on the acceptance criteria described in Section 4.2, figures of merit in the non-LOCA events include CHFR, RCS pressure, SG pressure, and CNV pressure. These figures of merit are affected by the DHRS heat removal performance during Phase 3. In addition, Section 5.3.7.3 concludes that DHRS heat removal performance is also applicable during small break LOCA events prior to ECCS performance (corresponding to Phase 3). The high-ranked phenomena identified in the NPM non-LOCA PIRT focusing on the DHRS and its interfacing systems (i.e., RCS and reactor pool) during Phase 3 can be summarized as follows.

((

                                    }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                               }}2(a),(c) Therefore, the SG and DHRS heat exchanger are the primary focus for the NIST-2 DHRS testing. Table 5-17 identifies SG/DHRS testing performed at the NIST-2 facility and assessed with NRELAP5 as part of the non-LOCA EM development.

Table 5-17 NIST-2 integral effects SG/DHRS testing for NRELAP5 code validation DHRS Steam Line Test Identifier Initial DHRS Inventory Test Description Orifice Run-1 Base Base This is the base configuration for the testing. 5.3.7.6 Run 1 Test Description The NIST-2 facility had the following initial conditions at the start of the Run 1 test: ((

                                              }}2(a),(b),(c),ECI The NIST-2 valve configuration at the start of the Run 1 test is shown in Figure 5-197. ((
                                                                                                   }}2(a),(b),(c),ECI

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                               }}2(a),(b),(c),ECI The sequence of events for the Run 1 test is shown in Table 5-18.

Table 5-18 NIST-2 Run 1 sequence of events ((

                                                                                               }}(a),(b),(c),ECI 5.3.7.7            Run 1 Test Results Figure 5-214 through Figure 5-234 show the results of the code-to-data comparisons for Run 1.

Figure 5-214 shows the core power. Figure 5-215 shows the SG/DHRS active loop inventory. ((

                                                                                                    }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                    }}2(a),(c)

Figure 5-216 and Figure 5-217 show the DHRS heat exchanger and SG power comparison for the long-term and short-term respectively. Figure 5-218 and Figure 5-219 show the SG-DHRS loop flow comparison for the long-term and short-term respectively. ((

                             }}2(a),(c) Overall, the NRELAP5 calculation has reasonable to excellent agreement with the test data.

Figure 5-220 shows the code-to-data comparison for the PZR pressure and the secondary side pressure near the steam drum. Before the PZR heater is tripped, both data and code maintain the PZR pressure at approximately ((

                                                            }}2(a),(c) Overall the secondary side pressure has excellent agreement with the test data.

Figure 5-221 shows the code-to-data comparison for the DHRS heat exchanger and SG levels. ((

                                                                                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                         }}2(a),(c) Nevertheless, the overall predicted DHRS heat exchanger and SG levels have reasonable to excellent agreement with the test data.

Figure 5-222 shows the code-to-data comparison for the steam drum collapsed liquid level. ((

                                                      }}2(a),(c)

Figure 5-223 shows the code-to-data comparison for the DHRS heat exchanger inlet and outlet temperatures. Figure 5-224 shows the code-to-data comparison for the SG secondary side inlet and outlet temperatures. The SG secondary side outlet temperature and DHRS heat exchanger inlet temperatures are both close to the loop saturation temperature (Figure 5-225). The code predictions of these two temperatures both have reasonable to excellent agreement with the data, similar to the secondary side pressure. From Figure 5-223 and Figure 5-224, the predicted DHRS heat exchanger outlet temperature and SG secondary side inlet temperature are ((

                            }}2(a),(c)

Figure 5-226 shows the code-to-data comparison of the RCS primary temperatures at two locations: at the SG primary side's inlet (upper plenum) and outlet (mid downcomer). ((

                                                                                                        }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                        }}2(a),(c)

Figure 5-227 shows the code-to-data comparison of the PZR level. Figure 5-228 shows the code-to-data comparison of the RPV level. ((

                                                             }}2(a),(c)

Figure 5-229 shows the code-to-data comparison of the RCS primary flow. The prediction has excellent agreement with the data during the transient. ((

                                                                                      }}2(a),(c)

Figure 5-230 shows the code-to-data comparison of the CPV level. The prediction has excellent agreement with data ((

                                                                                }}2(a),(c)

Figure 5-233 shows the code-to-data comparison of the DHRS steam line differential pressure. Excellent agreement is observed. Figure 5-234 shows the comparison of the DHRS steam line orifice differential pressure. Reasonable agreement is observed. ((

                            }}2(a),(c) Code-to-data comparisons of other differential pressures show that most along the SG-DHRS loop have excellent agreement.

((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

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

Figure 5-214 NIST-2 Run 1 core power comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-215 NIST-2 Run 1 steam generator and decay heat removal system active loop inventory comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-216 NIST-2 Run 1 steam generator and decay heat removal system heat exchanger power comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-217 NIST-2 Run 1 steam generator and decay heat removal system heat exchanger power comparison - short-term ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-218 NIST-2 Run 1 steam generator and decay heat removal system loop flow comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-219 NIST-2 Run 1 steam generator and decay heat removal system loop flow comparison - short-term ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-220 NIST-2 Run 1 primary and secondary pressure comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-221 NIST-2 Run 1 steam generator and decay heat removal system level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-222 NIST-2 Run 1 steam drum level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-223 NIST-2 Run 1 decay heat removal system heat exchanger inlet and outlet temperatures comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-224 NIST-2 Run 1 steam generator secondary side inlet and outlet temperatures comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-225 NIST-2 Run 1 steam generator secondary side saturation temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-226 NIST-2 Run 1 steam generator primary side inlet and outlet temperatures comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-227 NIST-2 Run 1 pressurizer level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-228 NIST-2 Run 1 reactor pressure vessel level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-229 NIST-2 Run 1 primary flow rate comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-230 NIST-2 Run 1 cooling pool vessel level comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-231 NIST-2 Run 1 cooling pool vessel level 5 temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-232 NIST-2 Run 1 cooling pool vessel level 6 temperature comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-233 NIST-2 Run 1 decay heat removal system steam line differential pressure comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-234 NIST-2 Run 1 decay heat removal system steam line orifice differential pressure comparison ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-235 NIST-2 Run 1 decay heat removal system heat exchanger tube lower middle section fluid temperatures - data only ((

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

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 5-236 NIST-2 Run 1 decay heat removal system heat exchanger tube bottom fluid temperatures - data only ((

                                                                                              }}2(a),(b),(c),ECI 5.3.7.8            NIST-2 Decay Heat Removal System Integral Effects Test Summary Based on the overall comparison between NRELAP5 and the NIST-2 test data for the simulated LOFW transient, there is reasonable to excellent agreement between code prediction and data. Predicted SG-DHRS loop flow and DHRS heat exchanger and SG power match the test data very closely during quasi steady state. Predicted secondary side pressure and temperatures, as well as DHRS heat exchanger and SG levels, have reasonable to excellent agreement with test data. SG-DHRS loop differential pressures are also predicted with reasonable to excellent agreement to test data. The comparisons demonstrate that NRELAP5 has the capability to accurately predict SG-DHRS performance during transients.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 5.4 Conclusions of NRELAP5 Applicability for Non-LOCA The high-ranked phenomena identified by the PIRT process for NPM non-LOCA transients were evaluated with respect to the high-ranked phenomena identified by the PIRT process for NPM LOCA scenarios, as well as the NRELAP5 assessments performed as part of the NuScale LOCA evaluation model development. A gap analysis was performed to identify high-ranked phenomena for non-LOCA transients that are not assessed as part of the NuScale LOCA evaluation model development. High-ranked phenomena for non-LOCA events that are not assessed as part of the NuScale LOCA evaluation model development were addressed in different ways:

1. Additional NRELAP5 code assessment performed against separate effects or integral effects test data

2. Code-to-code benchmark performed between NRELAP5 and independent system thermal-hydraulics code

3. Phenomenon is addressed as part of the downstream subchannel analysis

4. Phenomenon is addressed by specifying appropriately conservative input to the system transient analysis In addition to the NRELAP5 assessments performed as part of the LOCA evaluation model development, further assessments were performed to demonstrate NRELAP5 qualification for high rank non-LOCA PIRT phenomena.

Assessment of NRELAP5 against KAIST data demonstrates the codes ability to model heat transfer within tubes and appropriately model the key thermal-hydraulic phenomena associated with condensation within the DHRS heat exchanges tubes. The comparisons between the NIST-1 HP-03 experimental data and the code calculated values showed that NRELAP5 has the capability to accurately predict the energy transfer across the DHRS heat exchanger tubes to the CPV fluid resulting in reasonable to excellent agreement in capturing cooling pool heat up during these tests in which the cooling pool remains subcooled. The comparisons between the NIST-1 HP-04 experimental data and the code calculated values showed that NRELAP5 has the capability to accurately predict the energy transfer across the DHRS heat exchanger tubes to the CPV fluid. The HP-04 test duration was longer than that of HP-03. In HP-04, saturated conditions in the pool were reached and thermal stratification developed. Although the CPV fluid heat-up profiles for the HP-04 test data were not fully reproduced in the NRELAP5 simulations of HP-04, the code-to-data comparisons of DHRS heat removal rate were well-matched, demonstrating that NRELAP5 is capable of predicting the total energy removal rate of the DHRS to the cooling pool. The comparisons between the NIST-1 NLT-2a integral experimental data and the code calculated values showed that NRELAP5 has the capability to accurately predict primary coolant system heat up and pressurization due to a loss of feedwater flow. © Copyright 2022 by NuScale Power, LLC 404

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 NRELAP5 assessment of the NIST-1 NLT-2b phases 1 through 4 demonstrates the ability of the code to predict the heat transfer from primary side to the SG and from the DHRS to CPV. After completion of core power maneuvering and after full start of the DHRS loop, NRELAP5 calculated pressurizer pressure, core inlet and outlet temperature, SG pressure, energy transfer from the primary to secondary, energy transfer from DHRS to CPV, SG level, DHRS level, and CPV level results are all within reasonable to excellent agreement with NLT-2b test data. The pressurizer level for NLT-2b phase 1 was ((

                                }}2(a),(b),(c) However, even with these discrepancies, the results show that the important parameter of total energy transfer from the primary side to the SG and from the DHRS to CPV are well predicted.

For NLT-2b phases 1 through 4, the calculated condensate temperature is in minimal agreement with data, with the temperature being over predicted by NRELAP5. However, even with this discrepancy, the important parameter of heat transfer from primary side to the SG and from the DHRS to CPV are well predicted. For NLT-2b phases 1 through 4, although the CPV fluid heat-up profiles were not fully captured in the NRELAP5 simulations, the code-to-data comparisons of DHRS heat removal rate were well-matched, demonstrating that NRELAP5 is capable of capturing the important parameter of total energy removal rate of the DHRS to the cooling pool and that the integral response is insensitive to the CPV temperature profile. ((

                                         }}2(a),(c)

NRELAP5 assessment of the NIST-1 NLT-15p2 test demonstrates the ability of the code to predict the heat transfer from primary side to the SG and from the DHRS to CPV. The NRELAP5 calculated pressurizer and RPV level, core inlet and outlet temperature, primary flow rate, steam generator pressure, energy transfer from the primary to secondary, energy transfer from DHRS to CPV, SG level, DHRS heat exchanger level, and CPV level results are predicted reasonably compared to the NLT-15p2 test data. The NRELAP5 calculation of the initial pressurizer pressure response to decrease in secondary side heat transfer due to loss of feedwater, and the RPV pressure turn over after core power decreased to decay heat levels and DHRS heat removal flow was established, was well-predicted. ((

                     }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The calculated fluid temperature inside the DHRS heat exchanger tubes predicted the trends of the measured data. ((

                   }}2(a),(c)

The calculated differential pressure across the DHRS steam line ((

                                          }}2(a),(c)

The overall comparison between NRELAP5 and the NIST-2 test data for the simulated LOFW transient, shows reasonable to excellent agreement between code prediction and data. Predicted SG-DHRS loop flow and DHRS heat exchanger and SG power match the test data very closely during quasi steady state. Predicted secondary side pressure and temperatures, as well as DHRS heat exchanger and SG levels, have reasonable to excellent agreement with test data. SG-DHRS loop differential pressures are also predicted with reasonable to excellent agreement to test data. The comparisons demonstrate that NRELAP5 has the capability to accurately predict SG-DHRS performance during transients. Four different transients were performed for code-to-code benchmarking between NRELAP5 and RETRAN-3D: Reactivity insertion representative of a fast UCRW from full power conditions, reactivity insertion representative of a slow UCRW from full power conditions, negative reactivity insertion to reduce power from 100 percent to 50 percent power, and negative reactivity insertion simulating a dropped rod from 50 percent power. The results from all four of the transients showed that the comparison between the power and the total reactivity were consistently excellent, in that the calculation results of the two codes were nearly identically with one another. NuScales LOCA Topical Report (Reference 2) Section 7.4 discusses the validation of NRELAP5 for helical coil SG modeling. The validation was mainly against SIET TF-1 and TF-2 test data. The operating range of the helical coil SG primary and secondary side is demonstrated to be sufficiently covered by the validated range of NRELAP5. NRELAP5 showed reasonable to excellent agreement with test data for all phenomena at conditions important for the non-LOCA analysis. Computational fluid dynamics was also used to validate the model incorporated in NRELAP5 for the primary side heat transfer in the helical SGs. A nodalization sensitivity of the steam generator for a main steam line break scenario was performed comparing the effect of modeling the SG ((

                                                                                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                       }}2(a),(c)

DHRS actuation is also expected to occur during LOCA event sequences. For smaller LOCA break sizes, there may be an extended period where DHRS is in operation but ECCS is not yet actuated. In a small break LOCA sequence where there is such an extended period, the phenomena associated with DHRS performance are applicable. The DHRS heat removal before ECCS actuation is important to determine ((

                                      }}2(a),(c) The effect of these different ranges of conditions has been examined and the impact on DHRS performance is reasonable considering underlying physics of the integral system behavior. NRELAP5 sensitivity calculations show that the discharge of RCS mass and energy during a small break LOCA does not degrade the performance of DHRS from the transient initiation until ECCS actuation near the top of the riser. No new important phenomena are identified due to DHRS operation in conjunction with a small liquid or vapor break, before ECCS actuation. Therefore, conclusions of NRELAP5 applicability to simulate non-LOCA SG/DHRS performance can be extended to SG/DHRS performance during LOCA conditions, particularly small LOCA conditions with an extended phase of liquid flow over the top of the riser and DHRS heat removal prior to ECCS actuation.

A bottom-up code applicability evaluation for the non-LOCA EM focused on NRELAP5 model ranges for SG/DHRS boiling/condensation wall heat transfer phenomena. ((

                                                                       }}2(a),(c) It is concluded that the models and correlations implemented in NRELAP5 are applicable to simulate NPM SG/DHRS heat transfer.

A top-down scaling evaluation was performed to evaluate PI groups for high-ranked SG/DHRS phenomena in the non-LOCA PIRT and compare the relative importance between the prototype (NPM) and model (NIST-2 test facility). ((

                                                                                                         }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                                     }}2(a),(c) The top-down scaling analysis demonstrates the consistency in the dominant PI groups for SG/DHRS operation between the NPM and NIST-2 while phenomena/processes that are less important may be more distorted between the prototype and model. It is concluded that NRELAP5 is applicable for NPM non-LOCA integral analysis.

Considering the high-ranked phenomena identified from the PIRT process, the NRELAP5 code along with an NPM system model is applicable for calculation of an NPM system response for the non-LOCA short-term transient event progression as part of this EM based on separate effects and integral effects testing, code-to-code benchmarking, bottom-up and top-down evaluations of SG/DHRS performance, and appropriate conservative input for initial and boundary conditions. © Copyright 2022 by NuScale Power, LLC 408

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 6.0 NuScale NRELAP5 Plant Model This section discusses the NuScale NRELAP5 non-LOCA plant transient model. A summary overview of the plant components and features simulated by the NRELAP5 model is provided, including the reactor primary and secondary (SG) systems, core fuel rods and kinetics, ECCS and DHRS, containment and reactor pool, and trips and controls. The NRELAP5 plant model is developed to support the non-LOCA analysis methodology described in Section 7.0. The model is developed following the NRELAP5 code manual user guidelines, supplemented by NuScale-specific modeling guidelines. The guidelines describe how to model a NuScale Plant Module using the NRELAP5 code, and include directions on how to select code options, nodalizing the system, and selecting heat transfer correlations. The NuScale NRELAP5 non-LOCA plant transient model described in this section is considered typical and representative of the methodology. The base model for a specific NPM design reflects the parameters and features of that design. For example, an NPM design with fewer RVVs uses fewer NRELAP5 components to model the RVVs. Such NPM design differences do not affect the method used to develop the base model. 6.1 Thermal-Hydraulic Volumes and Heat Structures The NRELAP5 plant model contains multiple hydraulic components, heat structures and junctions. The model simulates the majority of a typical NPM (Figure 6-1) including the RPV and internals, the containment, and the reactor cooling pool. ((

                     }}2(a),(c) Both DHRS trains are included in the model, along with the ECCS consisting of the RVVs and RRVs. Control system components include variable and logical trips, control blocks and general tables. Figure 6-2 shows a typical nodalization diagram for the primary and secondary systems and is meant to convey the overall model structure rather than show nodalization details of any particular component.

Figure 6-3 shows a cut-away of the typical NPM reactor coolant system and CNV with the key nodalization regions included in the NRELAP5 model. The circled numbers in the figure represent RCS fluid regions and the numbers in squares represent containment regions. Table 6-1 lists the typical RCS regions and the associated NRELAP5 components. The NRELAP5 model of an NPM serves as the standalone baseline model for non-LOCA safety analysis, as well as various aspects of plant design support. The information presented herein describes a typical base model as it is configured for non-LOCA analysis. ((

                                                                                                  }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                                        }}2(a),(c)

Where precise noding is described in Section 6.1, the specified level of detail is considered the minimum level of detail required for the component of interest. Should additional detail be needed in the future, the relevant benchmarks, sensitivity studies, and transient analyses will be reviewed for continued applicability. If necessary, the relevant benchmarks, sensitivity studies, and transient analyses will be revised to demonstrate the higher level of detail for the component of interest is applicable to an NPM. Figure 6-1 NuScale Power Module (typical) © Copyright 2022 by NuScale Power, LLC 410

Figure 6-2 Typical primary and secondary side model (heat structures and component cell

     © Copyright 2022 by NuScale Power, LLC details excluded)

(( Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

                                                                                                                                             }}2(a),(c) 411

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-3 Typical NRELAP5 plant module volume regions ((

                                                                                                }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 6-1 Typical reactor coolant system regions and associated NRELAP5 components ((

                                                                                              }}2(a),(c) 6.1.1         Reactor Primary Reactor Pressure Vessel Downcomer

((

                                                 }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-4 Typical reactor pressure vessel downcomer model ((

                                                                                                  }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Steam Generator Primary The SG is a helical coiled, once-through HX, with the primary system on the shell side and the secondary system on the tube side. The primary system water is cooled as it flows over the outer surfaces of the SG tubes before passing over the feedwater plena that are located above the elevation of the conical transition riser fairing in the RPV. On the secondary side, feedwater enters the bottom of the SG tubes via the feedwater plena and is heated as it flows upward, with superheated steam exiting the tops of the tubes. Two independent sets of interwoven SG tube banks occupy the SG region, each having independent feedwater and steam plena. If the tube banks experience different secondary side conditions, the primary coolant does not experience any corresponding asymmetries because of the interwoven design of the helical coiled tubes. The heat transfer to the SG tubes occurs in the upper downcomer. The heat transfer and pressure drop resulting from the presence of the SG tubes was assessed with data from SIET TF-2 (Section 5.3.5). This assessment used special heat transfer model options to determine the methodology for accurately predicting the pressure drop in the region. ((

                                       }}2(a),(c)

Core and Lower Plenum ((

                                                                                                  }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

              }}2(a),(c)

Figure 6-5 Typical core and lower plenum model ((

                                                                                                    }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-6 Reflector / core bypass without fuel assemblies (for illustration only) Lower Riser ((

                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-7 Lower riser region, immediately above the core (for illustration only) © Copyright 2022 by NuScale Power, LLC 418

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-8 Typical reactor pressure vessel core and lower riser model ((

                                                                                                }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Upper Riser ((

                                                       }}2(a),(c) Normal flow in the riser is single-phase subcooled water. Transients that involve RPV depressurization or inventory loss can result in flashing and two-phase flow in the riser region.

((

                                                    }}2(a),(c)

Figure 6-9 Typical reactor pressure vessel upper riser model ((

                                                                                                       }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Pressurizer ((

                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-10 Typical reactor pressure vessel pressurizer model ((

                                                                                                 }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 6.1.2 Core Kinetics The separable point kinetics model is used to calculate reactivity feedback to the core power from the moderator, the fuel, and decay heat. The various point kinetic parameters are input based on the fuel burnup (for example, new core, BOC, or EOC) and control rod insertion amounts assumed for the analysis. NRELAP5 assumes an infinite core operating time at the initialized power when determining the decay heat power. The fission product decay type is specified as 'gamma-ac' with the 'ans73' model, which calculates decay heat in accordance with the 1973 ANS standard while adding the contribution from actinides. A fission product yield factor of 1.0 is specified in the base models, which can be changed to suit the scenario being analyzed. Control variable inputs to the core kinetics model are used to simulate control rod movement and reactivity feedback from a variety of parameters including, but not limited to fuel temperature, moderator temperature, and moderator density. Trips are used to enable or disable each reactivity feedback mechanism as needed. The fuel and moderator feedback controllers utilize volume weighting of the fuel and moderator temperatures to specify bounding reactivity feedback input; volume weighting is used for consistency with how the reactivity feedback design limits are confirmed in the core design. A scram table is used to simulate control rod insertion following reactor trip. Appropriately conservative scram curves are developed based on the core time-in-life, the initial power level, the location of the control bank, and other relevant factors to preserve the minimum shutdown margin. 6.1.3 Fuel Rod Design Input ((

                                               }}2(a),(c) Fuel performance data (Section 4.3.1.2) is incorporated into the NRELAP5 non-LOCA model via material thermal property tables for the UO2 fuel region, the gas gap, and the cladding. Because the density of UO2 changes with burnup, the thermal property tables can be revised as needed to match time in cycle.

As discussed in Section 4.3.1.1, the core power distribution is based on a nominal average axial power shape with power distributed solely in the fuel pellet. ((

                                                                }}2(a),(c) The gap thermal conduct calculated at bounding values burnup to ensure that the fuel volume average

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 temperature is appropriately bounded compared to fuel performance design data. This bounding method also accounts for gap closure over the fuel cycle. 6.1.4 Secondary System 6.1.4.1 Feedwater System In an NPM design two feedwater lines penetrate the CNV immediately downstream of the FWIVs. Each feedwater line splits into two lines before connecting to the SGs. ((

                                  }}2(a),(c) 6.1.4.2            Steam Generator Secondary The NRELAP5 specific helical coil SG component ('hlcoil') is used to simulate the helical coil SG that is characteristic of an NPM (Reference 2 describes the helical coil component). The typical steam generator nodalization is shown in Figure 6-11. The primary coolant flows through the SG shell side while the feedwater and steam flow through the tube side. The tube and shell side of the SG elevation nodalization schemes are one-to-one. The SG nodes are uniform, or may be finer towards inlet and coarser towards the exit if necessary to capture the

© Copyright 2022 by NuScale Power, LLC 424

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 phase change process in the lower region of the tubes. For non-LOCA transient calculations, finer nodalization is not needed near the tube exits due to the state of the fluid being single phase vapor. Section 5.3.5.4 summarizes SG nodalization sensitivity calculations. Based on these studies, modeling the helical coil SG with (( }}2(a),(c) nodes is expected to produce reasonably accurate results for the non-LOCA transients. Should additional detail be needed in the future, the relevant benchmarks, sensitivity studies, and transient analyses will be reviewed for continued applicability and updated as necessary to demonstrate the higher level of detail for the component is applicable to an NPM. ((

                                                                          }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-11 Typical steam generator model ((

                                                                                                          }}2(a),(c) 6.1.4.3            Main Steam System

((

                                }}2(a),(c) The typical MSS nodalization is shown in Figure 6-12.

((

                         . }}2(a),(c) DHRS modeling is discussed in Section 6.1.5.

((

                                                                                                         }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                         }}2(a),(c)

Figure 6-12 Typical main steam system model ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 6.1.5 Decay Heat Removal System An NPM incorporates two separate DHRSs that are treated individually in the NRELAP5 model. ((

                                     }}2(a),(c) Figure 6-13 shows the typical nodalization for DHRS loop

1. Loop 2 is modeled similarly. While each DHRS line in an NPM features two parallel actuation valves, ((

                            }}2(a),(c)

The number of hydrodynamic volumes in the DHRS piping and HX regions are based on results from NRELAP5 assessments using data from the NIST-1 facility (Section 5.3.2). ((

                                  }}2(a),(c)

In an actual NPM, the DHRS heat exchanger is located in the reactor cooling pool. ((

                              }}2(a),(c) The long-term use of DHRS is addressed separately in Reference 26.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-13 Typical decay heat removal system division 1 model ((

                                                                                                 }}2(a),(c)

Figure 6-14 Not used. © Copyright 2022 by NuScale Power, LLC 429

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 6.1.6 Emergency Core Cooling System The ECCS hydrodynamic components consist of two reactor recirculation valves (RRVs) and, depending on the NPM design, either two or three reactor vent valves (RVVs). ((

                                   }}2(a),(c)

Figure 6-15 Not used. 6.1.7 Containment Vessel ((

                                                                                                      }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                  }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 6-16 Typical containment and reactor pool model ((

                                                                                                 }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 6.1.8 Reactor Cooling Pool ((

                           }}2(a),(c) 6.2      Material Properties Thermal properties (thermal conductivity and volumetric heat capacity) are specified by user input in the NRELAP5 non-LOCA base models for the following materials typically used in the heat structures. These material properties may be amended or revised as the NPM design evolves:

1. fuel cladding (AREVAs M5 cladding)

2. inconel 690 (SG tubes)

3. uranium dioxide (UO2)

4. stainless steel (SA-240 304L)

5. fuel-to-cladding gas gap (initially pressurized helium at BOC; mixture of fission product gases and helium after irradiation)

6. carbon steel (SA-508)

7. martensitic stainless steel (SA-336, F6NM)

8. austenitic stainless steel (SA-965, FXM-19)

9. ((

                                                                                           }}2(a),(c) 6.3      Control Systems With its combination of trips, control functions, and user-defined tables, NRELAP5 provides flexibility to accurately simulate plant control and protection system responses during both steady-state and transient operation. The NRELAP5 non-LOCA base models contains logic for normal controls that simulate normal operational plant response, as well as user-convenience controls that make it easier to initialize the model for particular transients and easier to interpret the transient results. It also contains trip and control

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 logic that accurately simulates the MPS, i.e., the safety-related trips that protect the reactor core and fission product boundaries. 6.3.1 Module Control System (Nonsafety-related) The MCS as implemented in the NRELAP5 non-LOCA model provides control systems that may be used for model initialization and for simulating normal operating transients. These control systems can be used to initialize the model at new steady state conditions or to evaluate the nominal MCS response to a transient initiator. The MCS model allows the simulation of a variety of prototypic control scheme responses to transient conditions to provide nominal responses of these control schemes. The model also includes the capability to disable these control schemes to allow for conservative modeling applications as necessary. 6.3.1.1 Pressurizer Pressure Control (Nonsafety-related) The RPV pressure is controlled via the use of pressurizer heaters and spray to maintain the pressurizer steam pressure at a target value. When the pressure drops, the power to the PZR heaters increases, and if pressure increases, a portion of the CVCS recirculation flow is diverted to the pressurizer spray nozzles to collapse the steam space via condensation and reduce pressure. 6.3.1.2 Chemical Volume Control System Control (Nonsafety-related) The control systems implemented in the NRELAP5 non-LOCA model include simplified mechanisms for controlling CVCS recirculation flow, injection temperature and RCS inventory control (makeup and letdown). ((

                            }}2(a),(c) The water level in the pressurizer is controlled to a programmed setpoint by operation of the CVCS makeup pumps and the letdown control valve.

Operation of CVCS letdown may either occur automatically or with operator permission, depending on the design. Operator permission is required for CVCS makeup to be initiated. For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. 6.3.1.3 Reactor Coolant System Temperature Control (Nonsafety-related) The MCS model controls RCS average temperature by changing reactivity in the core to increase or decrease core power. This reactivity change is accomplished © Copyright 2022 by NuScale Power, LLC 434

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 with a control rod controller and a boron controller. The control rod controller uses design data to model a calculated rate of reactivity insertion due to maximum or nominal rod movement rates. ((

                                    }}2(a),(c) Neither controller accounts for all the actual core physics including the effect of xenon or other decay products or poisons that could be expected with control rod repositioning.

The average coolant temperature is controlled by adjusting core power, which is accomplished by moving the control rods or changing the boron concentration of the reactor coolant. The choice of which method is based on the desired rate of change for core power. The control rods are moved to achieve faster power changes to meet the target average coolant temperature; slower power changes are accomplished by changing the boron concentration of the reactor coolant. At full power, the rod control system is set to insert only mode to prevent automatic withdrawal of the control rods during a transient. 6.3.1.4 Steam Pressure Control (Nonsafety-related) In an NPM design, the turbine throttle and bypass valves are used to control steam pressure at the programmed values, ((

                                           }}2(a),(c) 6.3.1.5            Feedwater and Turbine Load Control (Nonsafety-related)

An NPM prototypic control scheme design for the feedwater system is based on turbine load demand. The feedwater pumps are variable speed and can provide variable flow for module operations over a wide range of power without adjustments to the feedwater regulating valve. ((

                                                }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                                                                     }}2(a),(c) 6.3.1.6            Containment Pressure Control (Nonsafety-related)

The containment pressure is established at sub-atmospheric conditions via operation of the containment evacuation system. The impact of this system continuing to operate is considered for the non-LOCA transient analyses. 6.3.2 Module Protection System (Safety-related) 6.3.2.1 Analytical Limits and Delays The MPS implemented in the NRELAP5 base models is intended for the purposes of performing safety analysis transient simulations. As such, the logic and actuation points are based on the NPM safety analysis analytical limits. Fixed delay times are specified considering different sensor response times. ((

                                    }}2(a),(c) In addition to the sensor delays, a given safety signal is subject to instrumentation string delays, an MPS processing delay, and an actuation delay. The NRELAP5 non-LOCA model incorporates the methodology assumption that a bounding total for these additional delays is applied as a signal delay in addition to the individual sensor delay.

