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LLC Submittal of Pressure and Temperature Limits Methodology, TR-130877, Revision 1
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LO-175289

October 29, 2024 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 Pressure and Temperature Limits Methodology, TR-130877, Revision 1

REFERENCE:

NuScale letter to NRC, NuScale Power, LLC, Revision 1 to Standard Design Approval Application, Part 2, Chapter 5, TR-130877-NP, "Pressure and Temperature Limits Methodology",

dated December 31, 2022 (ML23304A342)

NuScale Power, LLC (NuScale) hereby submits Revision 1 of the Pressure and Temperature Limits Methodology, (TR-130877). This revision includes changes made during the Standard Design Approval Application audit. The next revision of the Standard Design Approval Application (Revision 2) will reference this revision of TR-130877.

contains the proprietary version of the report entitled, Pressure and Temperature Limits Methodology, TR-130877, Revision 1. 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 requirement 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 Jim Osborn at 541-360-0693 or at josborn@nuscalepower.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 29, 2024.

Sincerely,

Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com LO-175289 Page 2 of 2 10/29/2024

Distribution: Mahmoud Jardaneh, Chief, New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC David Drucker, Senior Project Manager, NRC

Pressure and Temperature Limits Methodology, TR-130877-P, Revision 1, Proprietary Version : Pressure and Temperature Limits Methodology, TR-130877-NP, Revision 1, Nonproprietary Version : Affidavit of Mark W. Shaver, AF-175290

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com LO-175289

Pressure and Temperature Limits Methodology, TR-130877-P, Revision 1, Proprietary Version

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-175289

Pressure and Temperature Limits Methodology, TR-130877-NP, Revision 1, Nonproprietary Version

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

Pressure and Temperature Limits Methodology

TR-130877-NP Revision 1 Licensing Technical Report

Pressure and Temperature Limits Methodology

October 2024 Revision 1 Docket: 52-050

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Pressure and Temperature Limits Methodology

TR-130877-NP Revision 1 Licensing Technical Report

COPYRIGHT NOTICE

This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copi es 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 licens e, 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.

TR-130877-NP Revision 1 Licensing Technical Report

Department of Energy Acknowledgement and Disclaimer

This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.

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

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

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Table of Contents

Abstract................................................................... 1 Executive Summary.......................................................... 2 1.0 Introduction.......................................................... 3 1.1 Purpose.............................................................. 3 1.2 Scope................................................................ 3 2.0 Background.......................................................... 5 2.1 Regulatory Requirements and Recommendations............................. 5 2.1.1 General Design Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary................................................ 5 2.1.2 General Design Criterion 32 - Inspection of Reactor Coolant Pressure Boundary....................................................... 5 2.1.3 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation................ 5 2.1.4 10 CFR 50, Appendix G - Fracture Toughness Requirements.............. 6 2.1.5 10 CFR 50, Appendix H - Reactor Vessel Material Surveillance Program Requirements.................................................... 6 2.1.6 Generic Letter 96 Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits....... 6 2.1.7 Regulatory Guide 1.99 - Radiation Embrittlement of Reactor Vessel Materials........................................................ 6 2.1.8 10 CFR 50.61 - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events................................... 6 3.0 Analysis............................................................. 8 3.1 Materials.............................................................. 8 3.1.1 Neutron Fluence and Ferritic Materials............................... 13 3.2 Adjusted Reference Temperature......................................... 13 3.3 Scope of Pressure-Temperature Limits Analysis.............................. 14 3.3.1 Thermal Transients.............................................. 14 3.3.2 Heatup and Cooldown Rates....................................... 20 3.3.3 Fracture Mechanics.............................................. 20 3.3.4 Pressure and Temperature Limit Methodology......................... 20 3.4 Reactor Vessel Surveillance Program Consideration.......................... 27 3.5 Low Temperature Overpressure Protection.................................. 28 3.6 Pressurized Thermal Shock.............................................. 28

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Table of Contents

4.0 Results............................................................. 30 4.1 Adjusted Reference Temperature......................................... 30 4.2 Pressure Temperature Limits............................................. 30 4.3 Low Temperature Overpressure Protection Setpoint Limits..................... 39 4.3.1 Pressurized Thermal Shock Screening............................... 41 5.0 Summary and Conclusions............................................. 43 6.0 References.......................................................... 44 6.1 Source Documents..................................................... 44