The MPS actuation signals in an NPM design are typically based on the following types of parameters: power range power power range power rate intermediate range log power rate source range count rate source range log power rate RCS flow RCS temperature pressurizer pressure pressurizer level RPV riser level containment pressure containment water level © Copyright 2022 by NuScale Power, LLC 436

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 main steam pressure main steam superheat These parameters can be monitored in NRELAP5 to credit MPS actuations in the safety analysis. A single parameter may be used for multiple MPS actuations. For example, a high pressurizer pressure signal could result in reactor trip, containment isolation, and DHRS actuation. The MPS actuation signals may vary by NPM design and may be incorporated into the NRELAP5 base model for the design.The control logic for other MPS signals may be added if needed for a particular event analysis or as necessary to maintain consistency with the MPS design. 6.3.2.2 NRELAP5 Modeling The NRELAP5 components used to model the MPS include variable and logical trips that are used to sense the MPS trip limits, apply signal delays and generate actuation signals. The MPS instrumentation and sensors are typically modelled by comparing the volume or junction parameter at the location of the sensor and comparing it to the associated analytical limit, rather than explicitly modeling the sensor. The reactor trip system (RTS) logic is fairly simple, as each RTS signal is computed and compared against the trip setpoint on each timestep. The trips are organized together in a series of or logical trips. Should one trip limit be reached, the cascade of logical trips immediately reach the final RTS logical trip, which then causes the RPS actuation signal to become true after a fixed delay that conservatively accounts for signal processing and rod latch mechanism delays. Other subsystems such as containment isolation, DHRS actuation, ECCS actuation, etc. are modeled similarly. ((

                                                   }}2(a),(c)

The pressurizer level signal is generated by modeling the collapsed liquid level, ((

                                                                              }}2(a),(c)

The ECCS actuation logic includes the ability for the user to set additional electrical or mechanical conditions that are external to the NRELAP5 hydrodynamic model. ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 (( }}2(a),(c) The effects of external conditions can be added to any of the control system models as needed for a given transient scenario. © Copyright 2022 by NuScale Power, LLC 438

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.0 Non-LOCA Analysis Methodology 7.1 General The discussion in the following sub-sections is applicable for all transients unless the event-specific methodology states otherwise in Section 7.2. 7.1.1 Achieving Steady State Conditions This section identifies the initial and boundary conditions considered for biasing in non-LOCA analyses, including prioritization during the initialization process. While the majority of parameters identified herein are initial conditions relevant to the steady state, other parameters are considered as bounding input for an NPM response during the transient progression. 7.1.1.1 Background Establishing the appropriate initial conditions is paramount to obtaining an appropriate plant response for the transient of interest. To this end, it is important to ensure the NRELAP5 model achieves a valid steady state prior to initiating the transient. The effect of various initial conditions on the response to a specific acceptance criterion is assessed for each non-LOCA transient. Several means are available to perform the assessment. For example, the assessment may consist of a combination of: Qualitative engineering assessment in the calculation to identify why an initial condition bias is limiting or non-limiting Quantitative assessment by execution of appropriate sensitivity calculations in the calculation Reference to applicable regulatory precedent that identifies why a particular initial condition bias is not limiting for a particular transient or type of transient, or why a nominal condition is appropriate 7.1.1.2 Identification of Relevant Parameters A list of initial conditions to be considered for biasing was developed for the non-LOCA analyses, including the target value for the initial plant condition and acceptable tolerance to the target value. Other parameters needed to obtain a steady state include: certain fuel-related and core-related inputs; measurement uncertainties for various safety-related processes; plant operational limits assumed in the safety analyses; and the DHRS initial conditions. © Copyright 2022 by NuScale Power, LLC 439

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-1 identifies a typical list of initial conditions considered for non-LOCA analyses. This list of initial conditions includes both direct inputs and calculated results. Most of these parameters are directly input to NRELAP5. The calculated results are identified as a target because the analyst uses this parameter as a target during the initialization process. Under best-estimate primary flow conditions and various reactor powers, it has been shown that there is relatively little variation in liquid inventory in the steam generator. The inventory does not change appreciably when considering variations in primary coolant temperature and primary flow rates. Consequently, steam generator inventory is not a target during the initialization process. ((

                                                                 }}2(a),(c)

For reactor powers greater than or equal to 20 percent full power (FP) and best-estimate primary flow conditions, the steam superheat increases with power for the defined main feedwater conditions. ((

                                                                                   }}2(a),(c)

The removal of any restrictions regarding steam generator inventory and steam superheat allows use of the parameters in Table 7-1 for initialization without over-specifying or under-specifying the problem. © Copyright 2022 by NuScale Power, LLC 440

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-1 Typical list of initial conditions considered Region Parameter Core Power Calorimetric power uncertainty Volume-weighted core average fuel temperature (target) Moderator temperature coefficient Doppler temperature coefficient Effective delayed neutron fraction Ratio of effective delayed neutron fraction to prompt neutron lifetime 238U neutron capture to fission ratio Energy deposition factor Average core axial power shape Scram reactivity (shutdown margin) Control rod bank differential worth Decay heat and decay heat uncertainty Boron concentration RCS Primary system mass flow rate (target) RCS average fluid temperature (target) Pressurizer pressure Pressurizer level Total core bypass flow rate Pressurizer spray bypass flow rate Pressurizer heater power Heat losses to containment from the RPV and from piping inside containment Steam generator secondary, Feedwater mass flow rate feedwater system, and MSS Feedwater temperature Steam chest pressure (turbine header pressure) Steam generator tube plugging, fouling factor Containment Pressure Heat losses from containment to reactor pool and reactor building Temperature Reactor pool Temperature Level DHRS Liquid volume 7.1.1.3 Prioritization of Initial Conditions An important design feature of NPMs is the natural circulation of primary reactor coolant. This design feature may also limit the ability of the NRELAP5 non-LOCA transient model to achieve a given set of initial conditions. In particular, changes in core power or parameters that affect the SG secondary side heat removal rate alter the reactor coolant flow rate and density distribution in the primary coolant. The initial conditions are prioritized based on the greatest impact to the transient of interest. © Copyright 2022 by NuScale Power, LLC 441

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Initial conditions may be treated in the NRELAP5 non-LOCA transient model in different ways. For instance, the initial conditions can be specified directly as input or calculated by the code. Parameters identified as important initial conditions for a particular transient (whether directly input or indirectly determined by the code) are checked as part of the steady state balance to confirm that either: 1) the parameter is within the allowable tolerance to the target value based on design references; or, 2) the parameter conservatively bounds the target design value and is adequately steady (within acceptable tolerances). If the initial condition is calculated by the code but not identified as important for a transient, the parameter may or may not be directly checked as part of the steady state initialization process against design values. 7.1.1.4 Typical Initialization Process Prior to the initialization process, certain parameters critical to establishing the correct steady state for the event of interest are identified. Once the parameters of interest achieve steady state target values within acceptable tolerances, on the basis of engineering judgement, ((

                                                                  }}2(a),(c) A steady state solution is achieved when the change in the target value during the loop transits is within the variance band described for each parameter.

A typical list of critical parameters for initializing the primary and secondary systems is provided below. reactor power: tolerance (or bounded) and variance fuel temperature: tolerance (or bounded) and variance RCS temperature: tolerance (or bounded) and variance PZR pressure: tolerance (or bounded) and variance PZR level: tolerance (or bounded) and variance RCS flow: tolerance (or bounded) and variance steam pressure: tolerance (or bounded) and variance feedwater flow rate: variance feedwater temperature: tolerance (or bounded) and variance Null transients are used to ensure the re-initialization cases achieve the desired steady state conditions. ((

                                                                                                 }}2(a),(c) In addition, feedback from the point kinetics model is active for the null transients.

7.1.2 Treatment of Plant Controls Control systems are necessary to maneuver an NPM within the power range in accordance with normal operating transients. However, not all of these control © Copyright 2022 by NuScale Power, LLC 442

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 systems are relevant to the non-LOCA transient analyses because the action of the control system does not alter the event consequences. This section identifies the normal, nonsafety-related plant controls considered in non-LOCA transient analyses. The effect of various plant controls on the response to a specific acceptance criterion is assessed for each non-LOCA transient. Several means are available to perform the assessment. For example, the assessment may consist of a combination of: Providing a qualitative engineering assessment that identifies why operation of a PCS is limiting or non-limiting. Performing a quantitative assessment via appropriate sensitivity cases. Referencing to applicable regulatory precedent that identifies why a particular normal plant control is not limiting for a specific transient or type of transient. When considering operation of the various plant controls, the approach is based on the event consequences for a given acceptance criterion. Specifically, if operation of the control system leads to a less severe plant response, then the actions of the control system are not simulated for the transient of interest. Conversely, if operation of the control system causes the event consequences to be more severe, the PCS is assumed to operate as designed. If operation of the control system has minimal or no impact on the event consequences, then control system operation may either be modeled or not modeled. ((

                      }}2(a),(c)

The normal PCSs to be considered in design basis event analysis are: Pressurizer Pressure Control Pressurizer pressure is controlled via operation of the pressurizer spray and the pressurizer heaters. Pressurizer Water Level Control The water level in the pressurizer is controlled to a programmed setpoint by operation of the CVCS makeup pumps and the letdown control valve. Operation of CVCS letdown may either occur automatically or with operator permission, depending on the design. Operator permission is required for CVCS makeup to be initiated. For non-LOCA transient analyses where loss of inventory or inventory shrinkage is expected to indicate makeup is needed, the process by which the operator approves makeup is considered. Since approval to initiate makeup is a manual action by the operator, the non-LOCA transient analyses do not credit this action. For spurious inventory addition events where letdown could reduce the event consequences, credit for this action is not taken for the non-LOCA transient analyses. © Copyright 2022 by NuScale Power, LLC 443

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Core Average Coolant Temperature Control The average coolant temperature is controlled by moving the control rods or changing the boron concentration of the reactor coolant. This selection is based on the desired rate of change for core power. Hence, the control rods are moved to achieve faster power changes to meet the target average coolant temperature; slower power changes are accommodated by changing the boron concentration of the reactor coolant. At full power, the rod control system is set to insert only mode to prevent automatic withdrawal of the control rods during a transient. Steam Pressure Control Steam pressure is controlled to the desired value using the turbine throttle valves or the turbine bypass valves. The effect of these valves to a change in steam pressure is considered for the non-LOCA transient analyses. Turbine Load Control The mass flow rate and pressure provided by the feedwater pump is used to meet the desired turbine load, which reflects the power generation rate. The impact of the feedwater pumps continuing to operate until the feedwater line is isolated is considered for the non-LOCA transient analyses when DHRS is actuated. Containment pressure control The containment pressure is established at sub-atmospheric conditions via operation of the containment evacuation system. The impact of this system continuing to operate is considered for the non-LOCA transient analyses. 7.1.3 Loss of Power Conditions This section defines the term loss of normal power as applied to an NPM; describes the various power supplies (AC and DC); and, explains how the loss of these power supplies is treated by the non-LOCA transient analyses. 7.1.3.1 Background Chapter 15 of the SRP (Reference 15) does not ordinarily consider a loss of offsite power for events that require a malfunction of an active system for which power must be available; however, exceptions are made for some reactivity initiated events. The role of offsite power is less defined for an NPM plant than for traditional plants for several reasons, but the use of natural circulation for normal operation and safety systems is the fundamental reason. Consequently, these design features limit the impact of a power loss to an NPM plant compared to a traditional plant design that relies on forced circulation. The term loss of normal power is used herein because the normal power source for an NPM plant is not the offsite grid, and there are no onsite safety-related © Copyright 2022 by NuScale Power, LLC 444

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 power generating sources (AC or DC). Although NPM plant designs do not include a safety-related electrical power generation source, the designs include two different DC power systems, both with a battery backup. This battery backup allows the DC power system that supplies MPS to provide uninterrupted DC power for 72 hours of MPS operation. Because the MPS also provides power to keep the ECCS valves closed, if a loss of normal AC power occurs, the MPS sheds the load to the ECCS valves after 24 hours. The load shedding allows the ECCS valves without IAB to open at that time; ECCS valves with IAB will open when the pressure difference between the RPV and the containment is below the IAB threshold. The effect of various power availability scenarios on the response to a specific acceptance criterion is assessed for each non-LOCA transient. This assessment can be performed using several methods. For example, the assessment may consist of a combination of: providing a qualitative engineering assessment that identifies why operation of a power availability scenario is limiting or non-limiting performing a quantitative assessment via appropriate sensitivity cases referencing to applicable regulatory precedent that identifies why a particular power availability scenario is not limiting for a specific transient or type of transient When considering various power availability scenarios, the approach is based on the event consequences for a given acceptance criterion. Because nonsafety-related AC power sources are not credited for the non-LOCA event analyses, in all scenarios where AC power is lost, the ECCS valves open at some point during the transient. For non-LOCA transients that do not demand ECCS actuation during the short-term event progression, i.e., during the time frame in which the peak pressures, MCHFR, or peak fuel centerline temperature occur, the time at which the ECCS valves open is beyond the scope of this methodology document. Analysis of ECCS valve opening events is addressed using the Reference 2 methodology. 7.1.3.2 Consideration of Loss of Power to All Transients The subsections of SRP Chapter 15 (Reference 15) applicable to an NPM, and the Chapter 15 DSRS sections, were reviewed to identify which subsections explicitly address a loss of power or GDC 17 (Reference 4). This review concluded that given the differences between the design of an NPM and typical PWRs, particularly with respect to the safety system design and response, the effect of loss of power on each non-LOCA transient should be considered. 7.1.3.3 Electrical Systems with Important Loads A review of the electrical systems applicable to the NPM designs was undertaken to identify which electrical systems have loads important to the non-LOCA © Copyright 2022 by NuScale Power, LLC 445

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 transient analyses. Consideration was given to both direct and indirect impact to the transient analyses. A direct impact to the transient analyses is one that alters the primary or secondary side conditions if the electrical system is lost. An indirect impact to the transient analyses is one that affects the ability to monitor the system or affects a connecting system. The following typical electrical systems have loads important to the non-LOCA transient analyses. EHVS - High voltage (13.8 kV) AC electrical distribution system EMVS - Medium voltage (4.16 kV) AC electrical distribution system ELVS - Low voltage (480 V and 120 V) AC electrical distribution system EDNS - Normal DC power system EDSS - Highly reliable DC power system EDAS - Augmented DC power system The NPM designs utilize either an EDAS or EDSS, depending on the design. The EDNS and the EDAS/EDSS power systems are both designed with battery backups to allow a continuous power supply in the event the power supply to the chargers is not available. The duration of the battery backup is based on the relative importance of the loads. Because the loads supplied by the EDNS are of lower importance (i.e., all loads are non-essential) the batteries are typically sized to supply these loads for 40 minutes. In contrast, for the higher importance (essential) loads supplied by the EDAS/EDSS, with the exception of the ECCS valves, the batteries are sized to supply these loads for greater than 24 hours. The ECCS valves are unique because the MPS acts to shed the load for these valves at 24 hours after the loss of AC power to the EDAS/EDSS battery chargers. The loss of normal AC power to the EDAS/EDSS chargers also causes the MPS to initiate a reactor trip, actuate DHRS, and close the containment isolation valves. Based on the electrical system design for an NPM, a loss of normal AC power means a loss of power from the ELVS. Such a condition could be due to: 1) a failure within the ELVS; or, 2) a loss of power from the EHVS or EMVS. Since a failure of the EHVS or EMVS results in the same response from the electrical systems for non-LOCA transient analyses, these failures are not considered separately. With respect to the battery backups for the DC power supply systems, the availability of the battery backups is considered because the EDNS and EDAS/EDSS are not safety-related systems. © Copyright 2022 by NuScale Power, LLC 446

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.1.3.4 Timing for Loss of Power The review of the subsections of SRP or DSRS Chapter 15 discussed in Section 7.1.3.2 indicates the SRP/DSRS is not prescriptive regarding when the loss of power occurs. An example is provided in DSRS Subsection 15.1.5, which states: Assumptions as to the loss of offsite power (LOOP) and the time of loss should be made to study their effects on the consequences of the accident. A LOOP could occur simultaneously with the pipe break or during the accident, or offsite power may not be lost. For the non-LOCA transient analyses applicable to an NPM, the loss of normal AC power is considered at two different times during the event progression: coincident with event initiation; or, coincident with turbine trip. The basis for considering a loss of power coincident with event initiation is the loss of normal AC power could be caused by the event initiator or as a result of the event. For example, the mass and energy exiting a ruptured steam pipe could damage equipment that results in a loss of normal AC power. The basis for considering loss of power coincident with turbine trip is because the loss of power is a consequence of the event. Specifically, the loss of AC power occurs as a result of a disruption to the electrical grid following a turbine trip. Because the electrical output for an NPM is much smaller (less than 100 MWe) than the electrical output for a typical large plant (approxiately 1000 MWe), the loss of an NPM is not anticipated to significantly disrupt the electrical grid. Regardless, the possibility of causing such a disruption is assessed as part of the non-LOCA transient analyses. The random loss of a nonsafety-related system during an event is not typically required as part of the evaluation, and is not being postulated for the NPMs. If the EDNS or EDAS/EDSS design demonstrates that its failure is not linked to failure of the normal AC power supply, then the timing of a loss of EDNS or EDAS/EDSS is considered independently of the loss of normal AC timing described above. Assessment of non-LOCA events with and without the EDNS or EDAS/EDSS is consistent with their design as nonsafety-related. 7.1.4 Single Failures This section discusses active single failures in fluid systems and single failures in electrical systems for consideration in non-LOCA transient analyses, as well as the relevance of fluid system passive single failures. © Copyright 2022 by NuScale Power, LLC 447

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.1.4.1 Background The effect of various single active failures on the response to a specific acceptance criterion is assessed for each non-LOCA transient. Several means are available to perform the assessment. For example, the assessment may consist of a combination of: Providing a qualitative engineering assessment that identifies why a single failure is limiting or non-limiting. Performing a quantitative assessment via appropriate sensitivity cases. Referencing to applicable regulatory precedent that identifies why a particular single failure is not limiting for a specific transient or type of transient. When considering various single active failures, the approach is based on the event consequences for a given acceptance criterion. Only safety-related components are considered for possible single active failures. A nonsafety-related component whose action benefits the transient consequences is typically assumed inactive, unless the component is demonstrated to not be affected by the initiating event or the component is acting as backup protection. This treatment of nonsafety-related components is consistent with RG 1.206 (Reference 14 Section C.I.15.6.2), which states: Only safety-related systems or components should be used to mitigate transient or accident conditions. However, analyses may assume that nonsafety-related systems or components are operable for the following cases: (1) when a detectable and nonconsequential random and independent failure must occur in order to disable the system (2) when nonsafety-related components are used as backup protection For any nonsafety-related systems or components credited in the design-basis analyses for mitigating the event consequences, the applicant must provide proper justification. The application must take into account nonsafety-related systems or components that may adversely affect transient or accident analyses. Safety-related SSC are defined in 10 CFR 50.2 (Reference 16), as those SSC:

                      ... that are relied upon to remain functional during and following design basis events to assure:

1) The integrity of the reactor coolant pressure boundary;

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or

3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 10 CFR 100.11, as applicable.

A limited number of safety-related SSC interact with the reactor coolant system, secondary system, or containment. In most instances these safety-related SSC are considered using the definition presented in ANSI/ANS-58.9-2002 R2015 (Reference 17), i.e., an active failure is a malfunction of a component that relies on mechanical movement to complete its intended function upon demand. Passive failures of fluid systems are not considered for the non-LOCA transient analyses during the short term (up to 24 hours). 7.1.4.2 Consideration of Single Failures An NPM considers single active failures of fluid systems or single failures of electrical systems that could affect the non-LOCA transient analyses. Various means are used to identify these failures, including: 1) the application of design criteria and industry standards to the design of the equipment; 2) the arrangement and type of valves incorporated into the design; and, 3) the level of redundancy in equipment functionality incorporated into the design. 7.1.4.3 Consideration of Passive Single Failures An NPM does not consider passive single failures of fluid systems during the short term (up to 24 hours) non-LOCA transient analyses; however, passive single failures of electrical systems are considered. This approach is consistent with the regulatory position in SECY-94-084 (Reference 18) and SRP Chapter 15 (Reference 15). 7.1.4.4 Single Failures to Evaluate An NPM considers the worst active single failure in conjunction with the initiating event and with loss of power for the short term non-LOCA transient analyses. During the short term, passive single failures of fluid systems are not considered; however, passive single failures of electrical systems are considered. The following active single failures are considered for non-LOCA transient analyses. safety-related main steam isolation valve - failure to close when demanded safety-related main steam isolation bypass valve - failure to close when demanded safety-related feedwater isolation valve - failure to close when demanded © Copyright 2022 by NuScale Power, LLC 449

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 safety-related feedwater check valve - failure to close when demanded MPS - failure of a single channel to trip when demanded (e.g., for asymmetric reactivity events where trips based on asymmetric core response are credited for mitigation) ECCS reactor recirculation valve - failure to open when demanded (mechanical failure prevents opening on low P) ECCS reactor vent valve - failure to open when demanded (mechanical failure prevents opening on low P) Depending on the design, the following single failure may be active or passive electrical, but is considered for non-LOCA transient analyses. MPS - failure to signal one ECCS reactor recirculation valve and one RVV to open when demanded (each valve still opens on low P) A failure of the MPS which causes spurious opening of EECS valves is not considered for non-LOCA transient analyses. The ECCS valves are either prevented from opening by IAB or their opening is analyzed using the Reference 2 methodology. 7.1.5 Bounding Reactivity Parameters This section describes the bounding reactivity parameters used for the non-LOCA transient analyses. A brief discussion of each parameter is provided in the following sections. Moderator Temperature Coefficient (Section 7.1.5.1) Doppler Temperature Coefficient (Section 7.1.5.2) Decay heat contribution (Section 7.1.5.3) Scram worth (Section 7.1.5.4) Reactivity versus time for scram insertion (Section 7.1.5.5) 7.1.5.1 Moderator Temperature Coefficient Bounding values for MTC are generally selected to represent the most positive value at BOC and the most negative value at EOC. For most non-LOCA transient analyses, the coefficient is conservatively applied as a constant value for the full range of expected moderator temperatures. If the MTC at zero power is positive, this value could be used to bound all power levels. Alternately, the positive MTC could be applied over a limited range of power levels with a different, yet bounding, MTC value for the remaining power levels. The following MTC values are examples for a representative NPM. A review of the applicable core physics parameters is performed each cycle to confirm the bounding nature of the values utilized for the non-LOCA transient analyses. © Copyright 2022 by NuScale Power, LLC 450

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Most Positive at power > 25 percent RTP = 0 pcm / degrees F Most Positive at power 25 percent RTP = +6 pcm / degrees F Most Negative at hot full power = - 43.0 pcm / degrees F Most Negative at hot zero power = - 15.0 pcm / degrees F 7.1.5.2 Doppler Temperature Coefficient Bounding values for DTC are generally selected to represent the least negative value at BOC and the most negative value at EOC. For most non-LOCA transient analyses, the coefficient is conservatively applied as a constant value for the full range of expected fuel temperatures. The following DTC values are examples for a representative NPM. A review of the applicable core physics parameters is performed each cycle to confirm the bounding nature of the values utilized for the non-LOCA transient analyses. Minimum (least negative) = - 1.40 pcm/°F Maximum (most negative) = - 2.25 pcm/°F 7.1.5.3 Decay Heat Contribution Energy production in the core comes from fission and fission product decay. This latter component is the decay heat contribution. During the initialization process the specified kinetics inputs are used to determine the balance between fission power and fission product decay. Changing the decay heat contribution, therefore, changes the fission power because the total core power is not altered. Bounding values for decay heat are designated to represent the high contribution and the low contribution. Once specified, the decay heat contribution is utilized for the duration of the event of interest. For the non-LOCA transient analyses the decay heat contribution is based on the 1973 ANS decay heat standard, which is varied by utilizing different decay heat multipliers and specifying whether or not to include the actinide contribution. The following decay heat contribution values are examples for a representative NPM. Figure 7-1 provides an example of the decay heat as a function of time for an equilibrium cycle. A review of the applicable core physics parameters is performed each cycle to confirm the bounding nature of the values utilized for the non-LOCA transient analyses. Low = use multiplier of 0.8 while excluding the actinide contribution High = use multiplier of 1.0 while including the actinide contribution © Copyright 2022 by NuScale Power, LLC 451

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 7-1 Example of decay heat comparisons 20 18 16 14 Decay Heat Power (MWt) 12 10 8 6 4 2 0 1.0E-1 1.0E+0 1.0E+1 1.0E+2 1.0E+3 1.0E+4 1.0E+5 1.0E+6 1.0E+7 Time (seconds) ORIGEN Best Estimate 80% of 1973 w/o Actinides 100% 0f 1973 w/ Actinides 7.1.5.4 Scram Worth After a reactor trip, the insertion of the control and safety banks imposes negative reactivity on the core. The reactivity available for insertion varies based on the time-in-life and the location of the control bank prior to trip. The inserted reactivity is sufficient to not exceed SAFDLs under conditions of normal operation, including AOOs; and, to limit fuel damage during accidents to assure the capability to cool the core is maintained. As applied to the non-LOCA transient analyses, the scram worth is the magnitude of the reactivity needed to take the initial core power to the desired shutdown conditions at zero power (for example, 420 degrees F) while, at a minimum, accounting for the moderator defect, the Doppler defect, the maximum worth stuck rod, and the minimum shutdown margin. 7.1.5.5 Reactivity Versus Time for Scram Insertion The manner in which the scram worth is inserted into the core is derived as a bounding depiction of the reactivity insertion characteristics associated with a © Copyright 2022 by NuScale Power, LLC 452

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 reactor trip from hot full power. An example of a normalized trip worth as a function of time, for which the control rods are fully inserted 2.278 seconds after being released by the control rod drive mechanisms, is presented in Table 7-2. Table 7-2 Example of normalized trip worth vs. time after trip Time After Trip (sec) Normalized Trip Worth 0.0 0.0 0.428 0.011 0.616 0.044 0.766 0.099 0.900 0.176 1.022 0.276 1.138 0.397 1.220 0.502 1.250 0.540 1.458 0.706 1.952 0.893 2.278 1.0 7.1.6 Biasing of Other Parameters This section describes the biasing of non-reactivity parameters used for the non-LOCA transient analyses. A brief discussion of each parameter is provided in the following sections. Initial Conditions (Section 7.1.6.1) Valve Characteristics (Section 7.1.6.2) Analytical Limits and Response Times (Section 7.1.6.3) 7.1.6.1 Initial Conditions The initial conditions assumed for the non-LOCA transient analyses are the most adverse with respect to the acceptance criterion of interest. These conditions are normally consistent with steady state operation, allowing for calibration and instrument errors and steady state fluctuations. Recognizing that the initial conditions do not contribute equally to the severity of the event consequences, alternate approaches may be used to set these conditions. For instance, bounding values may be used for certain parameters to provide a more restrictive response for a specific acceptance criterion. Alternately, nominal conditions may be used if the event consequences are insensitive to a specific initial condition. A general description of the biasing of initial conditions, as applied to the non-LOCA transient analyses for an NPM, is provided below. © Copyright 2022 by NuScale Power, LLC 453

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Initial Core Power The initial core power is biased high by an amount equal to the heat balance uncertainty. As an example, consider full power operation with a heat balance uncertainty of +/-2 percent RTP. For this situation, the indicated core power is less than the actual core power by 2 percent RTP, i.e., the actual core power is 102 percent RTP while the indicated power is 100 percent RTP. Initial Reactor Coolant System Average Temperature At operating conditions the initial RCS average temperature is biased to either end of the range centered at the nominal value after consideration for any control system deadband and system/sensor measurement uncertainty. As an example, consider a control system deadband and system/sensor measurement uncertainty of +/- 10 degrees F for RCS average temperature. For this situation, the initial RCS average temperature is set to either -10 degrees F or

                  +10 degrees F relative to the nominal temperature for the core power level of interest.

Initial Pressurizer Pressure The initial pressurizer pressure is biased to either end of the range centered at the nominal value after consideration for any control system deadband and system/sensor measurement uncertainty. As an example, consider a control system deadband and system/sensor measurement uncertainty of +/- 70 psi for pressurizer pressure. For this situation, the initial pressurizer pressure is set to either -70 psi or +70 psi relative to the nominal pressure for the core power level of interest. Initial Pressurizer Level The initial pressurizer level is biased to either end of the range centered at the nominal value after consideration for any control system deadband and system/sensor measurement uncertainty. As an example, consider a control system deadband and system/sensor measurement uncertainty of +/- 8 percent for pressurizer level. For this situation, the initial pressurizer level is set to either -8 percent or +8 percent relative to the nominal level for the core power level of interest. Initial Containment Pressure The initial containment pressure is biased to either end of the range expected for normal operation. © Copyright 2022 by NuScale Power, LLC 454

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 As an example, consider a normal operational range of 0.037 psia to 2.0 psia for containment pressure. For this situation, the initial containment pressure is set to either 0.037 psia or 2.0 psia. Initial Steam Generator Pressure The initial SG pressure is biased to either end of the range centered at the nominal value after consideration for any control system deadband and system/sensor measurement uncertainty. As an example, consider a control system deadband and system/sensor measurement uncertainty of +/- 35 psi for SG pressure. For this situation, the initial SG pressure is set to either -35 psi or +35 psi relative to the nominal pressure for the core power level of interest. Initial Feedwater Temperature The initial feedwater temperature is biased to either end of the range centered at the nominal value after consideration for any control system deadband and system/sensor measurement uncertainty. As an example, consider a control system deadband and system/sensor measurement uncertainty of +/- 10 degrees F for feedwater temperature. For this situation, the initial feedwater temperature is set to either -10 degrees F or

                  +10 degrees F relative to the nominal temperature for the core power level of interest.