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List of Tables

Table 1-1 Abbreviations.................................................... 4 Table 3-1 Reactor Pressure Vessel Material Distribution........................... 8 Table 3-2 Containment Vessel Material Distribution.............................. 9 Table 3-3 Pressure and Temperature Requirements for the Reactor Pressure Vessel... 27 Table 4-1 Pressure-Temperature Limits for NuScale Power Module Reactor Pressure Vessel per 10 CFR 50, Appendix G.................................. 31 Table 4-2 Summary of Pressure-Temperature Limits - Normal..................... 32 Table 4-3 Summary of Pressure-Temperature Limits - Inservice Leak and Hydrostatic Testing........................................................ 32 Table 4-4 Recommended Low Temperature Overpressure Protection Pressure Setpoint as a Function of Cold Temperature........................... 40

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List of Figures

Figure 3-1 Two Dimensional Model Material Distribution.......................... 10 Figure 3-2 Two Dimensional Model Reactor Pressure Vessel Material Distribution...... 11 Figure 3-3 Two Dimensional Model Reactor Pressure Vessel Material Distribution...... 12 Figure 3-4 Transient Temperature for Heatup................................... 16 Figure 3-5 Power Ascent Transient Definition - Temperatures and Convection Coefficients..................................................... 17 Figure 3-6 Power Descent Transient Definition - Temperatures and Convection Coefficients..................................................... 18 Figure 3-7 Cooldown Transient Definition - Temperatures and Convection Coefficients..................................................... 19 Figure 3-8 Representation of Postulated Semi-Elliptical Circumfe rential Cracks in Reactor Pressure Vessel Wall...................................... 21 Figure 3-9 Representation of Postulated Semi-Elliptical Axial Cracks in Reactor Pressure Vessel Wall............................................. 21 Figure 4-1 Pressure-Temperature Limits for Transient Inservice Leak and Hydrostatic Testing (Composite of Transients)................................... 33 Figure 4-2 Pressure-Temperature Limits for Steady-State Inservice Leak and Hydrostatic Testing............................................... 34 Figure 4-3 Pressure-Temperature Limits for Bounding Heatup and Power Ascent Transient Inservice Leak and Hydrostatic Testing....................... 35 Figure 4-4 Pressure-Temperature Limits for Bounding Power Descent and Cooldown Transient Inservice Leak and Hydrostatic Testing....................... 36 Figure 4-5 Pressure-Temperature Limits for Bounding Normal Heatup and Power Ascent Transient................................................ 37 Figure 4-6 Pressure-Temperature Limits for Bounding Normal Power Descent and Cooldown Transient.............................................. 38 Figure 4-7 Pressure-Temperature Limits for Core Critical Heatup/Power Ascent and Power Descent/Cooldown Transients................................ 39 Figure 4-8 Recommended Low Temperature Overpressure Protection Setpoint........ 41

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Abstract

This report describes the methodology used to develop the pressure-temperature (P-T) limits and the low temperature overpressure protection (LTOP) setpoint for the NuScale Power, LLC, NuScale Power Module (NPM). Plant operation within these limits protects the reactor coolant pressure boundary (RCPB) from non-ductile fracture.

This report bases its requirements and met hodology for developing P-T limits on the requirements and the methodologies in Title 10 of the Code of Federal Regulations (CFR)

Part 50 (10 CFR 50), Appendix G, and the American Society of Mechanical Engineering (ASME)

Boiler and Pressure Vessel Code (BPVC)Section XI, Appendix G; the P-T limits in the reactor pressure vessel (RPV) account for vessel embrittlement due to neutron fluence in accordance with Regulatory Guide (RG) 1.99. Representative P-T limits for the NPM are in tables and figures displaying maximum allowable reactor coolant system (RCS) pressure as a function of RCS temperature.

The NPM reactor vessel uses an LTOP system to provide protection against non-ductile failure due to LTOP events during reactor start-up and shutdown operation. The basis of the LTOP methodology in this report is ASME BPVC Section XI, Appendix G. The LTOP setpoints account for the effects of neutron embrittlement.

The basis for representative limits in this report is the projected 57 effective full-power years (EFPY) neutron fluence over the 60-year design life of the module. The P-T limits and LTOP setpoints applicable to operating modules are mo dule-specific based on material properties of as-built reactor vessels. Plant licensees provide these limits, based on the methods provided in this report.

10 CFR 50.61 requires pressurized thermal shock (PTS) screening for the RPV beltline region of pressurized water reactors (PWRs). A PTS event is an event or transient in PWRs causing severe overcooling (thermal shock) concurrent with or followed by significant pressure within the RPV.