Initial Reactor Coolant System Flow Rate The initial RCS flow rate is biased to either end of the range expected for normal operation. As an example, consider a normal operational range of 535 kg/s to 690 kg/s for RCS flow rate at 100 percent RTP. For this situation, the initial RCS flow rate is set to either 535 kg/s or 690 kg/s. Initial Volume Weighted Core Average Fuel Temperature The initial volume weighted core average fuel temperature is biased to either end of the range expected assuming limiting power histories, power shapes, and core burnups. As an example, consider a BOC range of 960 degrees F to 1065 degrees F for volume weighted core average fuel temperature. For this situation, the initial fuel temperature is set to either 960 degrees F or 1065 degrees F. © Copyright 2022 by NuScale Power, LLC 455

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Initial Reactor Pool Temperature The initial reactor pool temperature is biased to either end of the range expected for normal operation. As an example, consider a normal operational range of 40 degrees F to 200 degrees F for reactor pool temperature. For this situation, the initial reactor pool temperature is set to either 40 degrees F or 200 degrees F. 7.1.6.2 Valve Characteristics The valve characteristics assumed for the non-LOCA transient analyses may differ depending on the type of valve. These differences are based on providing the most conservative characteristics for the acceptance criterion of interest. A general description of the biasing of valve characteristics, as applied to the non-LOCA transient analyses for an NPM, is provided below. Pressure Relief Valves The characteristics applied to the pressure relief valves for the non-LOCA transient analyses include: 1) the nominal lift setpoint; 2) the lift tolerance; 3) the set pressure drift; 4) the accumulation; 5) the blowdown; 6) the blowdown tolerance; and, 7) the stroke time. In general, the characteristics are applied to delay opening a specific valve for as long as possible and to minimize the time the valve is open. Isolation Valves As an example, the following characteristics are applied to the isolation valves for the non-LOCA transient analyses. In general, the characteristics are applied to delay closing a specific valve for as long as possible. primary MSIV closing stroke time = 5 seconds secondary MSIV closing stroke time = 30 seconds FWIV closing stroke time = 5 seconds feedwater regulating valve closing stroke time = 30 seconds CVCS containment isolation valves closing stroke time = 5 seconds Other Valves As an example, the following characteristics are applied to the DHRS valves for the non-LOCA transient analyses. In general, the characteristics are applied to delay opening for as long as possible. DHRS open stroke time = 30 seconds © Copyright 2022 by NuScale Power, LLC 456

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 As an example, the following characteristics are applied to the feedwater check valves for the non-LOCA transient analyses. In general, the characteristics are applied to provide the most adverse characteristics for the acceptance criterion of interest. feedwater safety-related check valve closing stroke time = 1.0 second when reverse flow is more limiting; otherwise, a high reverse loss coefficient is employed feedwater nonsafety-related check valve closing stroke time = 1.0 second when used to stop flow following a single failure of the feedwater safety-related check valve As an example, the following characteristics are applied to the turbine valves for the non-LOCA transient analyses. In general, the characteristics are applied to close as fast as possible. turbine stop valve closing stroke time = 0.1 second turbine control valve closing stroke time = 0.15 second 7.1.6.3 Analytical Limits and Response Times The analytical limits are the limits of measured or calculated variables (such as pressures, temperatures, and levels) established by the safety analysis to assure that a safety limit is not exceeded. In the non-LOCA transient analysis, when an analytical limit is calculated to be reached, a reactor trip or engineered safeguards actuation is signaled, and then an appropriate response time is accounted for. The analytical limits and response times assumed for the non-LOCA transient analyses are selected to provide sufficient operating margin to prevent spurious actuation, yet still provide adequate protection for the acceptance criterion of interest. In general, the response times are applied to delay action while bounding the expected instrumentation delays. As an example, a diagram of the different types of limits and setpoints created to account for margin, allowances, and uncertainties is provided in Figure 7-2. © Copyright 2022 by NuScale Power, LLC 457

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 7-2 Example of setpoint relationships The RTS response time is defined as the elapsed time from when the process reaches the analytical limit until the control rods are free to fall. Thus, the response time includes the control rod drive mechanism delatch time. The ESFAS response time definition is similar, but does not include the control rod drive mechanism delatch time. Specifically, the ESFAS response time is defined as the elapsed time from when the process reaches the analytical limit until the actuation signal is received at the component (e.g. valve solenoid). Examples of the analytical limits and actuation delays are summarized in Table 7-3. The values in Table 7-3 are example values used to aid the subsequent methodology discussion. Application of the methodology to a specific NPM design requires the use of analytical limits and actuation delays associated with that NPM design. © Copyright 2022 by NuScale Power, LLC 458

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-3 Examples of analytical limits and actuation delays (reactor trip system and engineered safety features actuation system) Signal Analytical Limit Actuation Delay High power 25% RTP (power < 15% RTP) 2.0 seconds 120% RTP (power 15% RTP) 2.0 seconds High count rate 5.0E+5 counts/second 2.0 seconds Startup rate 3 decades/minute 31 seconds (source range) 3 decades/minute 2 seconds (intermediate range) High power rate +/-15% RTP/minute 2.0 seconds High RCS riser temperature 610°F 8.0 seconds High containment pressure 9.5 psia 2.0 seconds High pressurizer pressure 2000 psia 2.0 seconds High pressurizer level 80% 3.0 seconds Low pressurizer pressure 1720 psia 2.0 seconds Low low pressurizer pressure 1600 psia 2.0 seconds Low pressurizer level 35% 3.0 seconds Low low pressurizer level 20% 3.0 seconds Low steam pressure 300 psia 2.0 seconds Low low steam pressure 100 psia 2.0 seconds High steam pressure 800 psia 2.0 seconds High steam superheat 150°F 8.0 seconds Low steam superheat 0°F 8.0 seconds Low RCS flow 1.7 ft3/second 6.0 seconds Low low RCS flow 0.0 ft3/second 6.0 seconds Low RCS level 350-390 inches 3.0 seconds High CNV water level 220-260 inches 3.0 seconds Low AC voltage 0 VAC 60.0 seconds 7.1.7 Credit for Nonsafety-related Components or Operator Actions There are three occasions where nonsafety-related equipment is credited for event mitigation by the non-LOCA transient analyses. Listed below is the equipment associated with these occurrences.

1. The nonsafety-related secondary MSIV serves as the backup isolation device to the safety-related primary MSIV for isolation of the MSS piping penetrating the containment. Similarly, the nonsafety-related primary main steam isolation bypass valve (MSIBV) serves as the backup isolation device to the safety-related primary MSIBV. (Section 7.1.4)

2. The nonsafety-related feedwater regulating valve serves as the backup isolation device to the safety-related FWIV for isolation of the feedwater system piping penetrating the containment. (Section 7.1.4)

3. The nonsafety-related feedwater check valve serves as the backup isolation device to the safety-related feedwater check valve for isolation of the DHRS when

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 reverse flow is experienced during a break in the feedwater system piping. (Section 7.1.4) Operator actions credited for the non-LOCA transient analyses are typically justified and consistent with plant operating procedures. For an NPM, there are no occasions where operator action is credited for event mitigation by the non-LOCA transient analyses. Operator actions taken to prevent abnormal operating events from resulting in more severe events are excluded from consideration. For example, very small leaks of reactor coolant from the CVCS that do not result in automatic reactor trip for more than 30 minutes are considered an abnormal operating event where operators are expected to identify and isolate the leak before it results in a more severe event. 7.2 Event Specific Methodology The non-LOCA event simulations are performed using conservative methodologies. Pertinent event-specific methodologies, as well as inputs and results for a representative NPM for non-LOCA event simulations are presented herein, and compared with the regulatory acceptance criteria listed in Table 7-4. Section 4.1 contains additional discussion of Chapter 15 design basis events and acceptance criteria. All acceptance criteria are considered for each event, and the criteria with the potential for being challenged are identified and evaluated in further detail (i.e., overcooling events do not challenge the acceptance criterion for primary side pressure, but may challenge the CHFR acceptance criteria). An event-specific parameter that is relevant to the acceptance criterion may be described as challenging in the event-specific summary, however, it is recognized that the parameter may not present the worst challenge for any event. Table 7-4 Regulatory acceptance criteria Description AOO Criteria IE Criteria Accident Criteria RCS pressure 110% of design 120% of design 120% of design SG pressure 110% of design 120% of design 120% of design CHFR(1) > Limit Note (3) Note (3) Maximum fuel centerline Limit Note (3) Note (3) temperature(1) Containment integrity(2) < Limits < Limits < Limits Escalation of an AOO to No No No an accident (AOO) or consequential loss of system functionality (IE or accident)? Dose(1) Normal operations < Limit < Limit

1. This criterion is confirmed as part of a separate follow-on analysis.

2. Containment integrity is evaluated by a separate analysis methodology.

3. If the minimum CHFR is less than or equal to the CHFR analysis limit, or if the maximum fuel centerline temperature exceeds the melting temperature, the fuel rod is assumed to be failed. If fuel failure is calculated, this is accounted for in the downstream radiological dose analysis.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 For each event analyzed following the non-LOCA evaluation model, a description of the event progression, significant inputs and results, and representative results of sensitivity studies are presented in the following sections. Sensitivity studies are performed to identify plant conditions that result in bounding transient analyses. Studies that identify acceptance criteria challenges or bounding transient forcing functions are discussed as well. Other sensitivity studies that determine bounding inputs for RCS flow, fuel parameters, etc. may not necessarily be discussed for every event. The selection of parameters to be studied is focused on the acceptance criteria challenged by the event. For events where a parameter has minimal or no impact on the consequences, the bias or conservatism identified for the parameter is that typically applied; use of an alternate assumption is also acceptable. The sensitivity study results provided herein are for a representative NPM. Sensitivity studies are repeated as necessary for different NPM designs. Initial RCS flow is typically biased to the low condition in all event simulations because this is bounding for MCHFR. ((

                 }}2(a),(c)

Steam generator tube plugging is considered for each event in the "Initial conditions, biases, and conservatisms" tables. The term, "Biased to the low condition" indicates no tube plugging is assumed. Biased to the high condition indicates (( }}2(a),(c) steam generator tube plugging. In an NPM design, the rod control system is set to insert only mode at full power to prevent automatic withdrawal of the control bank at full power. Although this plant feature exists, the feature is not credited during events where control rod withdrawal results in a bounding result. An NPM utilizes a nonsafety-related turbine bypass system sized to handle full steam flow rate at 100 percent RTP. As such, the turbine bypass valves open following a turbine trip to control the RCS temperature without steam relief to the atmosphere. Since the turbine bypass system enhances heat removal by the secondary system, these actions are not credited for the non-LOCA transient analyses. ((

                                                                                                   }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                                   }}2(a),(c)

Separate analyses for subchannel CHF, fuel centerline temperature, and containment pressure calculations are performed using the appropriate licensed NuScale methodologies. Extended cooldown via the DHRS is considered as part of the system design. 7.2.1 Decrease in Feedwater Temperature The methodology used to simulate a postulated decrease in feedwater temperature for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.1.1 General Event Description and Analysis Methodology The event is initiated by a feedwater system malfunction that causes a decrease in feedwater temperature, resulting in an unplanned overcooling of the RCS. The subsequent decrease in RCS temperature increases core reactivity due to moderator feedback, which raises reactor power. Decreasing average RCS temperature also prompts the control rod controller to withdraw the regulating bank from the core if automatic control is enabled. Rising reactor power typically causes RTS actuation on a high power or high power rate signal, and DHRS actuation and isolation of the secondary system occur post-reactor trip on other MPS signals. Closure of the FWIVs isolates the SGs from the colder feedwater, ending the overcooling event. Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient terminates with the NPM in a safe, stable condition. Table 7-5 lists the relevant acceptance criteria, SAF, and LOP scenarios. The limiting MCHFR typically occurs when the event is initiated from full power conditions. ((

                                                                                       }}2(a),(c) For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. ((
                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 To maximize the overall feedwater temperature change, the feedwater temperature transient starts at the initial (full power) feedwater temperature biased to the high condition, and terminates at the coldest temperature in the secondary, which is saturation temperature at condenser vacuum conditions. A sensitivity study on feedwater temperature cooldown rate is performed to identify the rate that results in limiting conditions. ((

                              }}2(a),(c)

Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation, covered by a separate methodology. Table 7-5 Acceptance criteria, single active failure, loss of power scenarios - decrease in feedwater temperature Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR CHF is the challenged acceptance criterion for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure The challenging cases typically occur when all equipment operates as designed. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event. 7.2.1.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-6. Table 7-6 Acceptance criteria - decrease in feedwater temperature Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the event, sensitivities that maximize primary pressure are not analyzed. Peak primary pressure resulting from a decrease in feedwater temperature is bounded by other AOO events. © Copyright 2022 by NuScale Power, LLC 463

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-6 Acceptance criteria - decrease in feedwater temperature (Continued) Acceptance Criteria Discussion Secondary pressure Pressure in the portion of the secondary system between the FWIVs and MSIVs increases rapidly post-DHRS actuation. However, due to the depressurizing nature of this cooldown event, sensitivities that maximize peak secondary pressure are not analyzed. Peak secondary pressure resulting from a decrease in feedwater temperature is bounded by other AOO events. CHFR Due to the increase in reactor power and subsequent reduction of MCHFR, this acceptance criterion is challenged for the decrease in feedwater temperature event. Consequently, sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.1.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-7 are considered in identifying a bounding transient simulation for MCHFR. Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition.

                                                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( Initial PZR pressure Biased to the high condition Initial PZR level Biased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Nominal. Moderator Temperature Coefficient Biased to EOC conditions. (MTC) Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Minimum feedwater temperature Biased to the low condition. Feedwater temperature cooldown Varied rate

                                                                                                          }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Varied Turbine bypass valves Disabled.

                                                                                                       }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                           }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. For example, sensitivity studies are performed to consider the effects of fuel-related parameters (initial fuel temperature, time in life). Representative results for these studies are presented in Table 7-8 and Table 7-9. ((

                                   }}2(a),(c)

The results of these studies are provided to demonstrate the process by which impact of the the bias of a parameter can be determined. For sensitivity studies that show a response similar to Table 7-8, the limiting bias should be selected. For sensitivity studies that show a response similar to Table 7-9, there is not a bias direction that is limiting and the parameter can be selected based on other analysis considerations. Other events in Section 7.2 use the same general sensitivity study process. © Copyright 2022 by NuScale Power, LLC 467

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-8 Representative fuel exposure study ((

                                                                                                     }}2(a),(c)

Table 7-9 Representative fuel temperature study ((

                                                                                                     }}2(a),(c)

Table 7-10 Not Used © Copyright 2022 by NuScale Power, LLC 468

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-11 Not Used 7.2.2 Increase in Feedwater Flow The methodology used to simulate a postulated increase in feedwater flow for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.2.1 General Event Description A feedwater system malfunction that causes an increase in feedwater flow results in an unplanned overcooling of the RCS. The subsequent decrease in RCS temperature increases core reactivity due to moderator feedback, which raises reactor power. Decreasing average RCS temperature also prompts the control rod controller to withdraw the regulating bank from the core if automatic control is enabled. Rising reactor power can cause RTS actuation on the high power or high pressurizer pressure signal. The feedwater flow increase can also cause RTS actuation on low steam line superheat or high steam line pressure. DHRS also actuates on one of the several signals that cause RTS. Closure of the FWIVs following DHRS actuation isolates the SGs from the feedwater source, ending the overcooling event. Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient calculation is terminated with the NPM in a safe, stable condition. Table 7-12 lists the relevant acceptance criteria, SAF, and LOP scenarios. A rapid (step) increase in feedwater flow is simulated. The limiting MCHFR typically occurs when the event is initiated from full power conditions, ((

                                                         }}2(a),(c) For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events.

((

                                                                         }}2(a),(c). Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s)

© Copyright 2022 by NuScale Power, LLC 469

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. Table 7-12 Acceptance criteria, single active failure, loss of power scenarios - increase in feedwater flow Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR CHF is challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure The challenging cases typically occur when all equipment operates as designed. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event. 7.2.2.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-13. Table 7-13 Acceptance criteria - increase in feedwater flow Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the event, sensitivities that maximize primary pressure are not analyzed. Peak primary pressure resulting from an increase in feedwater flow is bounded by other AOO events. Secondary pressure A rapid initial secondary pressure increase occurs prior to RTS actuation. Pressure in the secondary side continues to increase to the peak following DHRS actuation. This second pressure increase is expected behavior following DHRS actuation and is not a direct consequence of the increase in feedwater flow event itself. CHFR Due to the increase in reactor power and subsequent reduction of MCHFR, this acceptance criterion is challenged for the increase in feedwater flow event. Consequently, sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. © Copyright 2022 by NuScale Power, LLC 470

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-13 Acceptance criteria - increase in feedwater flow (Continued) Acceptance Criteria Discussion Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.2.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-14 are considered in identifying a bounding transient simulation. Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition. Initial PZR level Biased to the high condition. Initial feedwater temperature Varied.

                                                                                                           }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 471

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow (Continued) Parameter Bias / Conservatism Basis (( Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the low condition. Initial SG pressure(1) Varied SG heat transfer Varied. RSV lift setpoint Nominal SG tube plugging Varied. RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited.

                                                                                                          }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 472

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow (Continued) Parameter Bias / Conservatism Basis (( PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled to control to constant steam pressure. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed N/A. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                           }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. © Copyright 2022 by NuScale Power, LLC 473

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The increase in feedwater flow also creates the potential for overfilling the SG. The assumptions in Table 7-12 and Table 7-14 related to identifying the limiting MCHFR do not necessarily result in the maximum SG level. For example, a single failure of a feedwater isolation valve to close is expected to maximize SG level. Additional sensitivity studies are performed on the Table 7-14 parameters, as necessary, to identify the case(s) with a potentially limiting SG level. Table 7-15 Not Used Table 7-16 Not Used 7.2.3 Increase in Steam Flow The methodology used to simulate a postulated increase in steam flow for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.3.1 General Event Description In an NPM design, two main steam lines exit the CNV, and combine to form a common main steam line that connects to the high pressure turbine. The turbine bypass and main steam safety valves are located in the common main steam line, which is downstream of the MSIVs. A spurious opening of the turbine bypass valve or a main steam safety valve would cause an increase in steam flow, which results in an unplanned overcooling of the RCS. The subsequent decrease in RCS temperature increases core reactivity due to moderator feedback, which raises reactor power. Decreasing average RCS temperature also prompts the control rod controller to withdraw the regulating bank from the core if automatic control is enabled (rod withdrawal at 100 percent power is inhibited, however, it is conservatively allowed in the analysis). Rising reactor power typically causes RTS actuation on a high power or high power rate signal. Additionally, a large increase in steam flow would rapidly depressurize the secondary system and could cause RTS and isolation of the secondary system on the low steam pressure or low steam superheat signal. Isolation of the secondary system, if it has not already occurred, and actuation of DHRS occur post-trip on other MPS signals. Closure of the FWIVs and MSIVs isolates the steam generators, which ends the overcooling event. Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient calculation is terminated with the NPM in a safe, stable condition. Table 7-17 lists the relevant acceptance criteria, SAF, and LOP scenarios. © Copyright 2022 by NuScale Power, LLC 474

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The limiting MCHFR typically occurs when the event is initiated from full power conditions, and the increase in steam flow is such that the immediate RTS on low steam pressure is avoided and RTS is actuated sometime later (after the minimum CHFR conditions develop). For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events. The increase in steam flow event starts at the initial (full power) steam flow. Sensitivity studies are performed on the degree of steam flow increase to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. Table 7-17 Acceptance criteria, single active failure, loss of power scenarios - increase in steam flow Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR CHF is challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure Note that a single active failure of a FWIV to close would occur after RTS and DHRS actuation, subsequent to when the MCHFR occurs. Consequently, the MCHFR occurs before the single active failure of an FWIV to close could affect the transient. Otherwise, the challenging cases typically occur when all equipment operates as designed. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event. 7.2.3.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-18. © Copyright 2022 by NuScale Power, LLC 475

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-18 Acceptance criteria - increase in steam flow Acceptance Criteria Discussion Primary pressure Primary pressure initially drops as inventory shrinks due to increased heat removal. As reactor power increases and as the PZR heaters respond, an increase (typically less than 100 psi) in primary pressure is observed. Secondary pressure Peak secondary pressure does not change significantly during the initial phase of the transient due to the relatively small increase in steam flow rate analyzed in the challenging case. It is only after DHRS actuation and closure of the FWIVs and MSIVs that secondary pressure begins to increase rapidly. Steam generator pressure increase resulting from MS isolation is expected and is not a direct consequence of the increase in steam flow event itself. CHFR Due to the increase in reactor power and subsequent reduction of MCHFR, this acceptance criterion is challenged for the increase in steam flow event. Consequently, sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.3.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-19 are considered in identifying a bounding transient simulation for MCHFR. © Copyright 2022 by NuScale Power, LLC 476

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition. Initial PZR level Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Biased to the high condition. SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Steam flow increase Varied

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 477

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow (Continued) Parameter Bias / Conservatism Basis (( RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves N/A. Turbine bypass valves N/A. Feedwater and Turbine Load Control feedwater pump speed Disabled.

                                                                                                         }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 478

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow (Continued) Parameter Bias / Conservatism Basis (( CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

2. ((

                                                                                                         }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. Table 7-20 Not Used Table 7-21 Not Used 7.2.4 Steam System Piping Failure Inside or Outside of Containment The methodology used to simulate a postulated steam system piping failure for an NPM, and an evaluation of the acceptance criteria listed in Table 7-4, are presented below. Since both split breaks (relatively higher event frequency) and double-ended guillotine breaks (relatively lower event frequency) are analyzed, the more restrictive AOO criteria for system pressures, critical heat flux ratio, and fuel centerline melt applicable to breaks with higher event frequency are used in the evaluation. Radiological dose consequences are assessed as part of the downstream accident radiological dose analysis, documented in a separate report, and compared against the appropriate acceptance criteria. © Copyright 2022 by NuScale Power, LLC 479

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.4.1 General Event Description The steam line break event ranges from small breaks to double ended ruptures of a main steam line causing an increase in steam flow and an over cooling of the RCS. This event can occur inside or outside the containment vessel (CNV). A break inside CNV causes a rapid pressurization of the CNV resulting in a reactor trip and CNV isolation with a DHRS actuation. This break location is non-limiting for pressure and MCHFR but potentially challenging to the DHRS as one loop is disabled with the break inside the CNV. A steam line break outside of the CNV causes an increase in steam flow event that causes either a low SG pressure trip or a high core power trip due to the reactor power response from the decreased RCS temperature. The break flow is stopped by the MSIVs closing and depressurization of the steam system piping. Smaller breaks cause a slower loss of secondary pressure due to the increased steam demand that could cause a high core power trip. These smaller breaks can result in a significant delay in detection time, making the small break cases potentially challenging for MCHFR. Reactor trip and transition to stable DHRS flow eventually terminate the transients and bring the NPM to a safe, stable condition. For this overcooling event, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events. ((

                                                                               }}2(a),(c)

The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-22. © Copyright 2022 by NuScale Power, LLC 480

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-22 Acceptance criteria, single active failure, loss of power scenarios - steam line break Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR Critical heat flux is potentially challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) Radiological consequences A postulated break in the main steam line is evaluated for radiological consequences. Failure of one MSIV to close on the train with break MSIV single failure typically has no effect on MCHFR since DHRS actuation is not before the time when MCHFR occurs. MSIV single failure is potentially bounding for mass releases for break locations outside the CNV if mass release is calculated for use in the downstream radiological analysis. Failure of one FWIV to close on the train with the break FWIV single failure typically has no effect on MCHFR since DHRS actuation is not before the time when MCHFR occurs. FWIV single failure is potentially bounding for mass releases for break locations inside the CNV if mass release is calculated for use in the downstream radiological analysis. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event. 7.2.4.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-23. Table 7-23 Acceptance criteria - steam line break Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of this cooldown event, primary pressure remains below the acceptance criterion for peak primary pressure. Secondary pressure Due to the depressurizing nature of this cooldown event, secondary pressure remains below the acceptance criterion for peak secondary pressure. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Sensitivity cases are performed to support the follow-on MCHFR evaluation. © Copyright 2022 by NuScale Power, LLC 481

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-23 Acceptance criteria - steam line break (Continued) Acceptance Criteria Discussion Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation to a more serious accident or consequential This criterion is satisfied by demonstrating stable loss of functionality RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.4.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-24 are considered in identifying a bounding transient simulation for MCHFR and mass release. In the radiological analysis, primary-to-secondary leakage is assumed, which allows primary coolant to be released with the SG break flow. The transient SG mass release can be calculated for use as an input to the downstream radiological analysis. Alternatively, bounding assumptions for primary coolant release can be used in the radiological analysis to eliminate the need for calculating SG mass release. If such bounding assumptions are used, transient analysis to maximize SG mass release is not required. Table 7-24 Initial conditions, biases, and conservatisms - steam line break Parameter Bias / Conservatism Basis(1) (( Initial reactor power Biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition. Initial PZR level Biased to the high condition. Initial feedwater temperature Varied.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 482

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( Initial fuel temperature Biased low for EOC conditions and high for BOC conditions. MTC Both EOC and BOC conditions. Kinetics Both EOC and BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(2) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. RCS Temperature Control Automatic rod control Enabled. (MCHFR) Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Enabled.

                                                                                                          }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 483

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                       }}2(a),(c)

((

                                                                                                       }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 484

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response and challenging mass releases for this overcooling event. Table 7-25 Not Used 7.2.5 Containment Flooding / Loss of Containment Vacuum The methodology used to simulate a postulated containment flooding / loss of containment vacuum for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. This AOO is unique to the NPM designs. 7.2.5.1 General Event Description and Methodology The loss of containment vacuum terminology refers specifically to vapor / air or minimal water leakage into the containment vessel (CNV) that does not result in water build-up inside the CNV. The containment evacuation (CE) system is used to maintain a vacuum in the CNV during normal operation. If the leakage rate exceeds the CE systems capacity, a high containment pressure (> 9.5 psia in the example Table 7-3) signal can occur, which trips the reactor and isolate all containment penetrations. The containment flooding terminology refers to liquid build-up in the CNV that can potentially cause loss of CNV vacuum and result in containment isolation and reactor trip. The potential CNV flooding sources considered are piping failures inside the CNV. Depending on the CNV operating pressure, some of these lines inside the CNV carry fluid at a lower temperature than the CNV saturation temperature. If one of these low temperature lines fails, flashing inside the CNV does not occur instantly, so the high containment pressure analytical limit is not immediately reached. Therefore these low temperature line failures can potentially create different challenges than failures in other lines carrying higher temperature fluid such as main steam and feedwater lines. The CNV flooding sources due to a piping failure inside containment include: feedwater (FW), main steam (MS), chemical and volume control system (CVCS), high point vent (HPV), and reactor component cooling water (RCCW). Only a RCCW piping failure inside the CNV is considered as a flooding source for the CNV flooding / loss of CNV vacuum event. Other break locations are evaluated as separate initiating events. © Copyright 2022 by NuScale Power, LLC 485

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 During this event, heat removal from the RCS increases due to loss of CNV vacuum or CNV flooding from RCCW. As a result, the CNV flooding/loss of CNV vacuum event does not introduce more challenging conditions for RCS pressure and secondary side pressure compared to other AOOs such as the overheating events. Due to the overcooling effect of the event, an increase in reactor power and subsequent reduction of MCHFR may occur. A series of CNV flooding sensitivities are conducted to assess the effects on MCHFR from RCCW flow, RCCW temperature, containment pressure, primary pressure, and pool temperature. ((

                                                                                          }}2(a),(c) The purpose is to show that there is little observable difference between the two cases with the change in reactor power and the additional heat loss is minimal. Any loss of CNV vacuum cases that cause high CNV pressure trip should be bounded by the results obtained from the CNV flooding cases that result in reactor trip.

The relevant acceptance criteria, single active failure, and loss of power scenarios are listed in Table 7-26. Table 7-26 Acceptance criteria, single active failure, loss of power scenarios - containment flooding / loss of containment vacuum Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR MCHFR is potentially challenged during this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure The challenging cases typically occur when all equipment operates as designed. No loss of power A loss of AC power at event initiation is typically non-limiting because reactor trip and containment isolation mitigate the event, while a loss of AC power coincident with turbine trip typically does not alter the MCHFR because the time of reactor trip is not changed. 7.2.5.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-27. © Copyright 2022 by NuScale Power, LLC 486

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-27 Acceptance criteria - containment flooding / loss of containment vacuum Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the event, sensitivities that maximize primary pressure are not analyzed. Peak primary pressure resulting from CNV flooding/loss of CNV vacuum is bounded by other AOO events. Secondary pressure Due to the primary system depressurization of this cooldown event, secondary pressure remains below the acceptance criterion for peak secondary pressure. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.5.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-28 are considered in order to identify the bounding transient simulation for MCHFR for the CNV flooding/loss of CNV vacuum event. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied

                                                                                                           }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 487

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( Initial PZR level Nominal Initial feedwater temperature Nominal Initial fuel temperature Nominal MTC Biased to the EOC condition. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Initial containment pressure Varied Initial pool temperature Varied RCCW leak flow Varied RCCW temperature Varied RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 488

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( PZR Pressure Control PZR spray Disabled. PZR heaters Enabled. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled. (CNV flooding) N/A. (CNV vacuum loss)

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 489

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. As an example, the initial reactor power, pressurizer pressure, RCS average coolant temperature, RCS flow rate, containment pressure, pool temperature, RCCW leak flow rate, RCCW temperature were varied for a representative NPM for the CNV flooding/loss of CNV vacuum event. The sensitivity studies indicated that the CNV flooding cases are more challenging to MCHFR than loss of CNV vacuum. The sensitivity study results also indicated that for a variety of initial RCS conditions, reactor pool conditions, and condition of liquid or air ingress to containment, loss of containment vacuum or containment flooding results in a slow overcooling transient that is non-limiting with respect to MCHFR compared to other AOOs. Table 7-29 Not Used 7.2.6 Turbine Trip / Loss of External Load The methodology used to simulate a postulated turbine trip/loss of external load (LOEL) for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.6.1 General Event Description Turbine trip initiates with a turbine stop valve closure while loss of external load initiates with a turbine control valve closure. Otherwise, these transients are essentially equivalent and result in the sudden removal of the secondary side heat sink, overpressurization of the secondary system, and overheating of the RCS. Rising system pressures typically result in RTS actuation on the high PZR or steam pressure signal. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. Table 7-30 lists the relevant acceptance criteria, SAF, and LOP scenarios. The limiting pressure responses typically occur when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. Sensitivity studies on initial primary temperature and primary/secondary pressures are performed to identify the conditions that maximize peak primary and secondary pressures. Additional sensitivity studies © Copyright 2022 by NuScale Power, LLC 490

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 are performed on other parameters, as necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures. Table 7-30 Acceptance criteria, single active failure, loss of power scenarios - turbine trip

                                         / loss of external load Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure                    Primary and secondary pressures are challenged during this overheating event.

Failure of one FWIV to close Typically challenging for secondary side pressurization cases (negligible effect for primary side pressurization cases, which assume loss of AC power at event initiation and therefore feedwater is lost). Loss of AC power at transient initiation Typically maximizes primary pressure (feedwater is lost). No loss of power Typically maximizes secondary pressure. 7.2.6.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-31. Table 7-31 Acceptance criteria - turbine trip / loss of external load Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation and continued FW operation, in addition to the actual turbine trip or loss of external load. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. © Copyright 2022 by NuScale Power, LLC 491

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-31 Acceptance criteria - turbine trip / loss of external load (Continued) Acceptance Criteria Discussion Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.6.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-32 are considered in identifying the bounding transient simulation for primary and steam generator pressure. Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased to the high condition to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the high condition MTC Consistent with BOC conditions.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 492

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load (Continued) Parameter Bias / Conservatism Basis (( Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. Steam generator heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Disabled. Turbine bypass valves Disabled.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 493

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load (Continued) Parameter Bias / Conservatism Basis (( Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. The maximum RCS pressure is limited by the RSV lifting and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. Table 7-33 Not Used 7.2.7 Loss of Condenser Vacuum The methodology used to simulate a postulated loss of condenser vacuum for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4 are presented below. 7.2.7.1 General Event Description Loss of condenser vacuum initiates with a turbine stop valve closure. Also, a loss of condenser vacuum is postulated to lead to a loss of feedwater flow. Turbine trip and loss of feedwater result in the sudden removal of the secondary side heat sink, pressurization of the secondary system, and overheating of the RCS. Rising system pressures typically result in a rapid RTS actuation on either high PZR or steam pressure. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. © Copyright 2022 by NuScale Power, LLC 494

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The loss of condenser vacuum event is essentially equivalent to the turbine trip / loss of external load events discussed in Section 7.2.6. The main difference is that the loss of condenser vacuum event includes a loss of feedwater flow at event initiation. However, because the turbine trip / loss of external load events consider a loss of normal AC power at event initiation, those events also model a loss of feedwater flow at event initiation. As a result, the scenarios analyzed as part of Section 7.2.6 address the loss of condenser vacuum event. Therefore, the relevant acceptance criteria, SAF, and LOP scenarios from Table 7-30 are also applicable to the loss of condenser vacuum event. Table 7-34 Not Used 7.2.7.2 Acceptance Criteria The evaluation of the most challenging case(s) relative to the acceptance criteria for the turbine trip / loss of external load events presented in Table 7-31 is applicable to the loss of condenser vacuum event. Table 7-35 Not Used 7.2.7.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-32 for the turbine trip / loss of external load events are applicable to the loss of condenser vacuum event. Table 7-36 Not Used Table 7-37 Not Used © Copyright 2022 by NuScale Power, LLC 495

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.8 Main Steam Isolation Valve(s) Closure The methodology used to simulate a postulated main steam isolation valve closure for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4 are presented below. 7.2.8.1 General Event Description A closure of one or both main steam isolation valves results in a pressurization of the secondary system and overheating of the RCS. Rising pressures typically result in a rapid RTS actuation on the high pressurizer pressure or high steam pressure signals. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. The relevant acceptance criteria, single active failure, and loss of power scenarios are listed in Table 7-38. The MSIV closure event can occur when one or both MSIVs close unexpectedly. The limiting pressure responses typically occur when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. Sensitivity studies on number of MSIVs closing, initial primary temperature and primary/secondary pressures are performed to identify the conditions that maximize peak primary and secondary pressures. Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures. Table 7-38 Acceptance criteria, single active failure, loss of power scenarios - main steam isolation valve closure Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure Primary and secondary pressures are challenged during this overheating event. No single failure Typically challenging for primary pressure. Failure of one FWIV to close Typically challenging for steam generator pressure. Loss of AC Power at transient initiation Typically maximizes primary pressure. No loss of power Typically maximizes secondary pressure. 7.2.8.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-39. © Copyright 2022 by NuScale Power, LLC 496

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-39 Acceptance criteria - main steam isolation valve closure Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation and continued FW operation, in addition to the actual main steam isolation valve closure. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.8.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-40 are considered in identifying the bounding transient simulation for primary and steam generator pressure. Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Varied.

                                                                                                           }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 497

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure (Continued) Parameter Bias / Conservatism Basis (( Initial feedwater temperature Nominal. Initial fuel temperature Biased to the high condition MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. Steam generator heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 498

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure (Continued) Parameter Bias / Conservatism Basis (( Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. Table 7-41 Not Used 7.2.9 Loss of Nonemergency AC Power The methodology used to simulate a postulated loss of nonemergency (normal) AC power for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. © Copyright 2022 by NuScale Power, LLC 499

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.9.1 General Event Description The low voltage AC electrical distribution system (ELVS) supplies AC power to plant motors, heaters, packaged equipment, and battery chargers. Loss of normal AC power to the station auxiliaries can result from electrical grid-related failures, failures in plant or switchyard equipment, or external weather events. The nonsafety-related EDNS and EDAS/EDSS may remain available via battery operation; the primary loads for these systems include the module control system (MCS) and the MPS. Loss of the EDNS and/or EDAS/EDSS batteries with the loss of normal AC power is considered as described in Section 7.1.3. A loss of AC power results in a pressurization of the secondary system and overheating of the RCS. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. Table 7-42 lists the relevant acceptance criteria, SAF, and LOP scenarios. The electrical system design may vary by NPM design. Review of the electrical system is performed to determine the impact on plant equipment from the loss of power. In general, the following typically occur from the loss of power (AC or DC or both): turbine trip occurs the feedwater pumps and CVCS pumps stop the PZR heaters turn off control rods begin to drop due to loss of power to control rod drive mechanisms (CRDMs) the MPS actuates reactor trip, DHRS, and various system isolations within 60 seconds after event initiation (if not already actuated during that time), due to loss of AC power to the EDAS/EDSS battery chargers the MPS actuates ECCS within 24 hours after event initiation, due to loss of AC power to the EDAS/EDSS battery chargers Of the various scenarios, the limiting one is typically the scenario that does not result in an immediate reactor trip or full control rod insertion at event initiation. Typically this scenario is the loss of the ELVS at event initiation, with EDNS and EDAS/EDSS available. Cases considering the loss of EDSS battery backup coincident with the initiating event are considered as described in Section 7.1.3. Consequently, the limiting pressure responses typically occur when the event is initiated from full power conditions, reactor trip is delayed until MPS actuation, and the initial conditions are biased in the conservative directions. Sensitivity studies on initial primary temperature and primary/secondary pressures are performed to identify the conditions that maximize peak primary and secondary pressures. Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures. © Copyright 2022 by NuScale Power, LLC 500

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-42 Acceptance criteria, single active failure, loss of power scenarios - loss of normal AC power Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure Primary and secondary pressures are challenged during this overheating event. No single failure The challenging cases for primary pressure typically occur when all equipment is operational. Since feedwater is lost at transient initiation, peak secondary pressures are insensitive to the single failure of an FWIV to isolate. Loss of AC power at transient initiation Initiating event. 7.2.9.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-43. Table 7-43 Acceptance criteria - loss of normal AC power Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation, in addition to the actual loss of normal AC power. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. © Copyright 2022 by NuScale Power, LLC 501

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.9.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-44 are considered in identifying the bounding transient simulation for primary and secondary pressure. Table 7-44 Initial conditions, biases, and conservatisms - loss of normal AC power Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the high condition MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 502

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-44 Initial conditions, biases, and conservatisms - loss of normal AC power (Continued) Parameter Bias / Conservatism Basis (( PZR Pressure Control PZR spray Enabled. PZR heaters Disabled. PZR Level Control Charging Not credited. Letdown Enabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

2. Loss of normal AC power initiating event results in a loss of system function by loss of power to the system thereby making the system control not relevant to the event.

© Copyright 2022 by NuScale Power, LLC 503

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. Table 7-45 Not Used 7.2.10 Loss of Normal Feedwater Flow The methodology used to simulate a postulated loss of normal feedwater flow for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.10.1 General Event Description A postulated fault results in a partial or complete loss of feedwater flow, and the water in the steam generators boils off. The loss of steam generators as a heat sink leads to a rise in the RCS temperature and pressure until the reactor typically trips due to high PZR pressure. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-46. Table 7-46 Acceptance criteria, single active failure, loss of power scenarios - loss of normal feedwater flow Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure System pressures are challenged during this overheating event. No single failure The challenging cases typically occur when all equipment is operational. Loss of AC power at turbine trip Typically maximizes system pressures. 7.2.10.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-47. © Copyright 2022 by NuScale Power, LLC 504

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-47 Acceptance criteria - loss of normal feedwater flow Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation, in addition to the actual loss of feedwater. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.10.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-48 are considered in identifying a bounding transient simulation for primary and steam generator pressure. Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Varied. Initial fuel temperature Biased to the high condition. MTC Consistent with BOC kinetics.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 505

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow (Continued) Parameter Bias / Conservatism Basis (( Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Feedwater flow decrease Varied RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 506

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow (Continued) Parameter Bias / Conservatism Basis (( Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed N/A. CNV Pressure Control CNV evacuation system Disabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest primary and secondary pressures, varying the magnitude of the feedwater flow rate decrease (other parameters are biased as indicated in Table 7-48). Table 7-49 Not Used 7.2.11 Inadvertent Decay Heat Removal System Actuation The methodology used to simulate a postulated inadvertent DHRS actuation for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. This event is unique to the NPM designs. 7.2.11.1 General Event Description Inadvertent actuation of the DHRS may result from either an unexpected DHRS valve actuation or a spurious DHRS actuation signal. Due to the DHRS design © Copyright 2022 by NuScale Power, LLC 507

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 and configuration, several scenarios exist for consideration. The specific scenarios considered are discussed below. The sequence of events and the MPS signals that are reached first depend on the scenario. Reactor trip and transition to stable DHRS flow terminates the transients with the NPM in a safe, stable condition. The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-50. Scenario 1: The unexpected opening of a single DHRS valve can occur at full power or reduced power conditions. At low power conditions, a portion of the DHRS liquid inventory drains into the feedwater line, which momentarily increases feedwater flow and causes overcooling. This overcooling event is not considered further because it is bounded by other, more limiting overcooling events (i.e., increase in feedwater flow). The most challenging conditions for this heatup scenario occur at full power with initial conditions biased in the conservative directions. Since feedwater flow tends to increase in response to the reduced steam enthalpy and turbine load, limiting the feedwater response maximizes the heatup. Scenario 2: An inadvertent actuation signal isolates one steam generator, and initiates one DHRS train. This scenario is typically bounded by Scenario 3 discussed below. Scenario 3: An inadvertent actuation signal isolates both steam generators, and initiates both DHRS trains. This scenario represents a complete loss of normal heat removal from the RCS. The limiting conditions for this heatup scenario occur at full power with initial conditions biased in the conservative directions. Scenario 4: An inadvertent isolation of one steam generator. The valves associated with one SG are closed, but the associated DHRS train is not actuated. The heatup caused by the isolation of one SG causes an increase in primary pressure that results in reactor trip and RSV opening. The actuation of DHRS occurs later in the transient after MPS signals are reached. Secondary pressure peaks after DHRS actuation. Scenario 5: An inadvertent isolation of both steam generators. The valves associated with both SGs are closed, but neither DHRS train is actuated. The response is similar to Scenario 4, except the increase in primary pressure is more rapid and results in earlier reactor trip and RSV opening. The actuation of DHRS occurs later in the transient after MPS signals are reached. Secondary pressure peaks after DHRS actuation. Each scenario is considered to determine the limiting cases for the acceptance criteria. Sensitivity studies on initial primary and secondary conditions are performed as needed to identify the conditions that maximize peak system pressures. © Copyright 2022 by NuScale Power, LLC 508

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-50 Acceptance criteria, single active failure, loss of power scenarios - inadvertent decay heat removal system actuation Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure Challenged during the heatup scenarios of this event. No single failure The challenging primary pressurization cases typically occur when all equipment is operational. Failure of one FWIV to close Potentially challenging for secondary pressure, but typically bounded by other heatup events. No loss of power Maximizes system pressures. 7.2.11.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-51. Table 7-51 Acceptance criteria - inadvertent decay heat removal system actuation Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation. The DHRS heat removal is limited by the DHR condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. © Copyright 2022 by NuScale Power, LLC 509

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.11.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-52 are considered in identifying the bounding transient simulation for primary and steam generator pressure. Table 7-52 Initial conditions, biases, and conservatisms - inadvertent decay heat removal system actuation Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the high condition MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. Steam generator heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 510

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-52 Initial conditions, biases, and conservatisms - inadvertent decay heat removal system actuation (Continued) Parameter Bias / Conservatism Basis (( RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 511

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this event. Table 7-53 Not Used 7.2.12 Feedwater System Pipe Break Inside or Outside Containment The methodology used to simulate a postulated feedwater system pipe break for an NPM, and an evaluation of the acceptance criteria listed in Table 7-4, are presented below. Since both split breaks (relatively higher event frequency) and double-ended guillotine breaks (relatively lower event frequency) are analyzed, the more restrictive AOO criteria for system pressures, critical heat flux ratio, and fuel centerline melt applicable to breaks with higher event frequency are used in the evaluation. 7.2.12.1 General Event Description A feedwater line break can occur inside or outside of containment, and can range in size from a small split crack to a double ended rupture. A large feedwater line break inside containment results in a loss of containment vacuum and a high containment pressure MPS signal that actuates a reactor trip, isolates the secondary system and CVCS, and opens the DHRS valves. The steam generator, DHRS piping, and DHRS condenser on the affected side drain through the break. The non-affected steam generator system and DHRS loop provide cooling to the RCS via heat transfer to the reactor pool. The response of smaller feedwater line breaks inside containment is similar except that other MPS setpoints, such as high PZR pressure, may be reached before high containment pressure. A feedwater line break outside containment causes a loss of feedwater flow to the steam generators and a heatup of the RCS. Larger breaks result in rapid heatup events that pressurize the RCS beyond the high PZR pressure analytical limit. Smaller breaks cause a more gradual heatup, loss of secondary pressure, and reactor trip and DHRS actuation may occur on other MPS signals. The DHRS provides cooling to the RCS via heat transfer to the reactor pool. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. ((

                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-54 lists the relevant acceptance criteria, SAF, and LOP scenarios. Sensitivity studies on primary and secondary conditions, and break size/location are performed to identify the conditions that maximize peak primary and secondary pressures. Table 7-54 Acceptance criteria, single active failure, loss of power scenarios - feedwater line break Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure System pressures are challenged during this overheating event. Failure of one FWIV to close on the train with the break The challenging cases typically occur when FWIV single failure is assumed. Loss of AC power at transient initiation Typically maximizes system pressures. 7.2.12.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-55. Table 7-55 Acceptance criteria - feedwater line break Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation, in addition to the actual FWLB. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation to a more serious accident or consequential This criterion is satisfied by demonstrating stable loss of functionality RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. © Copyright 2022 by NuScale Power, LLC 513

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.12.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-56 are considered in identifying the bounding transient simulation for primary and steam generator pressure. Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Varied. Initial fuel temperature Biased to the high condition. MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition.

                                                                                                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break (Continued) Parameter Bias / Conservatism Basis (( SG tube plugging Nominal. Break size / location Varied. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled.

                                                                                                         }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break (Continued) Parameter Bias / Conservatism Basis (( CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. Table 7-57 Not Used 7.2.13 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions The methodology used to simulate a postulated uncontrolled CRA bank withdrawal from subcritical or low power startup conditions for an NPM, and an evaluation of the acceptance criteria listed for an AOO in Table 7-4, are presented below. The range of initial power levels associated with low power startup conditions for an NPM is based on the low setting for the high power signal (below 15 percent RTP for the example in Table 7-3). When core power reaches the low setting level, a hold point is established to alter the high power setting. Thus, low power startup conditions exist until reactor power reaches the low setting level. 7.2.13.1 General Event Description and Methodology The limiting event consequences to an uncontrolled CRA bank withdrawal from subcritical or low power startup conditions typically (for most PWR designs) occur for cases with very low initial power levels (~1 Watt). The primary reason for this behavior is the flux rate signals associated with the source range and intermediate range are typically either not safety related or not of sufficient quantity to adequately address single failures. An NPM, however, incorporates a safety related signal for each of these channels from each core quadrant into the MPS. © Copyright 2022 by NuScale Power, LLC 516

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Consequently, the limiting event consequences for an NPM typically occur for cases with higher initial power levels. A spectrum of constant reactivity insertion rates is evaluated. These reactivity insertion rates encompass the credible range resulting from a single control bank withdrawal. If necessary, this range is supplemented to include the reactivity insertion rates associated with an inadvertent decrease in boron concentration event (Section 7.2.16). Two event scenarios with different protection schemes are evaluated to determine which scenario produces the limiting event consequences. The first scenario arises when the high count rate signal is available because the intermediate range channel does not have an established signal. In this instance, the high power rate signal is not active (below 15 percent RTP in the Table 7-3 example), so core protection is provided by the high count rate signal and the startup rate (source range) signal. ((

                                                                          }}2(a),(c)

The second scenario arises when the high count rate signal is not available because the intermediate range channel has an established signal. In this instance, the high power rate signal is not active (below 15 percent RTP in the Table 7-3 example), thus core protection is provided by the high power signal and the startup rate (intermediate range) signal. The event scenario with the highest core power typically corresponds to the initial power level and reactivity insertion rate that cause the high power signal (low setting) and the startup rate (intermediate range) signal to occur at nearly the same time. This scenario is typically limiting because it represents the maximum rate of power change at the maximum core power. If the initial power is increased, the reactor trips on the high power signal but at a slower rate of power increase. Similarly, if the reactivity insertion rate is increased, the reactor trips on the startup rate but at a lower core power. Before initiating an approach to critical, the reactor coolant is heated to the minimum temperature for criticality. The heating of the reactor coolant is performed by the Module Heatup System (MHS) via the CVCS. At least one feedwater pump is operating at these RCS temperatures. Since feedwater flow is provided to both SGs, it may continue to provide decay heat removal following an uncontrolled CRA bank withdrawal from subcritical or low power startup conditions. If normal feedwater flow is not available, then depending on the point at which the event occurs in the startup process, the flooded containment or the DHRS provides decay heat removal. The peak power © Copyright 2022 by NuScale Power, LLC 517

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 and duration of the power spike are not sufficient to cause a significant temperature or pressure increase. Hence, the maximum power and minimum CHFR occur shortly after reactor trip while the RCS pressure and MS pressure do not challenge the relevant acceptance criteria. Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR or fuel centerline temperature. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. Table 7-58 lists the relevant acceptance criteria, SAF, and LOP scenarios. Table 7-58 Acceptance criteria, single active failure, loss of power scenarios - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR and maximum fuel centerline temperature. MCHFR and maximum fuel centerline temperature are challenged during this reactivity anomaly event. No single failure. The challenging cases typically occur when all equipment is operational. No loss of power. The challenging cases typically occur when AC power is available for the event duration. 7.2.13.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-59. Table 7-59 Acceptance criteria - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Acceptance Criteria Discussion Primary pressure This criterion is not an acceptance criterion listed in Section 15.4.1 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for primary pressure compared to other AOOs. Secondary pressure This criterion is not an acceptance criterion listed in Section 15.4.1 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for secondary pressure compared to other AOOs. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. © Copyright 2022 by NuScale Power, LLC 518

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-59 Acceptance criteria - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions (Continued) Acceptance Criteria Discussion Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS (if actuated) pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.13.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-60 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature. Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Nominal. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Nominal. Initial PZR level Nominal. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to BOC conditions.

                                                                                                           }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions (Continued) Parameter Bias / Conservatism Basis (( Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Reactivity insertion rate Varied. RCS Temperature Control Automatic rod control N/A. Boron concentration Not credited. PZR Pressure Control PZR spray Enabled. PZR heaters Enabled. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves N/A. Turbine bypass valves Enabled. Feedwater and Turbine Load Control feedwater pump speed N/A.

                                                                                                          }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions (Continued) Parameter Bias / Conservatism Basis (( CNV Pressure Control CNV evacuation system Enabled.

                                                                                                            }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR and highest fuel centerline temperature for this reactivity event. Sensitivity study results typically demonstrate the lack of challenging MCHFR values predicted for this event by NRELAP5, which is further supported by the values predicted with the approved subchannel methodology. ((

                                    }}2(a),(c)

Table 7-61 Not Used 7.2.14 Uncontrolled Control Rod Assembly Bank Withdrawal at Power The methodology used to simulate a postulated uncontrolled control rod assembly (CRA) bank withdrawal at power for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.14.1 General Event Description and Methodology As stated in Section 7.2.13, low power startup conditions exist for an NPM until reactor power reaches the low setting level (15 percent RTP for the example in Table 7-3). Accordingly, the uncontrolled CRA bank withdrawal at power event extends from the low setting level to HFP. The withdrawal of the control bank causes a reactivity insertion that increases reactor power and leads to a rise in coolant temperature, pressurizer level, and © Copyright 2022 by NuScale Power, LLC 521

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 RCS pressure. Reactivity feedback from the rising fuel temperature partially counteracts the reactivity insertion, slowing the power increase, which continues until the system trips on high power, high power rate, high pressurizer pressure, or high RCS temperatures. The maximum power and minimum MCHFR occur just after the resulting scram, while the peak primary pressure occurs some time later, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established by the end of the transient. Following reactor trip and subsequent turbine trip, the turbine bypass to the condenser opens to control the RCS temperature. However, the actions of the turbine bypass system are not credited, so as to minimize heat removal by the secondary side. Although turbine load is an input to the feedwater controller, no changes are made to this controller because the RCS responses are not sufficient to affect feedwater control. The limiting MCHFR typically occurs for a reactivity insertion rate that results in reactor trip on core power, pressurizer pressure, and RCS riser temperature signals at approximately the same time. These conditions typically arise for events initiated with lower reactivity insertion rates because higher reactivity insertion rates cause the MPS to trip much earlier on high power rate. The earlier reactor trip reduces the energy added to the reactor coolant, thereby producing a higher MCHFR. The range of reactivity insertion rates considered is sufficient to identify the point of transition to the high power rate signal. Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. The maximum fuel centerline temperature typically occurs when core power exceeds its analytical limit. This condition typically arises for events initiated from full power with the highest reactivity insertion rate as determined from the resulting bank worth and control rod step speed. A spectrum of constant reactivity insertion rates is evaluated. These reactivity insertion rates encompass the credible range resulting from a single control bank withdrawal. If necessary, this range is supplemented to cover the reactivity insertion rates associated with an inadvertent decrease in boron concentration event (Section 7.2.16). Table 7-62 lists the relevant acceptance criteria, SAF, and LOP scenarios. Table 7-62 Acceptance criteria, single active failure, loss of power scenarios - uncontrolled control rod bank withdrawal at power Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR and maximum fuel centerline temperature MCHFR and maximum fuel centerline temperature are challenged during this reactivity anomaly event. © Copyright 2022 by NuScale Power, LLC 522

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-62 Acceptance criteria, single active failure, loss of power scenarios - uncontrolled control rod bank withdrawal at power (Continued) Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest No single failure The challenging cases typically occur when all equipment is operational. No loss of power The challenging cases typically occur when AC power is available for the event duration. 7.2.14.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-63. Table 7-63 Acceptance criteria - uncontrolled control rod bank withdrawal at power Acceptance Criteria Discussion Primary pressure This criterion is not an acceptance criterion listed in Section 15.4.2 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for primary pressure compared to other AOOs. Secondary pressure This criterion is not an acceptance criterion listed in Section 15.4.2 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for secondary pressure compared to other AOOs. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.14.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-64 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature. © Copyright 2022 by NuScale Power, LLC 523

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power Parameter Bias / Conservatism Basis (( Initial reactor power Varied. RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Biased to the high condition. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the low condition. MTC Biased to BOC conditions. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition.

                                                                                                            }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power (Continued) Parameter Bias / Conservatism Basis (( Reactivity insertion rate Varied Maximum RCS Temperature Control Automatic rod control N/A. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                           }}2(a),(c)

1. (( }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. Table 7-65 Not Used 7.2.15 Control Rod Misoperation The methodology used to simulate a postulated control rod misoperation for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.15.1 General Event Description and Methodology The rod control system is used to move (insert or withdraw) the control rod assemblies (CRAs) in response to an operator action or an automatic control. Since these transients are initiated by a malfunction in the rod control system, a variety of reactivity related conditions can result. Specific reactivity conditions for an NPM include: 1) withdrawing a single CRA; 2) dropping one or more CRAs; or,

3) leaving one or more CRAs behind when inserting or withdrawing a control bank. The consequences for each of these reactivity conditions are discussed below.

Table 7-66 lists the relevant acceptance criteria, single active failure, and loss of power scenarios. Withdrawal of a Single CRA The withdrawal of a single CRA causes a reactivity insertion that increases reactor power and leads to a rise in coolant temperature, pressurizer level, and RCS pressure. Feedback from the rising fuel temperature is not sufficient to counteract the reactivity insertion, so the power increases until the system trips on high power, high power rate, high pressurizer pressure, or high RCS temperatures. The maximum power and minimum MCHFR occur just after the resulting scram, while the peak primary pressure occurs some time later, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. The limiting MCHFR typically occurs for a reactivity insertion rate that results in reactor trip on core power, pressurizer pressure, and RCS temperature signals at approximately the same time. These conditions arise for events initiated from partial power with lower reactivity insertion rates because higher reactivity © Copyright 2022 by NuScale Power, LLC 526

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 insertion rates cause the MPS to trip much earlier on high power rate. The earlier reactor trip reduces the energy added to the reactor coolant, thereby producing a higher MCHFR. The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. Consequently, the range of reactivity insertion rates considered is sufficient to identify the point of transition to the high power rate signal (using the lowest reading ex-core detector based on the minimum after to before event initiation ratio of the radial peaking factors for the outer row of fuel assemblies). Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. In contrast to the uncontrolled CRA bank withdrawal at power event (Section 7.2.14), the highest reactivity insertion rate as determined from the resulting rod worth and control rod step speed is significantly lower for the withdrawal of a single CRA event. Dropping One or More CRAs Based on the minimum worth at any time during the cycle for a given core power, i.e., with the control bank positioned at the PDIL, dropping a single CRA causes a reactivity insertion that decreases reactor power. Feedback from the decreasing fuel temperature and the actions of the rod control system to restore power are generally not sufficient to counteract the reactivity insertion, so the power decreases until the system trips on high power rate. For event scenarios without a return to power, the maximum core power, peak primary pressure, and MCHFR occur at event initiation. The peak secondary system pressure occurs some time after the scram, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. The potential for a return to power exists only for events initiated from less than RTP because the reduced worth of the dropped rod gives the rod control system time to act. The corresponding MCHFR for a dropped rod event with a return to power is typically greater than the MCHFR for events initiated from HFP. Hence, the limiting MCHFR cases typically occur at HFP conditions. Following reactor trip and subsequent turbine trip, the turbine bypass to the condenser opens to control the RCS temperature. However, the actions of the turbine bypass system are not credited, so as to minimize heat removal by the secondary side. Although turbine load is an input to the feedwater controller, no changes are made to this controller because the RCS responses are not sufficient to affect feedwater control. The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. The power input to the high power rate signal uses the highest reading ex-core detector, multiplying the core average power by the maximum after drop to before drop ratio of the radial © Copyright 2022 by NuScale Power, LLC 527

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 peaking factors for the outer row of fuel assemblies. Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. The maximum fuel centerline temperature typically occurs at event initiation for those event scenarios with an immediate reactor trip. If the event scenario has a return to power, the maximum fuel centerline temperature is typically bounded by the fuel centerline temperature at HFP because the associated power peak is less than full power. As an alternative to performing a system transient analysis, the MCHFR and linear heat generation rate of the dropped rod event can be confirmed to be bounded by other events. Most rod drops result in a reactor trip on high power rate because of the immediate decrease in power from the dropped rod. Figure 7-3 shows the decrease in power from the dropped rod ((

                             }}2(a),(c) as a function of rod worth for a variety of initial power levels. The figure is based on a representative NPM core design and also overlays a representative high power rate trip. ((
                                                                                }}2(a),(c) Sensitivity studies were performed to confirm this behavior using several representative NPM core designs, a range of initial power levels, and a variety of rod worths. ((
                                                                                                        }}2(a),(c) The alternative bounding method for the rod drop uses this result to screen rod drop cases into two groups: those that result in reactor trip within a short period and those that do not.