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Executive Summary

There are a number of Nuclear Regulatory Commi ssion (NRC) regulations related to reactor coolant pressure boundary (RCPB) integrity, including General Design Criterion (GDC) 31; GDC 32; Title 10 of the Code of Federal Regulations (CFR) Part 50.60 (10 CFR 50.60);

10 CFR 50.61; 10 CFR 50, Appendix G; and 10 CFR 50, Appendix H. Collectively, these regulations require a licensee to ensure that the RCPB has sufficient margin to prevent non-ductile failure during all phases of operation, including postulated accident condi tions, accounting for material changes due to neutron fluence and temperature history over the life of the RCPB.

develop reactor vessel pressure-temperature (P-T) limits for the reactor pressure vessel (RPV), which are limitations on reactor operating pressure as a function of reactor coolant temperature for various operating conditions.

develop and maintain an appropriate surveillance program to monitor reduction in material toughness in ferritic materials over the life of the reactor vessel.

This report presents the methodologies used to demonstrate that the regulatory requirements identified above are met or are not applicable to the NuScale Power Module (NPM) reactor vessel. Historically, P-T limits were in the plants technical specifications. The NRC guidance in Generic Letter (GL) 96-03 provides a means of relocating the P-T limits to a pressure-temperature limits report (PTLR), which fa cilitates modifications to P-T limits as needed over the life of the plant. Moving the P-T limits from the technical specifications to the PTLR requires the licensee to develop methods and programs to address each of the following aspects:

neutron fluence calculation method adjusted reference temperature (ART) calculation method to account for the effects of neutron embrittlement minimum temperature requirements for the reactor vessel during various operational and testing modes reactor vessel surveillance program (RVSP) for ferritic steel the low temperature overpressure protection (LTOP) setpoint calculation method

This report addresses each of these topics as applicable to the NPM design. A licensee may use the methods found in this report to develop a PTLR rather than maintaining P-T limits in the plants technical specifications. This report also includes the pressurized thermal shock (PTS) screening results.

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1.0 Introduction

1.1 Purpose

This report describes the methodology used to develop the NuScale Power Module (NPM) heatup and cooldown curves (pressure-temperature (P-T) curves) and low temperature overpressure protection (LTOP) setpoints. Operation within these limits protects the reactor vessel from brittle frac ture. This report also provides an embrittlement analysis in accordance with Regulatory Guide (RG) 1.99 (Reference 6.1.1) and outlines whether the design requires a reactor vessel surveillance program (RVSP). This report includes the pressurized thermal shock (PTS) screening results.

1.2 Scope

This report provides a methodology for developm ent of P-T limits for the NPM reactor coolant pressure boundary (RCPB) including heatup and cooldown curves and P-T limits for normal operation.

the P-T limits for in-service leak and hydrostatic tests.

the LTOP setpoints.

In addition, this report provides values for each of these items based on assumed material properties at an exposure of 57 e ffective full-power years (EFPY) fluence, which represents the end-of-design-life neutron exposure based on a 60-year design life of the module with an assumed 95 percent capacity factor. This report does not provide P-T limits for use in an as-built NPM; the P-T limits must be created on a module-specific basis with consideration of the material properties of the as-built reactor pressure vessel (RPV). Licensees may reference the methods contained in this report to develop their module-specific pressure-temperature limits report (PTLR), or they may choose to develop an alternative methodology.

This report includes the PTS screening results.

In accordance with Generic Letter (GL) 96-03 (Reference 6.1.2), this report addresses the following five methodology aspects:

neutron fluence calculation method the adjusted reference temperature (ART) calculation method to account for the effects of neutron embrittlement, in accordance with Reference 6.1.1 minimum temperature requirements for the reactor vessel during various operational and testing modes based on Appendix G of Reference 6.1.3 the RVSP for ferritic steel the LTOP setpoint calculation method

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Table 1-1 Abbreviations Term Definition ART adjusted reference temperature ASME American Society of Mechanical Engineers BPVC Boiler and Pressure Vessel Code CNV containment vessel EFPY effective full-power years GDC General Design Criterion ISLH inservice leak and hydrostatic testing LTOP low temperature overpressure protection NPM NuScale Power Module NRC Nuclear Regulatory Commission P-T pressure and temperature PTLR pressure and temperature limits report PTS pressurized thermal shock RCPB reactor coolant pressure boundary RG Regulatory Guide RPV reactor pressure vessel RTNDT nil-ductility reference temperature RVSP reactor vessel surveillance program RVV reactor vent valve SIF stress intensity factor

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2.0 Background

This report outlines the P-T limits methodology and the LTOP setpoints methodology that can be used by a licensee to create a module-specific PTLR for an NPM. In addition, this report outlines the neutron fluence calculation method, ART calculation method, minimum P-T curves, RVSP recommendations, LTOP setpoint calculation method, and PTS screening results.