For the larger group of cases that result in the early reactor trip ((

                                                     }}2(a),(c), the dropped rod, and subsequent reactor trip, cause a decrease in global power. The dropped rod does cause an asymmetry and results in an increase in local power peaking. The subsequent reactor trip eliminates the asymmetry and the associated local power peaking.

Figure 7-4 shows the impact on the local power from the decrease in global power and the increase in local peaking for a representative NPM rod drop case that results in early reactor trip. ((

                              }}2(a),(c) The relevant acceptance criteria in Table 7-66 for the first group of the rod drop event can be assessed by the limiting steady-state subchannel analysis.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 For the smaller group of cases that do not result in the early reactor trip ((

                                                                                                }}2(a),(c)

However, because power overshoot may occur during the event, these cases are bounded instead by comparison to the withdrawal of a single CRA. ((

                                                                            }}2(a),(c) the conditions associated with the dropped rod cases are bounded by the limits used in the analysis of the withdrawal of a single CRA. Therefore, the limiting MCHFR and linear heat generation rate for the dropped rod cases that do not result in early reactor trip are bounded by the results of the single rod withdrawal analysis. This comparison is performed on a design-specific basis. A point that falls outside the single rod withdrawal analysis limits, ((
                                }}2(a),(c), would require use of the system transient approach described earlier in this section. When the design-specific comparison is bounding, the relevant acceptance criteria in Table 7-66 for the second group of the rod drop event can be assessed by the limiting single rod withdrawal subchannel analysis.

Misalignment of One or More CRAs The misalignment of CRAs occurs as a result of one or more CRAs being left behind when inserting or withdrawing the control bank. These conditions are not evaluated with NRELAP5 as part of the non-LOCA event methodology because this event is not a transient. Instead, the MCHFR is determined as part of a detailed subchannel evaluation. Table 7-66 Acceptance criteria, single active failure, loss of power scenarios - control rod misoperation Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR and maximum fuel centerline temperature MCHFR and maximum fuel centerline temperature are challenged during this reactivity anomaly event. Single failure of an ex-core flux detector This single failure is typically limiting because it delays time of reactor trip by requiring actuation based on responses of the least affected core quadrants. No loss of power A loss of AC power at event initiation is typically non-limiting (early reactor trip), while a loss of AC power coincident with reactor trip typically does not alter the MCHFR or fuel centerline temperature because the time of reactor trip is not changed. © Copyright 2022 by NuScale Power, LLC 529

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 7-3 Example power reduction vs. initial power and worth for dropped rod event at (( }}2(a),(c) ((

                                                                                               }}2(a),(c)

Figure 7-4 Example components of local power vs. time for dropped rod event ((

                                                                                               }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 7-5 Example comparison of non-tripped rod drop events to single rod withdrawal analysis limits ((

                                                                                              }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.15.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-67. Table 7-67 Acceptance criteria - control rod misoperation Acceptance Criteria Discussion Primary pressure This criterion is not an acceptance criterion listed in Section 15.4.3 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for primary pressure compared to other AOOs. Secondary pressure This criterion is not an acceptance criterion listed in Section 15.4.3 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for secondary pressure compared to other AOOs. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates, and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.15.3 Biases, Conservatisms, and Sensitivity Studies Withdrawal of a Single CRA The biases and conservatisms presented in Table 7-68 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature for the withdrawal of a single CRA event. © Copyright 2022 by NuScale Power, LLC 532

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the low condition. MTC Biased to BOC conditions. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Reactivity insertion rate Varied. RCS Temperature Control Automatic rod control N/A. Boron concentration Not credited.

                                                                                                             }}2(a),(c)

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal (Continued) Parameter Bias / Conservatism Basis (( PZR Pressure Control PZR spray Varied. PZR heaters Varied. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. (( }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. © Copyright 2022 by NuScale Power, LLC 534

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-69 Not Used Dropping One or More CRAs The biases and conservatisms presented in Table 7-70 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature for the dropped CRA(s) event. If the alternative bounding approach is used, where cases that result in early trip are bounded by steady-state condtions and cases that do not result in early trip are bounded by the single rod withdrawal, the biases and conservatisms in Table 7-70 associated with system transient analysis are not applicable. However, the alternative bounding approach considers combinations of initial core power, dropped CRA worth, core time-in-life, axial offset, flow, and temperature to screen the cases into groups and compare the non-tripped cases to single rod withdrawal analysis limits. Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition. Initial PZR level Nominal. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Varied. Kinetics Varied.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 535

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies (Continued) Parameter Bias / Conservatism Basis (( Decay heat Biased to the high condition. Initial SG pressure Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Dropped CRA worth Minimum RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited. PZR Pressure Control PZR spray Nominal. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 536

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies (Continued) Parameter Bias / Conservatism Basis (( Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. Table 7-71 Not Used 7.2.16 Inadvertent Decrease in Boron Concentration The methodology used to simulate an inadvertent decrease in boron concentration for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.16.1 General Event Description and Methodology The boric acid blend system incorporated into the NuScale plant design permits the operator to control the boron concentration of the reactor coolant via the charging fluid chemistry. While the NuScale plant design incorporates both automatic and manual controls, strict administrative procedures govern the process for adjusting the boron concentration of the reactor coolant. These administrative procedures establish limits on the rate and duration of the dilution. The primary means of causing an inadvertent decrease in boron concentration is failure of the blend system, either by controller or mechanical failure, or operator error. The event is terminated by isolating the source for the diluted water, i.e., by closing the demineralized water system (DWS) isolation valves. © Copyright 2022 by NuScale Power, LLC 537

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 For Mode 1 plant operating conditions, the perfect mixing model and the wave front model are both evaluated. The perfect mixing model is evaluated for Mode 1 operating conditions because it provides a slower reactivity insertion rate, delaying detection, potentially allowing further loss of shutdown margin. The wave front model is physically conservative because it assumes the maximum amount of reactivity as the diluted slug of water sweeps through the core. This model does not assume any axial blending to ensure that this reactivity insertion rate is maximized. For all other operating modes where boron dilution is allowed and limited mixing exists, a wave front model is used. These mixing models are generally performed as a hand calculation, but may be automated via a spreadsheet or other process. The following mathematical expression is used to determine the time required to erode the shutdown margin with the perfect mixing model. Equation 7-1 shows that the reactivity insertion rate depends on the dilution rate and the total RCS mass. 1 C i t dil = - ---- ln ----- Equation 7-1 K C f where: t dil = time required to dilute from the initial boron concentration to the final boron concentration, s K = -( Q in in ) ( 60V r r ) Q in = dilution flow rate of unborated water, gpm in = dilution water density, lbm/ft3 V r = effective water volume of the RCS, gal r = density of the water in the RCS, lbm/ft3 C i = initial boron concentration (maximum critical boron concentration including uncertainties), ppm C f = final boron concentration (boron concentration at which shutdown margin is lost), ppm The following mathematical expressions are used to determine the number of wave fronts and the time required to erode the shutdown margin with the wave © Copyright 2022 by NuScale Power, LLC 538

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 front model. In this model, the boron concentration of the RCS is reduced in discrete steps at each time, t, corresponding to the time the wave front passes through the core. The reactivity insertion rate is determined from the reactivity step change calculated as the product of the change in boron concentration and boron worth, divided by the core transport time. The core transport time is calculated as the total mass in the core divided by the RCS flow rate. As shown in Equation 7-2, the change in boron concentration is also inversely proportional to the RCS flow rate, therefore the ratio of total reactivity step change and core transport time makes the initial reactivity insertion rate independent of the RCS flow rate. N W NC C N = C i -------------------------------- Equation 7-2 ( W D + W NC ) M RCSI M RCS

                                                                      - + ( N - 1 ) --------------------------------

t = ------------------------------- Equation 7-3 ( W D + W NC ) ( W D + W NC ) where: C N = the Nth front boron concentration, ppm Ci = initial boron concentration, ppm W D = dilution mass flow rate, lbm/s W NC = natural circulation mass flow rate, lbm/s M RCS = RCS fluid mass minus the pressurizer, lbm M RCSI =initial pass RCS fluid mass (mass between the CVCS injection point to core inlet), lbm N = number of times the wave front passes through the core Mode 1 (Operations) HFP to 25 percent RTP In this mode of plant operation, an inadvertent decrease in boron concentration causes a reactivity insertion that increases reactor power, which leads to a rise in coolant temperature, pressurizer level, and RCS pressure. A loss of shutdown margin would occur quickest for the highest reactivity insertion rate, i.e., the maximum dilution flow rate of 50 gpm (2 CVCS pumps) with unborated water. © Copyright 2022 by NuScale Power, LLC 539

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 However, operation with two CVCS pumps is typically limited to certain conditions by administrative controls. For example, two pump operation may only be allowed at certain power levels or when critical boron concentration is below a certain threshold. As a result, the single pump dilution flow rate of 25 gpm is also evaluated. The reactivity insertion rates associated with these configurations are determined using both the perfect mixing model (Equation 7-1) and the wave front model (Equation 7-2 and Equation 7-3). The times of reactor trip and isolation of the dilution source via closure of the DWS isolation valves are obtained from the NRELAP5 results for these reactivity insertion rates. The calculations performed with the perfect mixing model are also used to determine the shutdown margin available after isolation of the DWS, and the time at which the shutdown margin would be lost if the dilution source is not terminated. The system responses for all other acceptance criteria, such as MCHFR and peak RCS pressure, are comparable to the uncontrolled control rod bank withdrawal at power event (Section 7.2.14) by comparison of the reactivity insertion rates. Mode 1 (Operations) HZP In this mode of plant operation, an inadvertent decrease in boron concentration causes a reactivity insertion that increases reactor power, but does not lead to a rise in coolant temperature, pressurizer level, or RCS pressure. A loss of shutdown margin would occur quickest for the highest reactivity insertion rate, considering the maximum dilution flow rate scenarios discussed above. The reactivity insertion rates associated with these configurations are determined using both the perfect mixing model (Equation 7-1) and the wave front model (Equation 7-2 and Equation 7-3). The time of reactor trip and isolation of the dilution source via closure of DWS isolation valves is obtained from the NRELAP5 results for these reactivity insertion rates. The calculations performed with the wave front model are also used to determine the shutdown margin available after isolation of the DWS, and the time at which the shutdown margin would be lost if the dilution source is not terminated. The system responses for all other acceptance criteria, such as MCHFR and peak RCS pressure, are comparable to the uncontrolled CRA bank withdrawal from subcritical or low power startup conditions event (Section 7.2.13 by comparison of the reactivity insertion rates. Mode 2 (Hot Shutdown) and Mode 3 (Safe Shutdown) In these modes of plant operation, the MPS protection logic ensures the DWS is isolated when the RCS flow rate is less than the low flow setpoint (1.7 ft3/s in the example Table 7-3). This protection scheme precludes the possibility for an inadvertent decrease in boron concentration. When the RCS flow rate is greater than or equal to the low flow setpoint (1.7 ft3/s in the example Table 7-3), the reactivity insertion from the maximum dilution flow rate of 25 gpm (1 CVCS pump) with unborated water causes an increase in reactor power (neutron population). The increase in neutron flux is detected by the MPS count rate protection signal and used to close the DWS isolation valves. The calculations performed with the wave front model (Equation 7-2 and Equation 7-3) © Copyright 2022 by NuScale Power, LLC 540

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 determine the shutdown margin available after isolation of the DWS, and the time at which the shutdown margin would be lost if the dilution source is not terminated. Mode 4 (Transition) In this mode of plant operation, all CVCS connections to an NPM are disconnected, isolated, or locked out. Thus, the possibility of a design-basis inadvertent decrease in boron concentration is precluded. Mode 5 (Refueling) In this mode of plant operation, the Technical Specifications require the pool boron concentration to be sufficient to have appropriate shutdown margin. For some NPM designs, the Technical Specifications also require the pool level to be maintained within a narrow range. Surveillance of the boron concentration, and level if applicable, of the refueling pool is performed at appropriate intervals to prevent significant inadvertent dilution from flow paths to the reactor pool, or proximate water sources such as fire mains or feedwater piping. The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-72. Table 7-72 Acceptance criteria, single active failure, loss of power scenarios - inadvertent decrease in boron concentration Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest MCHFR MCHFR is challenged during this reactivity anomaly event. (Reactivity insertion rates are insufficient to challenge fuel centerline temperature.) No single failure The challenging cases typically occur when all equipment is operational. No loss of power The challenging cases typically occur when AC power is available to the CVCS equipment for the event duration. 7.2.16.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-73. © Copyright 2022 by NuScale Power, LLC 541

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-73 Acceptance criteria - inadvertent decrease in boron concentration Acceptance Criteria Discussion Primary pressure This criterion is not evaluated because an NPM does not introduce more challenging conditions for primary pressure compared to other AOOs. Secondary pressure This criterion is not evaluated because an NPM does not introduce more challenging conditions for secondary pressure compared to other AOOs. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis; or by demonstrating that a loss of shutdown margin does not occur. Maximum fuel centerline temperature This criterion is not evaluated because the reactivity insertion rates are insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS (if actuated) pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.16.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-74 are considered in order to identify the bounding conditions for loss of shutdown margin. Table 7-74 Initial conditions, biases, and conservatisms - inadvertent decrease in boron concentration Parameter Bias / Conservatism Basis Initial reactor power Excluded. Not part of mixing model. Initial RCS average temperature Biased to the high condition. ((

                                                                                                             }}2(a),(c)

Initial RCS flow rate Biased to the low condition. ((

                                                                                                             }}2(a),(c)

Initial PZR pressure Nominal. ((

                                                                                                             }}2(a),(c)

Initial PZR level Excluded. Not part of mixing model. Initial feedwater temperature Excluded. Not part of mixing model. Initial fuel temperature Excluded. Not part of mixing model. MTC Excluded. Not part of mixing model. Kinetics Biased to BOC conditions. ((

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 542

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-74 Initial conditions, biases, and conservatisms - inadvertent decrease in boron concentration (Continued) Parameter Bias / Conservatism Basis Decay heat Excluded. Not part of mixing model. Initial SG pressure Excluded. Not part of mixing model. SG heat transfer Excluded. Not part of mixing model. RSV lift setpoint Excluded. Not part of mixing model. SG tube plugging Excluded. Does not alter active RCS volume. Shutdown margin Biased to the low condition. ((

                                                                                                             }}2(a),(c)

Initial boron concentration Biased to the high condition (( }}2(a),(c) Boron worth Biased to the high condition. ((

                                                                                                             }}2(a),(c)

Active RCS volume Biased to the low condition. ((

                                                                                                             }}2(a),(c)

Makeup flow rate Biased to the high condition. ((

                                                                                                             }}2(a),(c)

Makeup temperature Biased to the low condition. ((

                                                                                                             }}2(a),(c)

RCS Temperature Control Automatic rod control Excluded. Not part of mixing model. Boron concentration Not credited. ((

                                                                                                             }}2(a),(c)

PZR Pressure Control PZR spray (normal) Excluded. Not part of mixing model. (bypass) Excluded. Not part of mixing model. PZR heaters (non-prop.) Excluded. Not part of mixing model. (prop.) Excluded. Not part of mixing model. PZR Level Control Charging Enabled. ((

                                                                                                             }}2(a),(c)

Letdown Enabled. ((

                                                                                                             }}2(a),(c)

Steam Pressure Control Turbine throttle valves Excluded. Not part of mixing model. Turbine bypass valves Excluded. Not part of mixing model. © Copyright 2022 by NuScale Power, LLC 543

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-74 Initial conditions, biases, and conservatisms - inadvertent decrease in boron concentration (Continued) Parameter Bias / Conservatism Basis Feedwater and Turbine Load Control feedwater pump speed Excluded. Not part of mixing model. CNV Pressure Control CNV evacuation system Excluded. Not part of mixing model. ((

                                                                         }}2(a),(c)

Table 7-75 Not Used Table 7-76 Not Used Table 7-77 Not Used Table 7-78 Not Used Table 7-79 Not Used 7.2.17 Chemical and Volume Control System Malfunction that Increases Reactor Coolant System Inventory The methodology used to simulate a postulated CVCS malfunction that increases RCS inventory for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. © Copyright 2022 by NuScale Power, LLC 544

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.17.1 General Event Description The NuScale reactor does not have a pumped ECCS, therefore, the unplanned increase in RCS inventory event is caused by a malfunction of the CVCS makeup pumps or pressurizer level control system. If borated water at the same concentration of the primary system is added to the RCS, the addition of large amounts of water to the primary system typically generates a reactor trip on high pressurizer (PZR) water level or high PZR pressure. Table 7-80 lists the relevant acceptance criteria, SAF, and LOP scenarios. The malfunction is assumed to isolate letdown and actuate both makeup pumps (each flowing at their maximum capacity), causing an unplanned increase in RCS inventory. The limiting pressure response occurs when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. The increase in RCS inventory event is typically terminated by CVCS isolation on high PZR level. (The CVCS containment isolation valves are dual safety related valves.) Sensitivity studies to identify the challenging conditions are performed, as necessary, to identify the case(s) with the potentially limiting system pressures. Cases that include the effects of CVCS recirculation, which increases the flow into the RCS, are also analyzed. Table 7-80 Acceptance criteria, single active failure, loss of power scenarios - reactor coolant system inventory increase Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Primary pressure, secondary pressure System pressures are challenged during this mass addition event. No single failure The CVCS is isolated via dual safety-related isolation valves. If one of the isolation valves were to fail, the other CVCS isolation valve would provide system isolation. No loss of power Continued operation of the CVCS typically maximizes system pressures. 7.2.17.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-81. Table 7-81 Acceptance criteria - reactor coolant system inventory increase Acceptance Criteria Discussion Primary pressure Primary pressure rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. © Copyright 2022 by NuScale Power, LLC 545

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-81 Acceptance criteria - reactor coolant system inventory increase (Continued) Acceptance Criteria Discussion Secondary pressure Secondary pressure increases rapidly to the peak value upon turbine trip, then decreases during the cool down with DHRS. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. 7.2.17.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-82 are considered in identifying the bounding transient simulation for primary pressure. Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the low condition MTC Consistent with EOC kinetics. Kinetics Biased to EOC conditions. Decay heat Biased to the high condition.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 546

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase (Continued) Parameter Bias / Conservatism Basis (( Initial SG pressure(1) Nominal. Steam generator heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Makeup temperature Varied RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Nominal. PZR Level Control Charging N/A. Letdown Disabled.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 547

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase (Continued) Parameter Bias / Conservatism Basis (( Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control Feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                             }}2(a),(c)

1. (( }}2(a),(c)

2. These inputs, in conjunction with least negative Doppler temperature coefficient, are selected to maximize the power response (if any) induced by the addition of colder CVCS water. However, since this event is driven by mass addition, reactivity effects prior to RTS actuation (if any) are small when compared to the pressurization associated with the increase in primary inventory.

Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event, including cases with the highest pressure responses for this inventory increase event. Table 7-83 Not Used 7.2.18 Failure of Small Lines Outside Containment The methodology used to simulate a postulated failure of a small line connected to the primary coolant system outside of containment for an NPM, and an evaluation of © Copyright 2022 by NuScale Power, LLC 548

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 the acceptance criteria for an infrequent event listed in Table 7-4, are presented below. A postulated break in a small line carrying primary coolant is typically only evaluated for radiological consequences. Neither the plant design nor the use of natural circulation flow for an NPM introduces a more challenging condition for other acceptance criteria. Evaluation of postulated small line breaks within the context of a 10 CFR 50 Appendix K regulatory LOCA analysis is covered separately in the LOCA-EM topical report. 7.2.18.1 General Event Description and Methodology The event is initiated by a break in a line connected to the primary coolant system outside containment that causes a decrease in pressurizer pressure and level. The rate of decrease for both parameters depends on the break location and size. The subsequent decrease in RCS pressure provides little core reactivity from moderator feedback, so the reactor power remains relatively constant until reactor trip. In the absence of a loss of power at event initiation, the decreasing pressurizer pressure and level causes RTS actuation on the low pressurizer pressure signal or the low pressurizer level signal. Regardless of the presence of a reactor trip, the sustained loss of reactor coolant from the break causes a continuous decrease in pressurizer pressure and level. Eventually an MPS low pressurizer pressure or level signal is generated. Depending on design, one or both of these MPS signals leads to a reactor trip (if not previously generated), DHRS actuation, and CVCS isolation. Closure of the FWIVs and MSIVs following DHRS actuation isolates the steam generators, while CVCS isolation terminates the loss of reactor coolant through the break (assumed to be in the CVCS system since this is the only primary coolant system pipe line that penetrates the containment). Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient terminates with the NPM in a safe, stable condition. Several lines within the CVCS qualify as candidates for evaluation. These lines include makeup lines (CVCS injection); letdown lines (CVCS discharge); pressurizer spray lines; and, the degassing (high point vent) line. However, depending on the design, the degassing line may be eliminated from evaluation based on similarity with the spray lines if: 1) the spray lines and degassing line are the same size; 2) both line types connect to the top of the pressurizer; and, 3) both line types penetrate the containment head. From a system response perspective, the response to a break in a spray line is nearly the same as the response to a break in the degassing line. Thus, a break in the degassing line may not be required to be explicitly modeled. When determining the magnitude of mass released for a break in a CVCS line outside containment the timing of CVCS isolation is critical, as CVCS isolation stops the release of mass from the RPV into the reactor building. Maximizing the © Copyright 2022 by NuScale Power, LLC 549

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 mass released from the RPV and the duration of the iodine spike maximizes the radiological consequences. Section 4.3.6 describes an example approach for performing the radiological consequence analysis using bounding inputs for the mass release and isolation time. If a bounding approach is used, then system transient results for mass release and isolation time are not required for the downstream radiological analysis. System transient analysis can be used to confirm the radiological analysis assumptions are bounding, if necessary. The nature of a spray line or degassing line break differs from that for a break in either the makeup line or letdown line because vapor is expelled from the break instead of liquid. As a result, steam production in the pressurizer is a key phenomenon for a spray line or degassing line break, but much less important for makeup or letdown line breaks. A brief description of each break location is presented below. Spray and Degassing Line Breaks Calculations are performed to assess the responses to a spectrum of break sizes. Specifically, each break is modelled as either a double-ended guillotine rupture, or a smaller size break, of the piping. ((

                                                                                                 }}2(a),(c) A break with an area of 100 percent of the line causes a rapid depressurization that quickly trips the reactor and isolates the break. As such, a break of this size is not typically limiting. While the rate of depressurization declines proportionally as the break area is reduced, a break area within the steam production capability of the pressurizer heaters allows the pressurizer pressure to be maintained until the reactor trips on a low pressurizer level signal. Due to reactor coolant volume shrinkage following reactor trip, the rate of decrease for both the pressurizer pressure and level is nearly independent of break flow, and any reduction in pressure reduces the mass released through the break. Regardless of the break size, a break in these lines is typically less limiting from a dose perspective than a break in either a makeup line or a letdown line because vapor is released instead of liquid, with no significant difference in event duration or iodine spiking time, thus the total mass released is smaller.

Makeup and Letdown Line Breaks The evaluation of the makeup and letdown line breaks is divided into two phases:

1. Before Reactor Trip This phase lasts until the reactor trips. During this phase, the total break mass release depends primarily on the liquid density at the location of the break.

The break is modelled as either a double-ended guillotine rupture, or a smaller size break, of the line piping. (( }}2(a),(c) © Copyright 2022 by NuScale Power, LLC 550

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 ((

                               }}2(a),(c)

2. After Reactor Trip This phase lasts from the time of reactor trip until isolation of the CVCS.

During this phase, the pressurizer level decreases independently of the break flow due to the mismatch between heat generation in the core and heat removal via the steam generators. Since the rate of decrease for both the pressurizer pressure and level is nearly independent of break flow, and any reduction in pressure reduces the mass released through the break, the mass released is maximized by increasing the break area to include both lines prior to CVCS isolation. In contrast, the spiking time is maximized when the break is restricted to a single location. Table 7-84 lists the relevant acceptance criteria, single active failure, and loss of power scenarios. Table 7-84 Acceptance criteria, single active failure, loss of power scenarios - breaks in small lines carrying primary coolant outside containment Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Radiological consequences A postulated break in a small line carrying primary coolant is evaluated for radiological consequences. No single failure The isolation valves on the makeup, letdown, and spray lines are safety grade and redundant. Therefore, failure of a single valve does not prevent isolation or significantly increase the radiological consequences. Loss of AC at event initiation The loss of heat transfer to the secondary system associated with a loss of AC power at event initiation typically results in the most challenging integrated mass release and spiking time. 7.2.18.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-85. © Copyright 2022 by NuScale Power, LLC 551

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-85 Acceptance criteria - breaks in small lines carrying primary coolant outside containment Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the small line breaks carrying primary coolant event, sensitivities that maximize primary pressure are not analyzed. The peak primary pressure experienced during this event does not vary significantly from the pressure at event initiation, unless a pressure increase occurs due to an assumed loss of power. In all instances, the peak primary pressure remains below the design pressure, thereby providing margin to the acceptance criterion. If a system transient analysis is not required as input for the radiological consequence analysis, maximum primary pressure can be selected as the relevant acceptance criterion to address with the system transient analysis. Secondary pressure Secondary pressure increases rapidly post-DHRS actuation. However, due to the depressurizing nature of this event, sensitivities that maximize peak secondary pressure are not analyzed. In all instances the peak secondary pressure experienced during this event remains below the design pressure, thereby providing margin to the acceptance criterion. Fuel cladding integrity Fuel failure directly relates to dose by dictating the isotopic concentration of the reactor coolant being released. Consequently, the MCHFR and fuel centerline temperature criteria are relevant to determining the appropriate source term for the downstream radiological analysis. The depressurization following a break in a small line carrying primary coolant is slow enough to preclude an increase in core power or significant reduction in core flow prior to reactor trip. After considering the NuScale CHF trend with pressure, and the ranges bounded by the steady-state subchannel analysis, the CHFR is not challenged during the transient prior to reactor trip. Therefore, the fuel cladding integrity acceptance criteria are not challenged and event-specific follow-on MCHFR and fuel centerline temperature evaluations are not necessary. Containment integrity Containment integrity is evaluated in a separate analysis methodology. © Copyright 2022 by NuScale Power, LLC 552

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-85 Acceptance criteria - breaks in small lines carrying primary coolant outside containment (Continued) Acceptance Criteria Discussion Consequential loss of system functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, the water level in the RPV remains above the top of the core throughout the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. Radiological consequences The radiological consequences acceptance criteria are evaluated by downstream radiological analysis using the mass release calculated in the non-LOCA transient analysis or are evaluated with bounding assumptions as described in Section 4.3.6. 7.2.18.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-86 are considered in identifying a bounding transient simulation for dose. Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment Parameter Bias / Conservatism Basis(1) (( Initial reactor power Varied. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition Initial PZR level Biased to the high condition. Initial feedwater temperature Varied.

                                                                                                             }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 553

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment (Continued) Parameter Bias / Conservatism Basis(1) (( Initial fuel temperature Varied. MTC Varied. Kinetics Nominal. Decay heat Biased to the high condition. Initial SG pressure(2) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. CVCS volume outside Biased to the high condition. containment RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 554

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment (Continued) Parameter Bias / Conservatism Basis(1) (( PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

                                                                                                         }}2(a),(c)

1. ((

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© Copyright 2022 by NuScale Power, LLC 555

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Table 7-87 Not Used Table 7-88 Not Used 7.2.19 Steam Generator Tube Failure The methodology used to simulate a postulated failure of a steam generator tube for an NPM, and an evaluation of the acceptance criteria for an accident listed in Table 7-4, are presented below. A postulated failure of a steam generator tube is typically only evaluated for radiological consequences. Neither the steam generator design nor the use of natural circulation flow for an NPM introduce a more challenging condition for other acceptance criteria. 7.2.19.1 General Event Description and Methodology The event is initiated by the failure of a steam generator tube that causes a decrease in pressurizer pressure and level. The rate of decrease for both parameters depends on the break location and size. The subsequent decrease in RCS pressure provides minimal core reactivity feedback from moderator feedback, so the reactor power remains relatively constant until reactor trip. In the absence of a loss of power at event initiation, the decreasing pressurizer pressure and level causes RTS actuation on the low pressurizer pressure signal or the low pressurizer level signal. Regardless of the presence of a reactor trip, the sustained loss of reactor coolant from the failed tube causes a continuous decrease in pressurizer pressure and level. Eventually an MPS low pressurizer pressure or level signal is generated. Depending on the design, one or both of these MPS signals actuates DHRS and closure of the FWIVs and MSIVs isolates the steam generator, which also terminates the loss of reactor coolant from the failed tube to the environment. Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient ends with the NPM in a safe, stable condition. © Copyright 2022 by NuScale Power, LLC 556

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 When determining the magnitude of mass released for a steam generator tube failure (SGTF) event the timing of SG isolation is critical, since SG isolation terminates the release of mass from the RPV to other plant areas. Maximizing the mass released to the environment and the duration of the iodine spike (elapsed time from reactor trip to steam generator isolation) maximizes the radiological consequences. In cases where SG isolation occurs on the same signal as RTS, spiking time is restricted to only the time required for valves to close. Section 4.3.6 describes an example approach for performing the radiological consequence analysis using bounding inputs for the mass release and isolation time. If a bounding approach is used, then system transient results for mass release and isolation time are not required for the downstream radiological analysis. System transient analysis can be used to confirm the radiological analysis assumptions are bounding, if necessary. Calculations are performed to assess the responses to a spectrum of break sizes. Specifically, each break is modelled as either a double-ended guillotine rupture, or a smaller size break, of the steam generator tube. ((

                           }}2(a),(c) The results of sensitivity studies on break type and location indicate a rupture of the steam generator tube at the top of the steam generator typically provides the greatest integrated mass released and the longest spiking time. While the rate of depressurization declines proportionally as the break area is reduced, the rate of decrease after reactor trip for both the pressurizer pressure and level is nearly independent of break flow, and any reduction in pressure reduces the mass released through the break.

An SGTF event initiated from HFP leads to a higher break flow, and thus a higher integrated mass released, because the pressure difference between the primary and secondary system is larger for longer. Similarly, an SGTF event initiated from HFP typically leads to a longer spiking time because the stored energy of the core is greater. Table 7-89 lists the relevant acceptance criteria, single active failure, and loss of power scenarios. Table 7-89 Acceptance criteria, single active failure, loss of power scenarios - steam generator tube failure Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Radiological consequences A postulated failure of a steam generator tube is evaluated for radiological consequences. © Copyright 2022 by NuScale Power, LLC 557

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-89 Acceptance criteria, single active failure, loss of power scenarios - steam generator tube failure (Continued) Acceptance Criteria / Single Active Failure / Loss Discussion of Power Scenarios of Interest Failure of primary MSIV to close on affected SG The primary MSIV is typically assumed to fail to isolate to maximize the radiological consequences. Delaying isolation until closure of the secondary MSIV leads to additional mass released to the environment and a longer iodine spike duration. No loss of power A loss of AC power at event initiation or coincident with turbine trip typically results in less integrated mass released and a shorter iodine spiking duration. 7.2.19.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-90. Table 7-90 Acceptance criteria - steam generator tube failure Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the steam generator tube failure event, sensitivities that maximize primary pressure are not analyzed. However, sensitivities were performed to demonstrate that, with the exception of those cases that involve a loss of power at event initiation, the peak primary pressure experienced during this event does not vary significantly from the pressure at event initiation. In all instances (i.e., with and without a loss of AC power at event initiation) the peak primary pressure remains below the design pressure, thereby providing margin to the acceptance criterion. If a system transient analysis is not required as input for the radiological consequence analysis, maximum primary pressure can be selected as a relevant acceptance criterion to address with the system transient analysis. © Copyright 2022 by NuScale Power, LLC 558

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-90 Acceptance criteria - steam generator tube failure (Continued) Acceptance Criteria Discussion Secondary pressure Pressure in the portion of the secondary system between the FWIVs and MSIVs increases rapidly post-DHRS actuation for both SGs, and the failed SG tube allows the pressure in the affected SG to approach RCS pressure conditions. In all instances, i.e., with and without a loss of AC power at event initiation, the peak secondary system pressure remains below the design pressure, thereby providing margin to the acceptance criterion. If a system transient analysis is not required as input for the radiological consequence analysis, maximum secondary pressure can be selected as a relevant acceptance criterion to address with the system transient analysis. Fuel cladding integrity Fuel failure directly relates to dose by dictating the isotopic concentration of the reactor coolant being released. Consequently, the MCHFR and fuel centerline temperature criteria are relevant to determining the appropriate source term for the downstream radiological analysis. The SGTF is a slow depressurization event that does not result in an increased core power or significantly reduced core flow prior to reactor trip. After considering the NuScale CHF trend with pressure and the ranges bounded by steady-state subchannel analysis, the CHFR is not challenged during the transient prior to reactor trip. Therefore, the fuel cladding integrity acceptance criteria are not challenged and event-specific follow-on MCHFR and fuel centerline temperature evaluations are not necessary. Containment integrity Containment integrity is evaluated in a separate analysis methodology. Consequential loss of system functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, the water level in the RPV remains above the top of the core throughout the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. Radiological consequences The radiological consequences acceptance criteria are evaluated by downstream radiological analysis using the mass release calculated in the non-LOCA transient analysis or are evaluated with bounding assumptions as described in Section 4.3.6. © Copyright 2022 by NuScale Power, LLC 559

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 7.2.19.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-91 are considered in identifying a bounding transient simulation for dose. Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure Parameter Bias / Conservatism Basis (1) (( Initial reactor power RTP biased to the high condition. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Biased to the high condition Initial PZR level Biased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Biased to the high condition. MTC Varied. Kinetics Varied. Decay heat Biased to the high condition.

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© Copyright 2022 by NuScale Power, LLC 560

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure (Continued) Parameter Bias / Conservatism Basis (1) (( Initial SG pressure(2) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Varied. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled.

                                                                                                            }}2(a),(c)

© Copyright 2022 by NuScale Power, LLC 561

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure (Continued) Parameter Bias / Conservatism Basis (1) (( Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

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((

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Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Table 7-92 Not Used © Copyright 2022 by NuScale Power, LLC 562

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.0 Representative Calculations The methodology of Chapter 7 is utilized in conjunction with the NRELAP5 model of Chapter 6 to provide representative transient results. The calculations are for a representative NPM design, consistent with the design characteristics in Chapter 3. The transients noted below were selected to demonstrate the application of the NuScale Non-LOCA methodology for analysis of the representative plant responses to a wide range of postulated equipment failures and malfunctions.

1. Cooldown and/or Depressurization of the RCS (Section 8.1)

2. Heatup and/or Pressurization of the RCS (Section 8.2)

3. Reactivity Anomaly (Section 8.3)

4. Increase in RCS Inventory (Section 8.4)

5. Decrease in RCS Inventory (Section 8.5)

The information included for each representative transient includes: an event description; the results for the acceptance criteria of interest; and, conclusions regarding the acceptance criteria of interest. These results are presented to demonstrate the application of the non-LOCA methodology to a representative NPM. Fuel rod and core physics parameter inputs for the representative transients were developed using COPERNIC (Reference 22) and SIMULATE5 (Reference 23) respectively. 8.1 Cooldown and/or Depressurization of the Reactor Coolant System 8.1.1 Decrease in Feedwater Temperature The purpose of this section is to present the thermal-hydraulic response of a representative NPM for a decrease in feedwater temperature event. This event is evaluated for MCHFR. 8.1.1.1 Event Description The general decrease in feedwater temperature (DFWT) event description can be found in Section 7.2.1.1. Based on Section 7.2.1.1, MCHFR is the only acceptance criterion that may be potentially challenged during the DFWT event. No single failure is applied since the challenging cases occur when all equipment operates as designed. No loss of power is applied since all loss of power scenarios terminate feedwater or trip the reactor, thus reducing the overcooling event. Chosen from a series of MCHFR sensitivity cases, the representative DFWT case presented here represents a case that could challenge MCHFR, based on the NRELAP5 MCHFR pre-screening. This case features the following conditions: Conservative initial condition biasing (as shown in Table 7-7) is applied in order to maximize the consequences of the overcooling event in terms of MCHFR. This representative case is initialized at 102 percent reactor power. RCS average temperature is biased at high condition (555 degrees F). RCS © Copyright 2022 by NuScale Power, LLC 563

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 flow rated is biased to the low condition (535 kg/s). Pressurizer pressure is biased to the high condition (1920 psia). Pressurizer level is biased to the high condition (53 percent). Initial feedwater temperature is biased to the high condition (307.5 degrees F). Transient feedwater temperature decreases linearly from 307.5 degrees F at the rate of 1.18 degrees F per second which minimize the MCHFR based on the MCHFR pre-screening process. Feedwater flow is maintained at a constant volumetric flow rate. This prevents the controller from reducing feedwater flow in response to decreasing feedwater temperature which maximizes RCS overcooling. Turbine trip on reactor trip and subsequent operation of the turbine bypass system is not credited in this case, rather the turbine boundary is conservatively held constant at the pre-reactor trip condition in order to maximize the overcooling of the RCS. EOC reactivity coefficients are applied which maximizes the reactor power response. A low fuel temperature bias (applied by increasing gap conductance) is applied in this case although a nominal temperature is acceptable. No operator action was credited in the representative case. Normal control system such as PZR spray, heater, letdown controls and automatic rod control are modeled based on the control status shown in Table 7-7. 8.1.1.2 Analysis Results The following describes the event sequence of the representative DFWT event. Table 8-1 summarizes the sequence of events. Figure 8-1 through Figure 8-9 show some key parameters during the representative DFWT event. The DFWT event begins at time zero. The hydraulic feedwater source boundary condition is linearly changed from a constant temperature at 307.5 degrees F at a rate of 1.18 degrees F per second (Figure 8-1). Since the feedwater flow control is held at a constant volumetric pump rate, decreasing feedwater temperature means feedwater density and mass flow rate increases. The RCS response to the overcooling event begins once the cold feedwater front propagates through the secondary system piping and reaches the SG. As heat removal from the RCS through the SG increases above its steady state value, downcomer temperature begins to decrease. The drop in average RCS temperature prompts the control rod controller to begin pulling the regulating bank out of the core at ~28 seconds. The addition of positive reactivity causes reactor power to increase. This increase in power is significant enough that moderator feedback stays slightly negative in response to the core heatup. Power and RCS riser temperature continue to rise until the high RCS riser temperature analytical limit (610 degrees F) is reached at ~139 seconds (Figure 8-2 and Figure 8-3). Because there is a total 8 second delay between the high RCS riser temperature © Copyright 2022 by NuScale Power, LLC 564

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 signal and reactor scram, power continues to rise until reaching the high power limit (125 percent RTP) at ~145 seconds; there is a 2 second delay between the high power signal and reactor scram. Both the high power signal and the high RCS riser temperature signal initiate reactor scram at ~147 seconds, which terminates the power excursion. Power peaks at 200.6 MWth before reactor scram. The high RCS riser temperature signal also initiates DHRS actuation at the same time as reactor scrams. This closes the FWIVs and MSIVs, which isolate the SG from the remaining secondary system and ends the overcooling transient. Steam generator pressure increase resulting from main steam isolation is expected and is not a direct consequence of the decrease in feedwater temperature event itself. SG pressure starts to decrease once the DHRS cooling is established (Figure 8-4). RCS pressure and level have an overall decreasing trend as inventory shrinks due to increased heat removal during the overcooling transient. Once the PZR level is lower than the low PZR level analytical limit (35 percent) at

                  ~725 seconds, PZR heaters are disabled and RCS pressure continues to decrease after that (Figure 8-5 and Figure 8-6).

After reactor trip and actuation of DHRS, RCS flow decreases rapidly and becomes stagnant and slightly reversed at ~180 seconds (Figure 8-7). Following that, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2 and shown in Figure 8-3, Figure 8-7, and Figure 8-8); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. At

                  ~30 minutes, RCS flow has stabilized, and the RCS temperature and pressure are steadily decreasing as the DHRS transfers decay heat from the RPV to the reactor pool (Figure 8-3, Figure 8-5, Figure 8-7, and Figure 8-8). It is concluded that by 30 minutes the overcooling transient has been terminated and that stable DHRS cooling has been achieved. Subcritical margin is verified as net reactivity remains less than 0.0 dollars at the time stable DHRS cooling has been achieved (Figure 8-9). No operator action was credited to mitigate this event.

Table 8-1 Decrease in feedwater temperature sequence of events Event Time (sec) Malfunction that initiates the decrease in feedwater temperature event. Feedwater 0 temperature is linearly decreased from 307.5°F at a rate of 1.18°F per second. Cold water front reaches core inlet. Regulating bank begins to withdraw in response to a decrease in average RCS temperature. Reactor power begins to 28 rise. High RCS riser temperature limit is reached (610°F). Control rod insertion begins 139 after 8 second delay. Peak RPV pressure is reached (1951 psia). 145 High reactor power limit is reached (125% RTP). Control rod insertion begins after 145 2 second delay. Peak reactor power is reached (200.6 MW). 147 © Copyright 2022 by NuScale Power, LLC 565

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 8-1 Decrease in feedwater temperature sequence of events (Continued) Event Time (sec) RTS actuation on both high power and high RCS riser temperature signals, control 147 rods begin to insert into the core. DHRS actuation on the high RCS riser temperature signal. DHRS actuation valves 147 open immediately. Limiting MCHFR is reached (2.591 as calculated by NRELAP5). 148 FWIVs and MSIVs are fully closed. Feedwater flow stops. 152 RCS flow is stagnant and slightly reversed ~ 180 Peak SG pressure is reached (1463 psia). 193 Low PZR level limit is reached (35%). PZR heaters are disabled after 3 second 725 delay. Low low PZR level is reached (20%). Containment and CVCS isolation begins 1767 after 3 second delay. Establishment of stable RCS flow. Pressure and temperature are steadily 1800 decreasing. End of calculation. Stable DHRS cooling has been established. Net reactivity 2700 remains < $0.0. Figure 8-1 Temperature of feedwater during the representative decrease in feedwater temperature event © Copyright 2022 by NuScale Power, LLC 566

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-2 Power response for the representative decrease in feedwater temperature event Figure 8-3 Core outlet temperature for the representative decrease in feedwater temperature event © Copyright 2022 by NuScale Power, LLC 567

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-4 Steam generator 2 pressure response for the representative decrease in feedwater temperature event Figure 8-5 Reactor pressure vessel pressure response for the representative decrease in feedwater temperature event © Copyright 2022 by NuScale Power, LLC 568

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-6 Pressurizer level for the representative decrease in feedwater temperature event Figure 8-7 Reactor coolant system flow rate for the representative decrease in feedwater temperature event © Copyright 2022 by NuScale Power, LLC 569

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-8 Core inlet temperature for the representative decrease in feedwater temperature event Figure 8-9 Net reactivity for the representative decrease in feedwater temperature event © Copyright 2022 by NuScale Power, LLC 570

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.1.1.3 Conclusion A representative case that could challenge MCHFR was identified for a decrease in feedwater temperature event. The results of this case, as presented in Section 8.1.1.2, are subsequently used as input to an MCHFR evaluation using the NuScale subchannel analysis methodology. 8.1.2 Increase in Steam Flow The purpose of this section is to present the thermal-hydraulic response of a representative NPM for an increase in steam flow event. This event is evaluated for MCHFR. 8.1.2.1 Event Description The general increase in steam flow event description can be found from Section 7.2.3.1. Based on Section 7.2.3.1, MCHFR is the only acceptance criterion that may be potentially challenged during the increase in steam flow event. No single failure is applied since the challenging cases occur when all equipment operates as designed. No loss of power is applied since all loss of power scenarios terminate feedwater or trip the reactor, thus reducing the overcooling event. Chosen from a series of MCHFR sensitivity cases, the representative increase in steam flow case presented here represents a case that could challenge MCHFR, based on the NRELAP5 MCHFR pre-screening. This case features the following conditions: Conservative initial condition biasing (as shown in Table 7-19) is applied in order to maximize the consequences of the overcooling event in terms of MCHFR. This representative case is initialized at 102 percent reactor power. RCS average temperature is biased at high condition (555 degrees F). RCS flow rated is biased to the low condition (535 kg/s). Pressurizer pressure is biased to the high condition (1920 psia). Pressurizer level is biased to the high condition (53 percent). SG heat transfer is decreased 30 percent by applying a heat transfer coefficient multiplier of 0.7 in the steady state initialization model. As identified in Section 7.2.3.3, this biasing has insignificant impact on the overall limiting MCHFR conditions for the transient. Steam flow is increased 14.45 percent instantly at the beginning of the event. A time-dependent junction that controls steam mass flow rate is used to model the turbine. During the increase in steam flow event, the feedwater pump speed remains constant and the pump curve allows a 1.0 lbm/s increase in feedwater flow for every 1 psi decrease in SG pressure. This maximizes the overcooling event by increasing the available source of secondary coolant. EOC reactivity coefficients are applied which maximizes the reactor power response. © Copyright 2022 by NuScale Power, LLC 571

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 A low fuel temperature bias (applied by increasing gap conductance) is applied in this case although a nominal temperature is acceptable. No operator action was credited in the representative case. Normal control system such as PZR spray, heater, letdown controls and automatic rod control are modeled based on the control status shown in Table 7-19. 8.1.2.2 Analysis Results The following describes the event sequence of the representative increase in steam flow event. Table 8-2 summarizes the sequence of events. Figure 8-10 through Figure 8-19 show some key parameters during the representative increase in steam flow event. The increase in steam flow event begins at time zero, as can be seen in Figure 8-10. The main steam flow rate at the secondary system hydraulic exit boundary is increased by 14.45 percent. Figure 8-11 shows that decreasing pressure in the secondary system also causes an increase in feedwater pump flow due to the pump curve, as shown in Figure 8-12. The RCS response to the overcooling event begins once steam flow through the steam generators starts to increase. As heat removal from the RCS through the SG increases above its steady state value, downcomer temperature begins to decrease (Figure 8-13). The drop in average RCS temperature prompts the CR controller to pull the regulating bank out of the core at ~5 seconds. The addition of positive reactivity causes reactor power to increase, as seen in Figure 8-14. Power continues to rise until peaking at

                  ~58 seconds. Peak power is calculated to be 199.97 MW, which is slightly lower than the high power analytical limit (200 MW or 125 percent RTP - increased from the analytical limit of 120 percent to conservatively account for the decalibration of the excore neutron detectors as downcomer density increases in response to an overcooling event). Therefore the representative increase in steam flow event is not tripped on high power. Soon after, the limiting MCHFR is reached at ~62 seconds. Figure 8-15 shows the RCS riser temperature, which continues to rise until the high RCS riser temperature limit (610 degrees F) is reached at ~68 seconds. There is a total of 8 second delay between the high RCS riser temperature signal and RTS/DHRS actuation. Once RTS and DHRS actuate, the subsequent reactor scram and steam generator isolation terminate the overcooling event at ~76 seconds.

DHRS actuation closes the FWIVs and MSIVs, which isolate the SG from the remaining secondary system and ends the overcooling transient. Steam generator pressure increase resulting from MS isolation is expected and is not a direct consequence of the increase in steam flow event itself. SG pressure starts to decrease once the DHRS cooling is established (Figure 8-11). RCS pressure and level have an overall decreasing trend as inventory shrinks due to increased heat removal during the overcooling transient. Once the PZR level is lower than the low © Copyright 2022 by NuScale Power, LLC 572

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 PZR level analytical limit (35 percent) at ~670 seconds, PZR heaters are disabled and RCS pressure continues to decrease after that (Figure 8-16 and Figure 8-17). After reactor trip and actuation of DHRS, RCS flow decreases rapidly and becomes stagnant and slightly reversed at ~112 seconds (Figure 8-18). Following that, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2 and shown in Figure 8-13, Figure 8-15, and Figure 8-18); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. By 40 minutes, RCS flow has stabilized, and the RCS temperature and pressure are steadily decreasing as the DHRS transfers decay heat from the RCS to the reactor pool (Figure 8-13, Figure 8-15, Figure 8-16 and Figure 8-18). It is concluded that by 40 minutes the transient has been terminated and that stable DHRS cooling has been achieved. Subcritical margin is verified as net reactivity remains less than 0.0 dollars at the time stable DHRS cooling has been achieved (Figure 8-19). No operator action was credited to mitigate this event. Table 8-2 Increase in steam flow sequence of events Event Time (sec) An instant increase of 14.45% is applied to the steam flow to model the spurious 0 opening of the turbine bypass valve or the main steam safety valve Cold water front reaches core inlet. Regulating bank begins to withdraw in response to a decrease in average RCS temperature. Reactor power begins to 5 rise. Peak reactor power is reached (199.97 MW). 58 Limiting MCHFR is reached (3.570 as calculated by NRELAP5). 62 High RCS riser temperature limit is reached (610°F). 68 Actuation of the RTS. Control rods begin to insert into the core. 76 Actuation of the DHRS. The DHRS actuation valves open immediately. The FWIVs 76 and MSIVs begin closing. Peak RPV pressure is reached (1991 psia). 77 FWIVs and MSIVs are fully closed. Steam generator isolation from the remaining 81 secondary system terminates the overcooling transient. RCS flow is stagnant and slightly reversed. 112 Peak steam generator pressure is reached (1248 psia). 128 Low PZR level limit is reached (35%). PZR heaters are disabled after 3 second 670 delay. Low low PZR level is reached (20%). Containment and CVCS isolation begins 1620 after a 5 second delay. End of calculation. Stable DHRS cooling has been established. Net reactivity 2400 remains < $0.0. © Copyright 2022 by NuScale Power, LLC 573

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-10 Main steam transient flow rate during the representative increase in steam flow event Figure 8-11 Steam generator 2 pressure response for the representative increase in steam flow event © Copyright 2022 by NuScale Power, LLC 574

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-12 Steam generator 2 secondary side flow for the representative increase in steam flow event Figure 8-13 Core inlet temperature for the representative increase in steam flow event © Copyright 2022 by NuScale Power, LLC 575

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-14 Power response for the representative increase in steam flow event Figure 8-15 Core outlet temperature for the representative increase in steam flow event © Copyright 2022 by NuScale Power, LLC 576

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-16 Reactor pressure vessel pressure response for the representative increase in steam flow event Figure 8-17 Pressurizer level for the representative increase in steam flow event © Copyright 2022 by NuScale Power, LLC 577

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-18 Reactor coolant system flow rate for the representative increase in steam flow event Figure 8-19 Net reactivity for the representative increase in steam flow event © Copyright 2022 by NuScale Power, LLC 578

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.1.2.3 Conclusion A representative case that could challenge MCHFR was identified for an increase in steam flow event. The results of this case, as presented in Section 8.1.2.2, are subsequently used as input to an MCHFR evaluation using the NuScale subchannel analysis methodology. 8.1.3 Main Steam Line Break The purpose of this section is to present the thermal-hydraulic response of a representative NPM for a main steam line break event. This event is evaluated for MCHFR, and mass releases are determined for input to downstream accident radiological dose analysis. A representative case evaluated for MCHFR is presented. 8.1.3.1 Event Description The general description for main steam line break event can be found from Section 7.2.4.1. Based on Section 7.2.4.1, MCHFR is the only acceptance criterion that may be potentially challenged during the main steam line break event. The MCHFR case assumes a failed MSIV on the affected SG train; however, the timing of MCHFR is well before the MSIV would have closed so it is concluded there is no limiting failure for the MCHFR case. No loss of power is applied since all loss of power scenarios terminate feedwater or trip the reactor, thus reducing the overcooling event. Chosen from a series of MCHFR sensitivity cases, the representative main steam line break case presented here represents a case that could challenge MCHFR, based on the NRELAP5 MCHFR pre-screening. This case features the following conditions: Conservative initial condition biasing (as shown in Table 7-24) is applied in order to maximize the consequences of the overcooling event in terms of MCHFR. The case is initialized at 102 percent reactor power with conservatively high RCS temperature and pressure. RCS average temperature is biased at high condition (555 degrees F). RCS flow rated is biased to the low condition (535 kg/s). Pressurizer pressure is biased to the high condition (1920 psia). Pressurizer level is biased to the high condition (58 percent). Feedwater temperature is biased to the high condition (310 degrees F). Minimum RCS design flow is assumed. SG heat transfer is increased 30 percent by applying a heat transfer coefficient multiplier of 1.3 in the steady state initialization model. As identified in Section 7.2.4.3, this biasing has insignificant impact on the overall limiting MCHFR conditions, spiking time or mass released. There is no tube plugging in the SGs for this representative calculation. The feedwater controller is based on FW pressure error rather than turbine load demand. This allows for the implementation of the feedwater pump flow response to the pressure loss events. In the steam line piping failure transient, the details of how the pumps speed controller will respond are ignored and © Copyright 2022 by NuScale Power, LLC 579

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 bounded by a conservative pump curve to maximize the flow response due to the drop in FW pressure. Steam pressure control is a flow based controller rather than a back pressure control. This is a better physical representation of the turbine response to a drop in steam pressure. During the transients the turbine is treated as a constant steam flow sink without automated runback. A low fuel temperature bias (applied via increase gap conductance) is applied which minimizes negative Doppler feedback following an increase in power. EOC reactivity coefficients are applied which maximizes the reactor power response. No operator action was credited in the representative case. Normal control system such as PZR spray, heater, letdown controls and automatic rod control are modeled based on the control status shown in Table 7-24. 8.1.3.2 Analysis Results - MCHFR Case As discussed in Section 7.2.4.1, in an NPM design the smaller breaks can result in a delayed detection time compared to larger breaks and be more challenging for MCHFR. The MCHFR case is identified as a small (3.3 percent of the pipe cross section area) split break in the MS piping just outside of containment. The following describes the event sequence of the MCHFR case for the steam line break event. Table 8-3 summarizes the sequence of events. Figure 8-20 through Figure 8-28 show some key parameters during the event. The main steam line break event begins at time zero. Steam flow increases following the initiation of the break, as seen in Figure 8-20. As heat removal from the RCS through the SG increases above its steady state value, RCS temperature begins to decrease (Figure 8-21). The drop in average RCS temperature prompts the CR controller to pull the regulating bank out of the core. The addition of positive reactivity causes reactor power to increase (Figure 8-22). At ~47 seconds, the high power analytical limit (200 MWth or 125 percent RTP) is reached (Figure 8-23). At ~49 seconds, power is peaked at 202.8 MWth when reactor starts to scram. Approximately at the same time, the limiting MCHFR calculated by NRELAP5 is reached. SG pressure decreases during the initial phase of the transient due to the break. As reactor power increases the SG pressure starts to increase and reaches the analytical limit of 800 psia at ~59 seconds (Figure 8-24). This is the time when DHRS is actuated. DHRS actuation closes the FWIVs and MSIVs, which isolate the SGs and the break from the remaining secondary system and ends the overcooling transient. Steam generator pressure increase resulting from MS isolation is expected. SG pressure starts to decrease once the DHRS cooling is established. Because of the single failure of the MSIV on the affected SG (failure to close), the affected SG and the associated DHRS train are depleted through © Copyright 2022 by NuScale Power, LLC 580

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 the break after DHRS actuation. The unaffected SG and the associated DHRS train remove the decay heat (Figure 8-20 and Figure 8-24). RCS level has an overall decreasing trend as inventory shrinks due to increased heat removal due to steam line break (Figure 8-25). The PZR level is still above the low PZR level analytical limit (35 percent) by the end of the transient so the PZR heaters can still function. The RCS pressure initially drops as inventory shrinks due to increased heat removal. As reactor power increases and as the PZR heaters respond, an increase in RPV pressure to ~1997 psia is observed. When the reactor is tripped, the RCS pressure has a sudden drop and then gradually returns to its setpoint, using pressurizer heaters (Figure 8-26). After reactor trip and actuation of DHRS, RCS flow decreases rapidly and becomes stagnant and slightly reversed at ~85 seconds (Figure 8-27). Following that, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2 and shown in Figure 8-20, Figure 8-21, Figure 8-27, and Figure 8-28); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. By 30 minutes, RCS flow has stabilized, and the RCS temperature and pressure are steadily decreasing as the DHRS transfers decay heat from the RCS to the reactor pool (Figure 8-21, Figure 8-26, Figure 8-27, and Figure 8-28). It is concluded that by 30 minutes the transient has been terminated and that stable DHRS cooling has been achieved. Subcritical margin is verified as net reactivity remains less than 0.0 dollars at the time stable DHRS cooling has been achieved (Figure 8-22). No operator action was credited to mitigate this event. Table 8-3 Main steam line break sequence of events Event Time (sec) Initiation of a split main steam line break (3.3% of the pipe cross sectional area) on 0 SG 2 High power (125% RTP) analytical limit is reached. 47 RTS is actuated and control rods begin to insert into the core. 49 Peak power is 202.8 MW. 49 MCHFR is reached (3.682 as calculated by NRELAP5) 49 Control rods fully inserted. 51 Peak RCS pressure is reached (~1997 psia). 52 High SG pressure analytical limit (800 psia) is reached. DHRS is actuated. (1) 59 RCS flow is stagnant and slightly reversed. ~85 Peak main steam system pressure (~1207 psia) is reached. 133 End of calculation. Stable DHRS cooling has been established. Net reactivity 1800 remains < $0.0. (1) In this example transient calculation the pressurizer heater trip on high SG pressure was not modeled. The calculation shows the heaters maintain the RCS pressure. The overall event progression is not affected because reactor trip and DHRS actuation have occurred. © Copyright 2022 by NuScale Power, LLC 581

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-20 Steam generators 1 (unaffected) and 2 (affected) secondary flow rates for the representative main steam line break event Figure 8-21 Core inlet temperature for the representative main steam line break event © Copyright 2022 by NuScale Power, LLC 582

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-22 Net reactivity for the representative main steam line break event Figure 8-23 Power response for the representative main steam line break event © Copyright 2022 by NuScale Power, LLC 583

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-24 Steam generators 1 (unaffected) and 2 (affected) pressure response for the representative main steam line break event Figure 8-25 Pressurizer level for the representative main steam line break event © Copyright 2022 by NuScale Power, LLC 584

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-26 Reactor pressure vessel pressure response for the representative main steam line break event Figure 8-27 Reactor coolant system flow rate for the representative main steam line break event © Copyright 2022 by NuScale Power, LLC 585

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-28 Core outlet temperature for the representative main steam line break event 8.1.3.3 Conclusion A representative case that could challenge MCHFR was identified for a main steam line break event. The results of this case, as presented in Section 8.1.3.2, are subsequently used as input to an MCHFR evaluation using the NuScale subchannel analysis methodology. 8.2 Heatup and/or Pressurization of the Reactor Coolant System 8.2.1 Loss of Normal Feedwater Flow The purpose of this section is to present the thermal-hydraulic response of a representative NPM for loss of normal feedwater flow. This event is evaluated for primary pressure and secondary pressure. Different initial condition biases and conservatisms are used for the RCS pressure case and the secondary pressure case. 8.2.1.1 Event Description - Reactor Coolant System Pressure Case The general loss of normal feedwater flow event description can be found in Section 7.2.10. Chosen from a series of RCS pressure sensitivity cases, the sample loss of normal feedwater flow case here represents a case that could challenge the RCS pressure acceptance criterion. No single failure is applied since the challenging cases occur when all equipment operates as designed. Normal AC power is lost at turbine trip since this maximizes the system pressure responses. This case features the following conditions: © Copyright 2022 by NuScale Power, LLC 586

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The initial power is 102 percent. RCS average temperature is biased at the high condition (555 degrees F). RCS flow rated is biased to the low condition (535 kg/s). Pressurizer pressure is biased to the low condition (1780 psia). Pressurizer level is biased to the high condition (58 percent). Initial feedwater temperature is biased to the high condition (312.5 degrees F). Initial SG pressure is biased to the high condition (535 psia). SG heat transfer is increased 30 percent by applying a heat transfer coefficient multiplier of 1.3 on both the primary and secondary sides of the steam generator tubes in the steady state initialization model. As identified in Section 7.2.10.3, this biasing does not significantly affect margin to the RCS pressure acceptance criteria. There is no SG tube plugging for this representative calculation. BOC reactivity coefficients are used since they are bounding for overheating events. No operator action was credited in the representative case. Normal control system such as PZR spray, heater, letdown controls and automatic rod control are modeled based on the control status shown in Table 7-48. 8.2.1.2 Analysis Results - RCS Pressure Case The following describes the event sequence of a representative case for the loss of normal feedwater flow event that could challenge the primary pressure. Table 8-4 summarizes the sequence of events. Figure 8-29 through Figure 8-37 show some key parameters during the representative loss of normal feedwater flow event. The feedwater flow is completely lost at time zero of the transient. Due to the overheating after loss of feedwater, pressurizer pressure and level start to increase (Figure 8-29 and Figure 8-30). At 17.6 seconds after transient initiation, the pressurizer pressure reaches the reactor trip analytical limit (Figure 8-29). At 18.6 seconds, the turbine trips (on the reactor trip signal), normal power is lost-causing MSIV closure-and steam generator pressures begin to increase (Figure 8-31). At 19.6 seconds, scram rod insertion begins, which causes the reactor power to decrease (Figure 8-32), resulting in a RCS flow decrease (Figure 8-33). After reaching high pressurizer pressure analytical limit, the DHRS is also actuated at 19.6 seconds, but conservative modeling of the DHRS actuation valve opening delays initiation of flow in the DHRS until 49.3 seconds. At 20.0 seconds, minimum MCHFR is reached (4.37 as calculated by NRELAP5). At 24.2 seconds, the pressurizer dome pressure reaches the 2137.3 psia Reactor Safety Valve 1 biased setpoint, Reactor Safety Valve 1 begins to stroke open. At 24.6 seconds, the pressure at the bottom of the reactor vessel reaches a maximum value of 2156.1 psia and begins to decrease (Figure 8-29). That maximum RCS pressure value is well below 110 percent of the RCS design pressure (2310 psia). At 33.1 seconds, Reactor Safety Valve 1 closes. For the © Copyright 2022 by NuScale Power, LLC 587

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 remainder of the transient, pressurizer pressure and level continue to decrease (Figure 8-29 and Figure 8-30). At 49.3 seconds, flows through the DHRS actuation valves and into the steam generator inlet plena begin. The steam generator pressure starts to increases. At

                  ~322 seconds, steam generator pressure reaches the maximum of 1243.4 psia and begins to decrease (Figure 8-31).