2.1 Regulatory Requirements and Recommendations

2.1.1 General Design Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary

General Design Criterion (GDC) 31 requires that the RCPB have sufficient margin to ensure that when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a non-brittle manner, and there is minimal probability of rapidly propagating fracture.

Changes in material properties must account for service temperatures and other conditions of the pressure boundary material under operating, maintenance, testing, and postulated accident conditions, as well as the uncertainties in determining material properties.

the effects of irradiation on material properties.

residual, steady state, and transient stresses.

size of flaws.

2.1.2 General Design Criterion 32 - Inspection of Reactor Coolant Pressure Boundary

General Design Criterion 32 requires that the RCPB be designed to permit periodic inspection and testing of important areas and an appropriate material surveillance program for the RPV.

2.1.3 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation

Regulation 10 CFR 50.60 (Reference 6.1.3) requires that light water reactors meet the fracture toughness and material survei llance program requirements set forth in Appendix G and Appendix H of Reference 6.1.3. Proposed alternatives to the requirements described in Appendix G and Appendix H of Reference 6.1.3 or portions thereof are allowed when the NRC grants an exemption under 10 CFR 50.12. The NPM design supports an exemption to 10 CF R 50.60 due to the absence of ferritic material in the RPV beltline region. The NPM design uses austenitic stainless steel in the lower RPV, which has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials, which increases the integrity and safety of the RCPB.

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2.1.4 10 CFR 50, Appendix G - Fracture Toughness Requirements

Appendix G of Reference 6.1.3 specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the RCPB of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation. Conditions of normal operation include anticipated operational occurrences and system hydrostatic tests to which the pressure boundary may be subjected over its service lifetime. The NPM design supports an exemption to 10 CFR 50.60 due to the absence of ferritic material in the RPV beltline region. The NPM design uses austenitic stainless steel in the lower RPV, which has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials, which increases the integrity and safety of the RCPB.

2.1.5 10 CFR 50, Appendix H - Reactor Vessel Material Surveillance Program Requirements

Appendix H of Reference 6.1.3 establishes the necessary material surveillance program to satisfy GDC 32 for light water reactors. Appendix H of Reference 6.1.3 requires that licensees establish and mainta in a material surveillance program to monitor changes in the fracture toughness proper ties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors. The materials in the reactor vessel beltline region undergo exposure to neutron irradiation and to the thermal environment. The NPM design supports an exemption to 10 CFR 50.60 due to the absence of ferritic material in the RPV beltline region. The NPM design uses austenitic stainless steel in the lower RPV, which has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials, which increases the integrity and safety of the RCPB. Upper RPV ferritic materials, which are outside the beltline, do not exceed the Appendix H of Reference 6.1.3 threshold for requiring an RVSP.

2.1.6 Generic Letter 96 Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits

Reference 6.1.2 provides information that describes the methodology that licensees may use to create PTLRs.

2.1.7 Regulatory Guide 1.99 - Radiation Embrittlement of Reactor Vessel Materials

Reference 6.1.1 provides general procedures that calculate the effects of neutron embrittlement of low-alloy steels used in light water reactor vessels.

2.1.8 10 CFR 50.61 - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Regulation 10 CFR 50.61 requires PTS screening for the RPV beltline region of pressurized water reactors (PWRs). A PTS event is an event or transient in PWRs causing severe overcooling (thermal shock) concurrent with or followed by significant pressure within the RPV. The NPM design supports an exemption to 10 CFR 50.61

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due to the absence of ferritic material in the RPV beltline region. The NPM design uses austenitic stainless steel in the lower RPV, which has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials, which increases the integrity and safety of the RCPB.

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3.0 Analysis

3.1 Materials

In accordance with Appendix G of Reference 6.1.3, the calculations in this report apply to the pressure-retaining components of the RCPB. Because Appendix G of Reference 6.1.3 only contains data and methods applicable to ferritic materials, and because the NPM lower RPV is not made of fe rritic materials, this report also evaluates ferritic materials in the upper RPV (i.e., the region above the upper flange).