After reactor trip and actuation of DHRS, RCS flow decreases rapidly and becomes almost stagnant at ~50 seconds (Figure 8-33). Following that, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2). The oscillations gradually diminish as the stable natural circulation flow is established in the RCS (Figure 8-31, Figure 8-33, and Figure 8-34). By 2500 seconds, DHRS operation and RCS flow are stable (Figure 8-33, Figure 8-34, and Figure 8-37); RCS pressure, temperatures, and steam generator pressure are trending downward (Figure 8-29, Figure 8-30, Figure 8-34 and Figure 8-35); the subcritical margin remains large at the end of the transient calculation (Figure 8-36). Table 8-4 Loss of normal feedwater flow sequence of events - reactor coolant system pressure case Event Time (sec) Total loss of feedwater 0.0 s Pressurizer pressure analytical limit (2000 psia) is reached. 17.6 s Turbine trips, normal AC power is assumed to be lost 18.6 s Scram rod insertion begins, and DHRS is actuated 19.6 s MCHFR is reached (4.37 as calculated by NRELAP5) 20.0 s Peak RCS pressure is reached (2156.1 psia) 24.6 s RCS flow is stagnant ~50 s SG pressures reach maximum (1243.4 psia) and begin to decrease 322.2 s End of calculation. Stable DHRS cooling has been established. Net reactivity 2500.0 s remains < $0.0. © Copyright 2022 by NuScale Power, LLC 588

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-29 Reactor pressure vessel pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure 8-30 Pressurizer level for the representative loss of normal feedwater flow event - reactor coolant system pressure case © Copyright 2022 by NuScale Power, LLC 589

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-31 Steam generator 2 pressure response for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure 8-32 Power response for the representative loss of normal feedwater flow event - reactor coolant system pressure case © Copyright 2022 by NuScale Power, LLC 590

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-33 Reactor coolant system flow rate for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure 8-34 Core inlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case © Copyright 2022 by NuScale Power, LLC 591

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-35 Core outlet temperature for the representative loss of normal feedwater flow event - reactor coolant system pressure case Figure 8-36 Net reactivity for the representative loss of normal feedwater flow event - reactor coolant system pressure case © Copyright 2022 by NuScale Power, LLC 592

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-37 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - reactor coolant system pressure case 8.2.1.3 Conclusion - RCS Pressure Case A representative case that could challenge the RCS pressure acceptance criterion was identified. The results of this case, as presented in Section 8.2.1.2, demonstrate the RCS pressure acceptance criterion is met. 8.2.1.4 Event Description - Secondary Pressure Case The general loss of normal feedwater flow event description can be found in Section 7.2.10. Chosen from a series of secondary pressure sensitivity cases, the sample loss of normal feedwater flow case here represents a case that could challenge the secondary pressure acceptance criterion. No single failure is applied since the challenging cases occur when all equipment operates as designed. Normal AC power is lost at turbine trip since this maximizes the system pressure responses. This case features the same conditions as for the RCS pressure case (shown in Section 8.2.1.1), except that the feedwater flow is not totally lost at the beginning of the event. Instead, only a partial feedwater flow is lost (2.3 percent). 8.2.1.5 Analysis Results - Secondary Pressure Case The following describes the event sequence of a representative case for the loss of normal feedwater flow event that could challenge the secondary pressure. Table 8-5 summarizes the sequence of events. Figure 8-38 through Figure 8-46 © Copyright 2022 by NuScale Power, LLC 593

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 show some key parameters during the representative loss of normal feedwater flow event. The transient for this case is initiated by a fault that is postulated to result in a partial (2.3 percent) loss of feedwater flow. Automatic rod control is conservatively disabled and therefore control rods are not modeled to insert to reduce power. The RCS heats up slowly (Figure 8-38 and Figure 8-39), and the RCS pressure and pressurizer level start to increase (Figure 8-40 and Figure 8-41). At 638.3 seconds after transient initiation, the RCS riser leg temperature reaches the reactor trip analytical limit. At 645.3 seconds, it is assumed that the turbine trips (on the reactor trip signal) and normal power AC power is lost. The loss of AC power causes MSIV closure, and the steam generator pressures and RCS pressure begin to increase (Figure 8-40 and Figure 8-42). At 646.3 seconds, scram rod insertion begins, which causes the reactor power to decrease (Figure 8-43), resulting in a RCS flow decrease (Figure 8-44). After reaching the high RCS riser leg temperature analytical limit, the DHRS is also actuated at 646.3 seconds, but conservative modeling of the DHRS actuation valve opening delays initiation of DHRS flow until 676.0 seconds. At 658.9 seconds, the RCS pressure reaches a maximum (1939.4 psia) and begins to decrease (Figure 8-40); the maximum value in this case is less than that for the RCS pressure case and the RSVs do not open. For the remainder of the transient, the RCS pressure and pressurizer level continue to decrease (Figure 8-40 and Figure 8-41). At 676.0 seconds, flow through the DHRS actuation valves and into the steam generator inlet plena begins (Figure 8-46). The steam generator pressure starts to increase. At ~716.3 seconds, steam generator pressure reaches the maximum of 1421.6 psia and begins to decrease (Figure 8-42). After reactor trip and actuation of DHRS, RCS flow decreases rapidly and briefly becomes almost stagnant at ~700 seconds (Figure 8-44). Following that, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2). The oscillations gradually diminish as the stable natural circulation flow is established in the RCS (Figure 8-38 and Figure 8-42 through Figure 8-44). By 2500 seconds, DHRS operation and RCS flow are stable (Figure 8-38 through Figure 8-44); RCS pressure, temperatures, and steam generator pressure are trending downward (Figure 8-38, Figure 8-39, Figure 8-40, and Figure 8-42); the subcritical margin remains large (Figure 8-45). © Copyright 2022 by NuScale Power, LLC 594

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 8-5 Loss of normal feedwater flow sequence of events - secondary pressure case Event Time (sec) Feedwater flow begins 0.1 s rampdown to 97.7% of initial value 0.0 s RCS riser leg temperature analytical limit (610°F) is reached 638.3 s Turbine trips, normal AC power is assumed to be lost 645.3 s Scram rod insertion begins, and DHRS is actuated 646.3 s Peak RCS pressure is reached (1939.4 psia) 658.9 s RCS flow briefly becomes almost stagnant ~700 s SG pressures reach maximum (1421.6 psia) and begin to decrease 716.3 s End of calculation. Stable DHRS cooling has been established. Net reactivity 2500.0 s remains < $0.0. Figure 8-38 Core inlet temperature for the representative loss of normal feedwater flow event - secondary pressure case © Copyright 2022 by NuScale Power, LLC 595

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-39 Core outlet temperature for the representative loss of normal feedwater flow event - secondary pressure case Figure 8-40 Reactor pressure vessel pressure response for the representative loss of normal feedwater flow event - secondary pressure case © Copyright 2022 by NuScale Power, LLC 596

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-41 Pressurizer level for the representative loss of normal feedwater flow event - secondary pressure case Figure 8-42 Steam generator 2 pressure response for the representative loss of normal feedwater flow event - secondary pressure case © Copyright 2022 by NuScale Power, LLC 597

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-43 Power response for the representative loss of normal feedwater flow event - secondary pressure cas Figure 8-44 Reactor coolant system flow rate for the representative loss of normal feedwater flow event - secondary pressure case © Copyright 2022 by NuScale Power, LLC 598

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-45 Net reactivity for the representative loss of normal feedwater flow event - secondary pressure case Figure 8-46 Steam generator 2 secondary flow for the representative loss of normal feedwater flow event - secondary pressure case © Copyright 2022 by NuScale Power, LLC 599

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.2.1.6 Conclusion - Secondary Pressure Case A representative case that could challenge secondary pressure acceptance criteria was identified. The results of this case, as presented in Section 8.2.1.5 demonstrate the secondary pressure acceptance criterion is met. 8.2.2 Loss of Normal AC Power The purpose of this section is to present the thermal-hydraulic response of a representative NPM for loss of normal AC power. This event is evaluated for primary pressure and secondary pressure. A representative case evaluated for primary pressure is presented. 8.2.2.1 Event Description The general event description for a loss of normal AC power is found in Section 7.2.9.1. Based on Section 7.2.9.1, the acceptance criterion for primary pressure is potentially challenged by the loss of normal AC power. The challenging case occurs when all equipment is operational. The representative case presented here represents a case that could challenge the primary pressure response (as shown in Table 7-44). Case features include the following conditions: Conservative initial condition biasing is applied in order to maximize the consequences of this event. This representative case is initialized at 102 percent reactor power. The RSV set pressure is biased to the high condition (2137.25 psia, which includes a 3 percent drift allowance) to maximize primary pressure response. The reactor pool temperature is biased to the maximum bounding value (200 degrees F) to provide a bounding high temperature for events/cases that require heat removal using the DHRS. The initial pressurizer pressure is biased to the maximum value (1920 psia) for the primary side pressurization case. The initial pressurizer level is biased to the maximum value (58 percent) to maximize the primary side pressure response. BOC reactivity coefficients are applied to maximize the primary side pressurization. 8.2.2.2 Analysis Results The following describes the event sequence of the loss of normal AC power event. Table 8-6 summarizes the sequence of events. Figure 8-47 through Figure 8-53 show some key parameters during the loss of normal AC power event. © Copyright 2022 by NuScale Power, LLC 600

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The loss of normal AC power occurs at zero seconds, which causes an immediate turbine trip. In this case, it is assumed that the DC power system (EDSS for the representative NPM) batteries are available. The control rods do not immediately insert on loss of AC power. Rather, a delay to reactor trip, DHRS actuation, and containment isolation are assumed to maximize RCS pressure. During this delay period, primary and secondary pressure increase due to the turbine trip. Primary temperatures (Figure 8-47) and system pressures (Figure 8-48, curves for SG1 and SG2 overlap) initially increase due to the mismatch between primary side heat production and secondary side heat sink. The analytical limit for high pressurizer pressure is reached at approximately 6 seconds, actuating reactor trip (Figure 8-49). DHRS valves begin to open at approximately 8 seconds (Figure 8-50). At approximately 12 seconds one of the two RSVs lifts (Figure 8-51). A peak pressure of 2155 psia is reached in the RPV. Subsequent to reactor trip and RSV lift, system pressures and temperatures decrease, and system shrinkage reduces the PZR level (Figure 8-53). After reactor trip and actuation of DHRS, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. Figure 8-47, Figure 8-50, and Figure 8-52 show that at ~15 minutes, RCS flow has stabilized, and the RCS temperatures are steadily decreasing as the DHRS is transferring decay heat from the RPV to the reactor pool. It is concluded that by 30 minutes the transient has been terminated and that stable DHRS cooling has been achieved. No operator action was credited to mitigate this event. Table 8-6 Sequence of events for loss of AC power Event Time (sec) Loss of AC power occurs 0.0 Turbine trip occurs 0.0 CVCS isolation occurs 0.0 High PZR pressure analytical limit is reached (2000 psia). 6 RTS actuation on high pressurizer pressure signal. 8 DHRS actuation on the high pressurizer pressure signal. DHRS actuation valves 8 begin to open. FWIVs and MSIVs begin to close. RSV1 opens 12 Peak RPV pressure is reached (2155 psia) 12 MSIVs are fully closed. 13 FWIVs are fully closed. 13 DHRS actuation valves are fully open. 38 Peak steam generator pressure is reached (1250 psia). 80 Establishment of stable RCS flow. Pressure and temperature are steadily 1500 decreasing. End of calculation. Stable DHRS cooling has been established. 1800 © Copyright 2022 by NuScale Power, LLC 601

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-47 Primary temperature response for the representative loss of AC power event Figure 8-48 System pressure response for the representative loss of AC power event © Copyright 2022 by NuScale Power, LLC 602

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-49 Reactor pressure vessel core power response for the representative loss of AC power event Figure 8-50 Decay heat removal system response for the representative loss of AC power event © Copyright 2022 by NuScale Power, LLC 603

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-51 RSV flow response for the representative loss of AC power event Figure 8-52 Reactor coolant system flow response for the representative loss of AC power event © Copyright 2022 by NuScale Power, LLC 604

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-53 Pressurizer level response for the representative loss of AC power event 8.2.2.3 Conclusion A representative case that could challenge primary pressure acceptance criterion was identified. The results of this case, as presented in Section 8.2.2.2, demonstrate that the primary pressure acceptance criterion is met. 8.2.3 Feedwater Line Break The purpose of this section is to present the thermal-hydraulic response of a representative NPM to a feedwater system piping failure. The challenging case occurs when AC power is lost at event initiation and all equipment is operational. This event is evaluated for primary pressure and secondary pressure. A representative case evaluated for primary pressure is presented. 8.2.3.1 Event Description The general event description for a feedwater system piping failure is found in Section 7.2.12.1. Based on Section 7.2.12.1, the acceptance criterion for primary pressure is potentially challenged by the feedwater system piping failure. © Copyright 2022 by NuScale Power, LLC 605

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The representative case presented here represents a case that could challenge the primary pressure response (as shown in Table 7-56). Case features include the following conditions: Conservative initial condition biasing is applied in order to maximize the consequences of this event. This representative case is initialized at 102 percent reactor power. The initial primary temperatures are biased to the maximum value (~545 degrees F) to produce a (slightly) higher energy system. The RSV set pressure is biased to the high condition (2137.25 psia, which includes a 3 percent drift allowance) to maximize primary pressure response. The initial feedwater temperature is biased to the maximum value (~305 degrees F) to produce a (slightly) higher energy system. The reactor pool temperature is biased to the maximum bounding value (~200 degrees F) to provide a bounding high temperature for events/cases that require heat removal using the DHRS. The initial pressurizer pressure is biased to the maximum value (1920 psia) for the primary side pressurization case. The initial pressurizer level is biased to the maximum value (58 percent) to maximize the primary side pressure response. BOC reactivity coefficients are applied to maximize the primary side pressurization. Automatic rod control is disabled. 8.2.3.2 Analysis Results The following describes the event sequence of the feedwater line break event. Table 8-7 summarizes the sequence of events. Figure 8-54 through Figure 8-61 show some key parameters during the feedwater line break event. A feedwater line break outside of containment occurs at zero seconds coincident with a loss of AC power, causing an immediate turbine trip and feedwater pump trip (Figure 8-54). In this case, it is assumed that the DC power system (EDSS for the representative NPM) batteries are available. To maximize RCS pressure, in this case it is assumed that after normal AC power is lost and MPS senses the loss of power to the EDSS battery chargers, there is a delay before MPS actuates reactor trip, DHRS and containment isolation. Since the reactor does not immediately trip, primary temperatures (Figure 8-55, curves for inlet temperature and RCS temperature overlap) and system pressures (Figure 8-56, curves for SG1 and SG2 overlap) initially increase due to the mismatch between primary side heat production and secondary side heat sink. The analytical limit for PZR pressure is reached at approximately 6 seconds. RTS actuation (Figure 8-57) occurs at approximately 8 seconds and RSV lift (Figure 8-58) occurs at approximately 11 seconds. A peak pressure of 2158 psia is reached in the RPV. © Copyright 2022 by NuScale Power, LLC 606

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The DHRS valves are fully open at approximately 36 seconds (Figure 8-59, curves for DHRS 1 and DHRS 2 overlap). Subsequent to reactor trip and RSV lift, system pressures and temperatures decrease, and system shrinkage reduces the PZR level (Figure 8-60). After reactor trip and actuation of DHRS, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. Figure 8-55, Figure 8-59, and Figure 8-61 show that at ~30 minutes, RCS flow has stabilized, and the RCS temperatures are steadily decreasing as the DHRS is transferring decay heat from the RPV to the reactor pool. It is concluded that by ~40 minutes the transient has been terminated and that stable DHRS cooling has been achieved. In this case, the EDSS batteries hold the ECCS valves closed for the duration of the transient calculation. No operator action was credited to mitigate this event. Table 8-7 Sequence of events for feedwater line break outside containment Event Time (sec) A double ended guillotine break in the SG 2 feedwater line occurs under the 0 bioshield (just outside of containment). AC power is lost resulting in turbine trip and FW pump trip 0 High PZR pressure analytical limit is reached (2000 psia) 6 RTS and DHRS ESFAS actuated 8 MSIVs close 8 Control rods fully inserted 10 RSV lift point is reached (2137 psia) 11 Peak RCS pressure reached (2158 psia) 12 FWIVs close (check valves already seated) 15 RSV reseats 22 DHRS actuation valves open 36 Peak pressure reached in SG 1 (1299 psia) 77 End of calculation. Stable DHRS cooling has been established. 3600 © Copyright 2022 by NuScale Power, LLC 607

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-54 Feedwater line break flow response for the representative feedwater line break event © Copyright 2022 by NuScale Power, LLC 608

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-55 Primary temperature response for the representative feedwater line break event Figure 8-56 System pressure response for the representative feedwater line break event © Copyright 2022 by NuScale Power, LLC 609

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-57 Reactor pressure vessel core power response for the representative feedwater line break event Figure 8-58 Reactor safety valve flow response for the representative feedwater line break event © Copyright 2022 by NuScale Power, LLC 610

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-59 Decay heat removal system response for the representative feedwater line break event Figure 8-60 Pressurizer level response for the representative feedwater line break event © Copyright 2022 by NuScale Power, LLC 611

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-61 Reactor coolant system flow response for the representative feedwater line break event 8.2.3.3 Conclusion A representative case that could challenge the RCS pressure acceptance criteria was identified. The results of this case, as presented in Section 8.2.3.2, demonstrate the RCS pressure acceptance criterion is met. 8.3 Reactivity Anomaly 8.3.1 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions The purpose of this section is to present the thermal-hydraulic and core neutronic responses of a representative NPM for an uncontrolled control rod assembly bank withdrawal from subcritical or low power startup conditions event. This event is evaluated for both MCHFR and maximum fuel centerline temperature. 8.3.1.1 Event Description The event description for a bank withdrawal from low power startup conditions can be found in Section 7.2.13. Based on Section 7.2.13, the acceptance criteria for MCHFR and maximum fuel centerline temperature are potentially challenged. For the representative NPM, the results of the sensitivity studies indicate the event scenario with the lowest MCHFR may differ from the event scenario with the © Copyright 2022 by NuScale Power, LLC 612

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 highest fuel centerline temperature; but a single active failure cannot make the event consequences worse. The representative case presented here corresponds to one of the more limiting peak power cases, and therefore a case that may challenge MCHFR. This case features the following conditions for a bank withdrawal from low power startup conditions: The initial core power is 15 percent RTP, which is the maximum power level for this event classification. The high count rate signal is inactive. The RCS core inlet temperature is the minimum allowed for criticality (420 degrees F). Heat removal is through the steam generators with coolant provided by the operating feedwater pump. The reactivity insertion rate is 35 pcm/s. No loss of normal AC power. No operator action was credited in the representative case. 8.3.1.2 Analysis Results The following describes the event sequence of the withdrawal of a control rod assembly bank from a low power startup condition. Table 8-8 summarizes the sequence of events. Figure 8-62 through Figure 8-69 show some key parameters for this case. The bank withdrawal occurs at time zero. The addition of positive reactivity (Figure 8-69) causes reactor power to increase (Figure 8-64). Since the high power rate signal is not active (below 15 percent RTP), core protection is provided by the high power signal and the startup rate (intermediate range) signal. In this instance, the analytical limit for the startup rate (intermediate range) is reached (3 DPM (decades per minute)) at 4 seconds. The control rods are free to fall after two seconds (Figure 8-69), while the DWS isolation valves begin to close. This latter action occurs as a precaution to the reactivity addition being caused by a dilution of the reactor coolant boron concentration. The peak core power (42.0 percent RTP) occurs at 7 seconds. The quick action of the MPS to trip the reactor precludes the addition of significant energy into the reactor coolant. For instance, the pressurizer pressure increases by less than 10 psia (Figure 8-62), while the increase in SG pressures is comparable (Figure 8-63). The delay between energy production in the fuel and heat addition to the reactor coolant causes the core outlet temperature (Figure 8-67) and RCS flow rate (Figure 8-68) to remain relatively constant until increasing shortly before reactor trip. Both parameters peak at ~10 seconds before decreasing rapidly as the core heat flux reduces. The core inlet temperature (Figure 8-65) and density (Figure 8-66) remain relatively constant © Copyright 2022 by NuScale Power, LLC 613

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 because the transient is terminated prior to completion of one loop transit. The MCHFR predicted by NRELAP5 (greater than 25) occurs at 7 seconds. At high initial power levels, the calculation would normally be continued to verify that the module transitions into passive and stable DHRS cooling. However, the system responses at this operating condition are negligible compared to the heat removal capability of the plant (normal feedwater, containment flooding, etc.). Since there is little additional energy to remove, core cooling is ensured and the evaluation can be quickly terminated. As noted above, the RCS flow peaked and is starting to decrease (Figure 8-68), while the core outlet temperature (Figure 8-67) and pressurizer pressure (Figure 8-62) are steady or decreasing. No operator action was credited to mitigate this event. Table 8-8 Withdrawal of a control rod assembly bank from a low power startup condition sequence of events Event Time (sec) Malfunction that initiates the withdrawal of a CRA bank 0 High startup rate (intermediate range) analytical limit (3 DPM) reached 4 RTS actuation (control rods are free to fall) 6 DWS isolation (valves begin to close) 6 Maximum core power occurs (42.0% RTP) 7 Lowest MCHFR occurs (> 25 as calculated by NRELAP5) 7 Maximum RCS pressure occurs (1873 psia) 9 DWS isolation valves closed 11 Transient terminated 20 © Copyright 2022 by NuScale Power, LLC 614

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-62 Pressurizer pressure response for the bank withdrawal from a low power startup condition Figure 8-63 Reactor pressure vessel and steam generator pressure responses for the bank withdrawal from a low power startup condition © Copyright 2022 by NuScale Power, LLC 615

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-64 Power response for the bank withdrawal from a low power startup condition Figure 8-65 Core inlet temperature for the bank withdrawal from a low power startup condition © Copyright 2022 by NuScale Power, LLC 616

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-66 Core inlet density for the bank withdrawal from a low power startup condition Figure 8-67 Core outlet temperature for the bank withdrawal from a low power startup condition © Copyright 2022 by NuScale Power, LLC 617

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-68 Reactor coolant system flow rate for the bank withdrawal from a low power startup condition Figure 8-69 Net reactivity for the bank withdrawal from a low power startup condition © Copyright 2022 by NuScale Power, LLC 618

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.3.1.3 Conclusion A representative challenging case regarding MCHFR was identified for a bank withdrawal from a low power startup condition. The results of this case, as presented in Section 8.3.1.2, are subsequently used as input to an MCHFR evaluation using the NuScale subchannel analysis methodology. 8.3.2 Control Rod Misoperation The purpose of this section is to present the thermal-hydraulic and core neutronic responses of a representative NPM for control rod misoperation events. Based on Section 7.2.15.1, the control rod misoperation analysis consists of two major transients: single rod withdrawal and rod drop. A representative single rod withdrawal transient is presented in this section. For the single rod withdrawal transient, MCHFR is the acceptance criterion that may be potentially challenged during the transient. 8.3.2.1 Single Rod Withdrawal MCHFR Case - Event Description The general description for the control rod misoperation event can be found from Section 7.2.15.1. Chosen from a series of MCHFR sensitivity cases, the representative single rod withdrawal case presented here represents a case that could challenge MCHFR, based on the NRELAP5 MCHFR pre-screening. No single failure is applied since the challenging cases occur when all equipment operates as designed. No loss of power is applied since loss of power scenarios trips the reactor and does not make the MCHFR worse. This case features the following conditions: The initial power is 75 percent. Conservative initial condition biasing (as shown in Table 7-68) is applied in order to maximize the consequences of the single rod withdrawal event in terms of MCHFR. RCS flow rate is biased at low conditions for MCHFR. Average RCS temperature is biased at low condition (535 degrees F) for maximum delay to the high RCS riser temperature trip. To maximize the RCS pressure at time of trip without causing an earlier trip on pressurizer pressure, letdown is active, the pressurizer heater is off, and the spray flow is set to 97.6 percent of the CVCS recirculation flow. Additionally, the pressurizer pressure and level were given biases of -70 psi and -3 percent, respectively. The reactivity insertion rate is 2.3 pcm/s, which is the maximum that does not result in an earlier trip on high power rate. Automatic rod control is disabled since it will counteract the reactivity insertion due to the rod withdrawal at the beginning of the event. BOC reactivity coefficients are applied, which is conservative for overpower events. © Copyright 2022 by NuScale Power, LLC 619

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.3.2.2 Single Rod Withdrawal MCHFR Case - Analysis Results The following describes the event sequence of the single rod withdrawal MCHFR case. Table 8-9 summarizes the sequence of events. Figure 8-70 through Figure 8-78 show some key parameters for the single rod withdrawal MCHFR case. The single rod withdrawal event begins at time zero. The addition of positive reactivity causes reactor power, pressurizer pressure and level to increase (Figure 8-70, Figure 8-71, and Figure 8-72). To bound the limiting single failure of an ex-core flux detector, the lowest reading ex-core detector is used to determine if the high power trip analytical limit of 120 percent is reached. Therefore the actual reactor power reaches a maximum of ~ 198 MWth, slightly exceeding the 120 percent limit, at ~ 147 s; this power level is not sufficient for the lowest reading ex-core detector to indicate that the 120 percent analytical limit is reached. As the power increases, the RCS riser temperature increases; in this case, the high RCS riser temperature limit of 610 degrees F is reached at ~139 seconds (Figure 8-73). There is a total 8 second delay between the high RCS riser temperature signal and RTS/DHRS actuation. RTS and DHRS are actuated at ~147 seconds. The limiting MCHFR is also reached at ~147 seconds. DHRS actuation closes the FWIVs and MSIVs. Steam generator pressure increases as a result of the MS isolation. SG pressure starts to decrease once the DHRS cooling is established (Figure 8-74, SG2 shown but SG1 response is identical). Pressurizer pressure reaches its peak value, less than 2100 psia, around 157 seconds and starts to decrease. Pressurizer level has a similar transient response (Figure 8-71 and Figure 8-72). After reactor trip and actuation of DHRS, RCS flow decreases rapidly and becomes stagnant and slightly reversed at ~184 seconds (Figure 8-75). Following that, flow oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2 and shown by Figure 8-73, Figure 8-75, and Figure 8-77). By 40 minutes, RCS flow and DHRS flow have stabilized (Figure 8-75 and Figure 8-76), and RCS temperature and pressure are steadily decreasing as the DHRS transfers decay heat from the RCS to the reactor pool (Figure 8-71, Figure 8-73, and Figure 8-77). It is concluded that by 40 minutes the transient has been terminated and that stable DHRS cooling has been achieved. Subcritical margin is verified as net reactivity remains less than 0.0 dollars at the time stable DHRS cooling has been achieved (Figure 8-78). No operator action was credited to mitigate this event. Table 8-9 Single rod withdrawal sequence of events - MCHFR case Event Time (sec) Malfunction that initiates the withdrawal of a single CRA. 0 High RCS riser temperature analytical limit (610°F) is reached 139 High pressurizer pressure analytical limit (2000 psia) is reached 145 © Copyright 2022 by NuScale Power, LLC 620

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 8-9 Single rod withdrawal sequence of events - MCHFR case (Continued) Event Time (sec) RTS is actuated and control rod insertion begins 147 DHRS is actuated. The DHRS actuation valves begin opening. The FWIVs and 147 MSIVs begin closing. Lowest MCHFR is reached (3.107 as calculated by NRELAP5) 147 RCS flow is stagnant and slightly reversed. ~184 End of calculation. Stable DHRS cooling has been established. Net reactivity 2400 remains < $0.0. Figure 8-70 Power response for the representative single rod withdrawal MCHFR case © Copyright 2022 by NuScale Power, LLC 621

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-71 Reactor pressure vessel pressure response for the representative single rod withdrawal MCHFR case Figure 8-72 Pressurizer level for the representative single rod withdrawal MCHFR case © Copyright 2022 by NuScale Power, LLC 622

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-73 Core outlet temperature for the representative single rod withdrawal MCHFR case Figure 8-74 Steam generator 2 pressure response for the representative single rod withdrawal MCHFR case © Copyright 2022 by NuScale Power, LLC 623

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-75 Reactor coolant system flow rate for the representative single rod withdrawal MCHFR case Figure 8-76 Steam generator 2 secondary flow for the representative single rod withdrawal MCHFR case © Copyright 2022 by NuScale Power, LLC 624

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-77 Core inlet temperature for the representative single rod withdrawal MCHFR case Figure 8-78 Net reactivity for the representative single rod withdrawal MCHFR case © Copyright 2022 by NuScale Power, LLC 625

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.3.2.3 Single Rod Withdrawal MCHFR Case - Conclusion A representative case that could challenge MCHFR was identified for the single rod withdrawal event. The results of this case, as presented in Section 8.3.2.2, are subsequently used as input to an MCHFR evaluation using the NuScale subchannel analysis methodology. 8.4 Increase in Reactor Coolant System Inventory 8.4.1 Chemical and Volume Control System Malfunction that Increases Reactor Coolant System Inventory The purpose of this section is to present the thermal-hydraulic response of a representative NPM for a CVCS malfunction that increases the RCS inventory. This event is evaluated for primary and secondary pressure. A representative case evaluated for primary pressure is presented. 8.4.1.1 Event Description The general event description associated with a malfunction of the CVCS that increases RCS inventory is provided in Section 7.2.17. Based on Section 7.2.17, the RCS primary pressure is the acceptance criterion that may be challenged during the CVCS malfunction event. A representative case that could challenge the RCS primary pressure is presented here. This case features the following conditions: Conservative initial condition biasing is applied to maximize the consequences of the RCS inventory increase event. This case is initialized at 102 percent reactor power, low pressurizer pressure and high pressurizer level. Minimum RCS design flow is assumed. Low RCS average temperature is assumed. Net CVCS mass flow rate of 5.4 lbm/s (40 gpm) into the RPV (this includes CVCS recirculation flow rate into and out of the RPV) High CVCS makeup temperature of 150 degrees F Pressurizer spray and letdown assumed unavailable 8.4.1.2 Analysis Results The following describes the event sequence of the representative CVCS malfunction that increases RCS inventory event. Table 8-10 summarizes the sequence of events. Figure 8-79 through Figure 8-91 show some key parameters during the representative increase in RCS inventory event. For this event, constant makeup flow is injected into the RCS until the CVCS is isolated (Figure 8-79). The CVCS also recirculates flow (Figure 8-80); no letdown flow is modeled (Figure 8-81). Therefore, there is a net increase of flow into the RCS (Figure 8-82 and Figure 8-83). © Copyright 2022 by NuScale Power, LLC 626

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Upon initiation of makeup, pressurizer level and RCS pressure were observed to rise. Pressurizer pressure and level are shown in Figure 8-84 and Figure 8-85, respectively. RCS pressure reached the high pressurizer pressure analytical limit, actuating reactor trip (Figure 8-86) and DHRS (Figure 8-88 and Figure 8-89, curves for SG1 and SG2 overlap). In this case, the normal control system is assumed to trip the turbine on reactor trip, which reduced secondary side heat removal, increasing the rate of RCS pressurization. The inventory addition continued following reactor trip because RCS flow remained above 0 lbm/s (Figure 8-87) and therefore CVCS remained un-isolated. After reactor trip and actuation of DHRS, RCS flow oscillations were observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2). Oscillations can be seen in Figure 8-87, Figure 8-90 and Figure 8-91. After reactor trip, the pressurizer pressure and level began to decrease for a short period of time but the makeup flow continued and the pressurizer level and primary pressure eventually began to increase again. Once the RSV 1 lift pressure was reached, RCS pressure was reduced as vapor from the pressurizer was vented into the containment vessel. A peak pressure of 2155 psia was reached in the RPV. Inventory increase continued for approximately 100 seconds further before the high pressurizer level analytical limit was reached. After reaching the high pressurizer level of 80 percent, CVCS was isolated and the event ended. Pressurizer level never exceeds 80 percent, thus ensuring a steam bubble in the pressurizer throughout this transient. The calculation is continued to verify that the module transitions into passive and stable DHRS cooling. Figure 8-84, Figure 8-87, Figure 8-88, Figure 8-89, and Figure 8-90 show that at ~ 1 hour, RCS and DHRS flows have stabilized, primary and secondary pressures are decreasing, and the RCS temperatures are steadily decreasing as the DHRS is transferring decay heat from the RPV to the reactor pool. Shutdown margin is maintained at the end of the transient calculation (Figure 8-91). It is concluded that by ~ 1 hour the transient has been terminated and that stable DHRS cooling has been achieved. Table 8-10 Increase in reactor coolant system inventory sequence of events Event Time (sec) CVCS malfunction initiates an excess CVCS mass flow rate of 5.4 lbm/s into the 0.0 RPV Analytical limit for high pressurizer pressure (2000 psia) is reached 510.9 Turbine stop valve closed 511.8 RTS1 and DHRS actuation on high pressurizer pressure analytical limit 512.9 Primary MSIVs fully closed 518.8 RSV 1 actuates 3370.9 © Copyright 2022 by NuScale Power, LLC 627

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Table 8-10 Increase in reactor coolant system inventory sequence of events (Continued) Event Time (sec) Peak RCS pressure is reached (2155 psia) 3371.0 Analytical limit for high pressurizer level (80%) is reached 3465.9 CVCS isolation actuation on high pressurizer level analytical limit 3468.9 CVCS isolation valves closed on high pressurizer level analytical limit 3473.9 End of calculation. Stable DHRS cooling has been established. 4000

1. Control rods are assumed to be fully inserted 2.278 seconds following RTS actuation.

Figure 8-79 Makeup flow for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 628

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-80 Recirculation pump flow for increase in reactor coolant system inventory Figure 8-81 Letdown flow for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 629

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-82 CVCS recirculation flow rate into the reactor pressure vessel for increase in reactor coolant system inventory Figure 8-83 CVCS recirculation flow rate out of the reactor pressure vessel for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 630

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-84 Pressure at the bottom of the pressurizer for increase in reactor coolant system inventory Figure 8-85 Pressurizer level for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 631

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-86 Reactor power for increase in reactor coolant system inventory Figure 8-87 Reactor coolant system flow for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 632

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-88 Decay heat removal system flow rate for increase in reactor coolant system inventory Figure 8-89 Steam generator pressure for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 633

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-90 Core inlet and exit coolant liquid temperature for increase in reactor coolant system inventory Figure 8-91 Total reactivity for increase in reactor coolant system inventory © Copyright 2022 by NuScale Power, LLC 634

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 8.4.1.3 Conclusion A representative case that could challenge the RCS pressure acceptance criterion was identified. The results of this case, as presented in Section 8.4.1.2, demonstrate the RCS pressure acceptance criterion is met. 8.5 Decrease in Reactor Coolant System Inventory 8.5.1 Small Line Break Outside of Containment The purpose of this section is to present the thermal-hydraulic and core neutronic responses of a representative NPM for a postulated break in a small line carrying primary coolant. For the reasons discussed in Section 7.2.18, the lines evaluated are the CVCS charging lines, the CVCS letdown lines, and the CVCS pressurizer spray lines. This event is evaluated for offsite and onsite radiological dose consequences. The transient analysis results are input to downstream radiological consequences evaluations and downstream subchannel evaluations. This analysis is not intended to be used for 10 CFR 50 Appendix K compliance for a LOCA as the separate LOCA-EM topical report encompasses this subject. 8.5.1.1 Event Description The event description for a break in a small line carrying reactor coolant can be found in from Section 7.2.18. Based on Section 7.2.18, the acceptance criteria for radiological consequences are potentially challenged by the break in a small line carrying reactor coolant event; but a single active failure cannot make the event consequences worse. Radiological Consequences The representative case presented here corresponds to one of the more limiting integrated mass released cases, and therefore a case with higher dose consequences. This case features the following conditions for a DEG break of 100 percent area in the letdown line: Conservative initial condition biasing (as shown in Table 7-86) is applied in order to maximize the consequences of this event. This representative case is initialized at 102 percent reactor power. The initial RCS average temperature is biased to the maximum value (555 degrees F) to maximize the mass released through the break. The initial feedwater temperature is biased to the minimum value (~290 degrees F) in this case. The initial pressurizer pressure is biased to the maximum value (1920 psia) to delay actuation of the low pressurizer pressure RTS and ESFAS protection signals. © Copyright 2022 by NuScale Power, LLC 635

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 The initial pressurizer level is biased to the maximum value (68 percent) to delay actuation of the low pressurizer level RTS and ESFAS protection signals. EOC reactivity coefficients are applied to obtain the mass released through the break associated with a specific time in core life. A loss of normal AC Power at event initiation is utilized, consistent with Table 7-87, to maximize the mass released through the break. No operator action was credited in the representative case. 8.5.1.2 Analysis Results The following describes the event sequence of the representative small break outside CNV event. Table 8-11 summarizes the sequence of events. Figure 8-92 through Figure 8-102 show some key parameters during the representative small break outside CNV event. A letdown line break of 100 percent area occurs at time zero with a coincident loss of AC power. Critical flow conditions are quickly established at the break location causing the break flow rate to stabilize (Figure 8-92). The reactor coolant lost through the break causes an immediate decrease in pressurizer level (Figure 8-93) and RPV pressure (Figure 8-94). In contrast, the SG pressure immediately increases (Figure 8-95) as a result of the turbine trip induced by the loss of AC power. The reduced secondary heat sink associated with tripping the feedwater pumps on loss of AC power is sufficient to cause the pressurizer level and RPV pressure to begin to increase. The exit pressures of the SGs reach the analytical limit for high steam line pressure (800 psia) at 11.7 seconds. Another two seconds is needed before the control rods are free to fall (Figure 8-96). Coincident with reactor trip at 13.7 seconds, the DHRS valves begin to open; the FWIVs begin to close; the MSIVs begin to close; and, the makeup line break opens fully (Figure 8-92). Opening the break in the makeup line causes the total break flow rate to increase sharply (Figure 8-92), then decrease and stabilize as critical flow conditions are established through the makeup line. Closing the FWIVs and MSIVs isolates the SGs from the remaining secondary system. The system shrinkage associated with the reactor trip works in conjunction with the increased break flow to cause the pressurizer level and RPV pressure to decrease. The analytical limit for low pressurizer level (35 percent) is reached at 86.3 seconds, and the pressurizer heaters are de-energized two seconds later. The pressurizer pressure continues to decrease until the analytical limit for low pressurizer pressure (1600 psia) is reached at 89.9 seconds. Containment isolation is initiated after a delay of 2 seconds. Closing the CVCS isolation valves terminates the break flow (Figure 8-92). The maximum integrated break flow of 12,940 lbm is reached at 97 seconds (Figure 8-97). After reactor trip and actuation of DHRS, oscillations are observed due to temperature and density differences between the riser and downcomer (as © Copyright 2022 by NuScale Power, LLC 636

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 discussed in Section 7.2); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. At ~25 minutes, RCS flow (Figure 8-98) has stabilized, and the RCS temperature (Figure 8-99) and pressure (Figure 8-100) are steadily decreasing as the DHRS transfers decay heat from the RPV to the reactor pool. The net reactivity (Figure 8-101) becomes negative shortly after reactor trip and remains negative during the transition to stable DHRS cooling. Lastly, the RPV level (Figure 8-102) remains well above the top of the core throughout this transient. It is concluded that, by 25 minutes, the transient has been terminated and that stable DHRS cooling has been achieved. No operator action was credited to mitigate this event. Table 8-11 Sequence of events for small line breaks carrying primary coolant outside containment Event Time (sec) Letdown line break (DEG, 100% area) occurs with coincident loss of normal AC 0 power; TSVs begin to close. Peak reactor power is reached (163.6 MW). 0 TSVs are fully closed. 0.1 Limiting MCHFR is reached (5.009 as calculated by NRELAP5). 1.0 High steam line 1 pressure limit is reached (800 psia). 11.7 High steam line 2 pressure limit is reached (800 psia). 11.7 RTS actuation on high steam line 1 pressure signal, control rods are inserted into 13.7 the core. CVCS makeup line break (DEG, 100% area) occurs. 13.7 DHRS actuation on the high steam line 1 pressure signal. DHRS actuation valves 13.7 begin to open. FWIVs and MSIVs begin to close. Peak RPV pressure is reached (1983 psia). 13.9 FWIVs and MSIVs are fully closed. 18.7 DHRS actuation valves are fully open. 43.7 Low PZR level limit is reached (35%). PZR heaters are disabled after 2 second 86.3 delay. Low PZR pressure is reached (1600 psia). Containment and CVCS isolation 89.9 begins after 2 second delay. CVCS isolation valves are fully closed. 96.9 Maximum integrated RCS break flow (11,940 lbm) occurs. 97 Establishment of stable RCS flow. Pressure and temperature are steadily 1500 decreasing. End of calculation. Stable DHRS cooling has been established. 3000 © Copyright 2022 by NuScale Power, LLC 637

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-92 Instantaneous break flow response (0 to 350 sec) for the representative small break outside containment event Figure 8-93 Pressurizer level response for the representative small break outside containment event © Copyright 2022 by NuScale Power, LLC 638

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-94 Reactor pressure vessel pressure response (0 to 350 sec) for the representative small break outside containment event Figure 8-95 Steam generator pressure responses for the representative small break outside containment event © Copyright 2022 by NuScale Power, LLC 639

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-96 Core power response for the representative small break outside containment event Figure 8-97 Integrated break flow response (0 to 350 sec) for the representative small break outside containment event © Copyright 2022 by NuScale Power, LLC 640

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-98 Reactor coolant system flow rate response for the representative small break outside containment event Figure 8-99 Core outlet temperature response for the representative small break outside containment event © Copyright 2022 by NuScale Power, LLC 641

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-100 Reactor pressure vessel pressure response (0 to 3000 sec) for the representative small break outside containment event Figure 8-101 Net reactivity response for the representative small break outside containment event © Copyright 2022 by NuScale Power, LLC 642

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-102 Level above top of core response for the representative small break outside containment event 8.5.1.3 Conclusion A representative challenging case regarding integrated mass released was identified for a break in a small line carrying reactor coolant. The results of this case, as presented in Section 8.5.1.2, are subsequently used as input to a dose evaluation using the NuScale radiological consequences methodology to demonstrate the respective acceptance criteria are met. 8.5.2 Steam Generator Tube Failure The purpose of this section is to present the thermal-hydraulic and core neutronic responses of a representative NPM for a steam generator tube failure event. This event is evaluated for offsite and onsite doses, which includes an assessment of the fuel cladding integrity. Thus, the transient analysis results are provided for input to downstream radiological consequences evaluations. 8.5.2.1 Event Description The event description for a failure of the steam generator tube can be found in Section 7.2.19. Based on Section 7.2.19, the acceptance criteria for radiological consequences are potentially challenged by a steam generator tube failure event. For the representative NPM, the results of the sensitivity studies indicate the event scenario with the highest integrated mass released to the environment may differ from the event scenario with the longest iodine spiking duration. The limiting © Copyright 2022 by NuScale Power, LLC 643

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 radiological dose consequences are associated with the event scenario having the highest integrated mass released to the environment. Radiological Consequences - Mass Released The representative case presented here corresponds to one of the more limiting integrated mass released to the environment cases, i.e., a case with higher dose consequences. This case features the following conditions for the failure of a steam generator tube: Conservative initial condition biasing (as shown in Table 7-91) is applied in order to maximize the consequences of this event. This representative case is initialized at 102 percent reactor power. Double-ended guillotine break of 100 percent area at top of a single steam generator tube. Single active failure of primary MSIV to close on affected SG. No loss of normal AC Power. No operator action was credited in the representative case. 8.5.2.2 Analysis Results The following describes the event sequence of the representative steam generator tube failure event. Table 8-12 summarizes the sequence of events. Figure 8-103 through Figure 8-113 show some key parameters during the representative steam generator tube failure event. The tube failure occurs at time zero. The reactor coolant lost through the break causes an immediate decrease in pressurizer level (Figure 8-103) and RPV pressure (Figure 8-104). In contrast, the SG pressure remains constant (Figure 8-104) as the flow from the tube failure is carried to the turbine. The pressurizer heater power increases in an attempt to compensate for the pressure reduction. The flow through the failed tube causes the pressurizer level to reach the analytical limit for low pressurizer level (35 percent) at 146.0 seconds. The pressurizer heaters are de-energized one second later, while the control rods are free to fall after two seconds (Figure 8-105). The system shrinkage associated with the reactor trip works in conjunction with the break flow to increase the rate of RPV depressurization. The pressurizer pressure continues to decrease until the analytical limit for low pressurizer pressure (1600 psia) is reached at 168.4 seconds. Two seconds later the DHRS valves begin to open; the FWIVs begin to close; and the MSIVs begin to close. The SGs are isolated from the remaining secondary system when the FWIVs and MSIVs close five seconds later. However, the SAF of the primary MSIV on the affected SG to close delays isolation of that SG until the secondary MSIV closes 30 seconds after the DHRS actuation. Isolating the SGs terminates the releases to the environment, but does not terminate flow from the failed tube (Figure 8-106). Following isolation of the SGs, the water level for the affected SG increases and ultimately goes off-scale high, while the level for the intact SG is maintained at the nominal post-trip level © Copyright 2022 by NuScale Power, LLC 644

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 (Figure 8-107). The maximum integrated break flow to the environment (8477 lbm) is reached at 200.5 seconds (Figure 8-108). After reactor trip and actuation of DHRS, oscillations are observed due to temperature and density differences between the riser and downcomer (as discussed in Section 7.2); therefore the calculation is continued to verify that the module transitions into passive and stable DHRS cooling. At 30 minutes, RCS flow (Figure 8-109) has stabilized, and the primary system temperature (Figure 8-110) and pressure (Figure 8-111) are steadily decreasing as the DHRS transfers decay heat from the RPV to the reactor pool. The net reactivity (Figure 8-112) becomes negative shortly after reactor trip and remains negative during the transition to stable DHRS cooling. The RPV level (Figure 8-113) remains well above the top of the core for the entire transient. Since stable DHRS cooling has been achieved and minimal flow exists through the failed tube, the transient is terminated at 60 minutes. No operator action was credited to mitigate this event. Table 8-12 Sequence of events for steam generator tube failure Event Time (sec) Steam generator tube fails in SG1. 0 Peak RPV pressure is reached (1931 psia). 0.5 Peak reactor power is reached (165.4 MW). 3.5 Limiting MCHFR is reached (5.297 as calculated by NRELAP5). 21.0 Low PZR level limit is reached (35%). PZR heaters are disabled after 1 second 146.0 delay. RTS actuation on low PZR level signal, control rods are inserted into the core. 148.0 Low PZR pressure is reached (1600 psia). Containment isolation and DHRS 168.4 actuation begins after 2 seconds delay. DHRS actuation on the low PZR pressure signal. DHRS actuation valves begin to open. FWIVs and MSIVs begin to close. Primary MSIV for faulted SG remains 170.4 open. FWIVs and MSIV (intact SG) are fully closed. 175.4 DHRS actuation valves are fully open. 200.4 Secondary MSIV (faulted SG) is fully closed. 200.5 Maximum integrated break flow to SG (8477 lbm) occurs. 200.5 Peak steam generator pressure is reached (1382 psia). 305.5 Establishment of stable RCS flow. Pressure and temperature are steadily 1800 decreasing. End of calculation. Stable DHRS cooling has been established and pressure 6000 difference across failed tube minimized. © Copyright 2022 by NuScale Power, LLC 645

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-103 Pressurizer level response for the representative steam generator tube failure event Figure 8-104 Reactor pressure vessel and steam generator pressure responses (0 to 500 sec) for the representative steam generator tube failure event (tube failure occurs in SG1) © Copyright 2022 by NuScale Power, LLC 646

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-105 Core power response for the representative steam generator tube failure event Figure 8-106 Instantaneous break flow response for the representative steam generator tube failure event © Copyright 2022 by NuScale Power, LLC 647

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-107 Steam generator level response for the representative steam generator tube failure event Figure 8-108 Integrated break mass release to steam generator before isolation (0 to 500 sec) for the representative steam generator tube failure event © Copyright 2022 by NuScale Power, LLC 648

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-109 Reactor coolant system flow rate response for the representative steam generator tube failure event Figure 8-110 Core inlet and exit temperature responses for the representative steam generator tube failure event © Copyright 2022 by NuScale Power, LLC 649

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-111 Reactor pressure vessel and steam generator responses (0 to 6000 sec) for the representative steam generator tube failure event Figure 8-112 Net reactivity response for the representative steam generator tube failure event © Copyright 2022 by NuScale Power, LLC 650

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 Figure 8-113 Level above top of core response for the representative steam generator tube failure event 8.5.2.3 Conclusion A representative challenging case regarding integrated mass released was identified for the failure of a steam generator tube. The results of this case, as presented in Section 8.5.2.2, are subsequently used as input to a dose evaluation using the NuScale radiological consequences methodology to demonstrate the acceptance criteria are met. © Copyright 2022 by NuScale Power, LLC 651

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 9.0 Quality Assurance The NuScale Power, LLC Quality Assurance Program Description, MN-122626 (Reference 3), complies with the requirements of 10 CFR 50 Appendix B (Reference 19) and Quality Assurance Requirements for Nuclear Facility Applications, ASME NQA-1 2008 and NQA-1a-2009 Addenda (Reference 20). As described in Reference 2, the NRELAP5 code was developed following the requirements of NuScales QAP. The non-LOCA system transient analysis is performed and documented in accordance with NuScales QAP. © Copyright 2022 by NuScale Power, LLC 652

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 10.0 Summary and Conclusions The NuScale evaluation model used to evaluate an NPM system short term transient thermal-hydraulic response to non-LOCA events is presented in this report. The non-LOCA system transient evaluation model is developed following a graded approach in accordance with the guidance provided in RG 1.203. The NPM plant designs for which this evaluation model is applicable is described in this report. An NPM is a natural circulation pressurized water reactor (PWR) with a reactor core, two helical coil SGs, and a pressurizer integral to the reactor vessel. Many of the events analyzed for operating PWRs and in recent design certification applications are applicable to the NuScale designs. NuScale-specific events reflect unique aspects of the NuScale designs such as the DHRS and normal operation of the containment at vacuum conditions. The NPM designs are evaluated in detail to assure that a sufficiently broad spectrum of transients, accidents, and initiating events have been included in the scope of design basis analyses presented in FSAR Chapter 15. The design-basis events are categorized by type and expected frequency of occurrence so that limiting cases in each group may be quantitatively analyzed and specific acceptance criteria applicable to each postulated initiating event are applied. The NPM design basis events for which the non-LOCA evaluation model is applicable are identified. NRELAP5 is NuScales system thermal-hydraulics code used to simulate an NPM system response during both the non-LOCA and LOCA short-term transient event progression. The NRELAP5 code is described in the separate NuScale LOCA evaluation model topical report. Applicability of the NRELAP5 code for non-LOCA system transient analysis is presented in this report. The NRELAP5 code is applicable for calculation of an NPM thermal-hydraulic system response for the non-LOCA short-term transient event progression as part of this EM. This conclusion is based on the high-ranked phenomena identified from the non-LOCA and LOCA PIRT processes, separate effects and integral effects testing, code to code benchmarking, and appropriately conservative input for initial and boundary conditions. The non-LOCA transient analysis process is described in this report. The methodology for conservatively biasing initial and boundary conditions for event analysis is presented. Then, each initiating event is considered to identify the acceptance criteria that may be challenged during the event. For each non-LOCA event, a description of the event is provided including biases and conservatisms applied, sensitivity studies performed, single active failures and loss of power scenarios that challenge the event acceptance criteria. For each transient event, the acceptance criteria where margin to the limit may be challenged are identified. For these acceptance criteria, sensitivity calculations are performed as appropriate to confirm that suitably conservative inputs are specified and to determine conditions that result in minimum margin. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed demonstrate that margin to the acceptance criterion is maintained. For non-LOCA initiating events that actuate the decay heat removal system, the EM is applicable for the short-term transient progression; during this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained. © Copyright 2022 by NuScale Power, LLC 653

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 For selected non-LOCA events, representative system transient results are provided to demonstrate application of the evaluation model for an NPM. System transient calculations are executed for sufficient duration to demonstrate that the initiating event is mitigated and stable cooling is established. Results of representative calculations show that the maximum primary system and secondary system pressure acceptance criteria are not challenged in an NPM design. The representative results indicate that the primary system pressure is limited by the reactor safety valve and peak values are less than 2200 psia, compared to acceptance criteria for the representative NPM of 2310 psia or 2520 psia, depending on the event classification; in the representative results the maximum secondary side pressure is less than 1600 psia, compared to acceptance criteria for the representative NPM of 2310 psia or 2520 psia, depending on the event classification. Margin to other quantitative acceptance criteria for MCHFR, fuel centerline temperature, and radiological dose limits applicable for the non-LOCA events are demonstrated as part of separate downstream subchannel or accident radiological analyses, presented in separate reports, which are not part of the scope of this topical report. © Copyright 2022 by NuScale Power, LLC 654

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4 11.0 References

1. U.S. Nuclear Regulatory Commission, Transient and Accident Analysis Methods, Regulatory Guide 1.203, December 2005.

2. NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422-P, Revision 3.

3. NuScale Power, LLC, NuScale Power, LLC Quality Assurance Program Description, MN-122626, Revision 0.

4. U.S. Code of Federal Regulations, General Design Criteria for Nuclear Power Plants, Appendix A, Part 50 Chapter I, Title 10, Energy, (10 CFR 50 Appendix A).

5. U.S. Code of Federal Regulations, Contents of applications; technical information, Section 52.47, Part 52 Chapter I, Title 10, Energy, (10 CFR 52.47).

6. NuScale Power, LLC, Subchannel Analysis Methodology, TR-0915-17564-P-A, Revision 2.

7. Not Used

8. NuScale Power, LLC, Accident Source Term Methodology, TR-0915-17565-P-A, Revision 4.

9. U.S. Nuclear Regulatory Commission, Standard Review Plan, Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment, NUREG-0800, Chapter 15, Section 15.6.2, Rev. 2, July 1981.

10. NuScale Power, LLC, Evaluation Methodology for Stability Analysis of the NuScale Power Module, TR-0516-49417-P-A, Revision 1.

11. S.M. Modro et al., Multi-Application Small Light Water Reactor Final Report, Idaho National Engineering and Environmental Laboratory, INEEL/EXT-04-01626, December 2003.

12. Electric Power Research Institute, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, Volumes 1-5, EPRI NP-7450-A, Revision 10, September 2014.

13. Moghanaki, S. K. and Rahgoshay, M., Pressurizer Modeling: Using Different Thermodynamic Models and Comparing Results with RELAP Code Results, Applied Mechanics and Materials (Volumes 423-426), pp. 1444-1448, 2013.

14. U.S. Nuclear Regulatory Commission, Combined License Applications for Nuclear Power Plants, Regulatory Guide 1.206, June 2007.

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Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Revision 4

15. U.S. Nuclear Regulatory Commission, Standard Review Plan, Transient and Accident Analyses, NUREG-0800, Chapter 15, Revision 3, March 2007 and various Chapter 15 subsections.

16. U.S. Code of Federal Regulations, Definitions, Section 50.2, Part 50 Chapter I, Title 10, Energy, (10 CFR 50.2).

17. American National Standards Institute / American Nuclear Society, Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems, ANSI/ANS-58.9-2002 R2015.

18. U.S. Nuclear Regulatory Commission, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs, SECY-94-084, March 1994.

19. U.S. Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Appendix B, Part 50 Chapter I, Title 10, Energy, (10 CFR 50 Appendix B).

20. American Society of Mechanical Engineers, Quality Assurance Program Requirements for Nuclear Facility Applications, ASME NQA-1-2008, NQA-1a-2009 Addenda.

21. NuScale Power, LLC, Rod Ejection Accident Methodology, TR-0716-50350-P, Revision 2.

22. NuScale Power, LLC, Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P-A, Revision 1.

23. NuScale Power, LLC, Nuclear Analysis Codes and Methods Qualification, TR-0616-48793-P-A, Revision 1.

24. Kim, S.J., Turbulent film condensation of high pressure steam in a vertical tube of passive secondary consensation system, PhD thesis, Korea Advanced Institute of Science and Technology, 2000.

25. SwUM-0304-17023, Revision 8, NRELAP5 Version 1.4 Theory Manual.

26. NuScale Power, LLC, Extended Passive Cooling and Reactivity Control Methodology, TR-124587-P, Revision 0.

27. U.S. Code of Federal Regulations, Contents of applications; technical information, Section 52.137, Part 52 Chapter I, Title 10, Energy, (10 CFR 52.137).

28. NuScale Power, LLC, Statistical Subchannel Analysis Methodology, TR-108601-P (Supplement 1 to TR-0915-17564-P-A, Revision 2), Revision 2.

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29. SwUM-0304-17023, Revision 10, NRELAP5 Version 1.6 Theory Manual.

30. NuScale Power, LLC, "Methodology for the Determination of the Onset of Density Wave Oscillations (DWO)," TR-131981-P, Revision 0.

© Copyright 2022 by NuScale Power, LLC 657

LO-133397 : Affidavit of Mark W. Shaver, AF-133398 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Licensing Manager of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the method by which NuScale develops its Non-Loss-of-Coolant Accident Analysis Methodology. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled Non-Loss-of-Coolant Accident Analysis Methodology. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § AF-133398 Page 1 of 2

552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on 01/05/23. Mark W. Shaver AF-133398 Page 2 of 2}}