Table 3-1 lists the materials in the RPV. Table 3-2 lists the materials in the containment vessel (CNV). Figure 3-1, Figure 3-2, and Figure 3-3 show the material distribution model for the RPV and the CNV.

Table 3-1 Reactor Pressure Vessel Material Distribution Component Material Lower Seismic Cap SA-693, Type 630, Condition H1100 Lower Head SA-965, Grade FXM-19 Lower Flange Core Region Shell SA-965, Grade FXM-19 Upper Flange Shell SA-508, Grade 3 Class 2 Upper Feed Plenum Shell SA-508, Grade 3 Class 2 Upper Steam Generator Shell SA-508, Grade 3 Class 2 Upper Support Ledge Shell SA-508, Grade 3 Class 2 Upper Support Ledge Shell Cladding Alloy 690 RPV - CNV Support Ledge SB-168, Alloy 690 Upper Steam Plenum Shell SA-508, Grade 3 Class 2 Upper Pressurizer Shell SA-508, Grade 3 Class 2 Upper Head SA-508, Grade 3 Class 2 Interior and exterior cladding, except for the RPV upper support 308L/309L ledge exterior cladding

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Table 3-2 Containment Vessel Material Distribution Component Material Lower Seismic Support Pads SA-479, Type 304 Lower Support Skirt SA-182, Grade F304 Lower Head SA-965, Grade FXM-19 Lower Core Region Shell SA-965, Grade FXM-19 Lower Transition Shell SA-965, Grade FXM-19 Buttering and Weld between the Lower Shell and Lower Transition Alloy 52/152 Shell Lower Shell SA-336, Grade F6NM Lower Flange SA-336, Grade F6NM Upper Flange SA-336, Grade F6NM Upper Support Ledge Shell SA-336, Grade F6NM Upper Steam Generator Access Shell SA-336, Grade F6NM Upper Intermediate Shell SA-336, Grade F6NM Upper Manway Access Shell SA-336, Grade F6NM Upper Seismic Support Shell SA-336, Grade F6NM Upper Head SA-336, Grade F6NM Control Rod Drive Mechanism To p Head Cover SA-182, Grade F6NM

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Figure 3-1 Two Dimensional Model Material Distribution

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

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Figure 3-2 Two Dimensional Model Reactor Pressure Vessel Material Distribution ((

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

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Figure 3-3 Two Dimensional Model Reactor Pressure Vessel Material Distribution ((

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

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3.1.1 Neutron Fluence and Ferritic Materials

Per Appendix G of Reference 6.1.3, the calculations in this report apply to the pressure-retaining components of the RCPB. The lower RPV (i.e., the region below the upper flange) undergoes exposure to higher neutron fluence than other portions of the RCPB; however, the NPM lower RPV is made of austenitic stainless steel rather than ferritic materials (Table 3-1). Despite the higher neutron fluence in the lower RPV region, the use of austenitic stainless steel ensures safety of the RCPB because austenitic stainless steel has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials.

Appendix G of Reference 6.1.3, provides the following definition of the RPV beltline.

Beltline or Beltline region of reactor vessel means the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

The NPM design does not require testing Charpy upper shelf energy per Appendix G of Reference 6.1.3, for the following reason. Appendix G of Reference 6.1.3 applies to ferritic materials, while the portion of the NPM in the beltline region is austenitic stainless steel. The ASME BPVC Section III, NB-2311, does not require impact testing for austenitic stainless steel because these materials do not undergo ductile-to-brittle transition temperature and have higher toughness than ferritic materials used for ASME BPVC Section III Class 1 pressure-retaining components. Because impact testing is not required for austenitic stainless steel, the nil-ductility reference temperature (RTNDT) cannot be calculated. The NRC endorsed ASME BPVC Section III in 10 CFR 50.55a.

Appendix H of Reference 6.1.3 specifically applies to ferritic steel because the requirements for an RVSP were developed for ferritic materials and there is no guidance for an RVSP for austenitic stainles s steel. Furthermore, austenitic stainless steel has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement than ferritic materials, which increases the integrity and safety of the RCPB. Because the lower RPV is aus tenitic stainless steel, the ferritic portion of the RPV that experiences the highest fluence is evaluated against the Appendix H of Reference 6.1.3 criteria requiring an RVSP. The upper RPV lower flange has a design life peak fluence less than 1E+17 n/cm 2, E > 1 MeV, Section III.A of Appendix H of Reference 6.1.3 does not require an RVSP.

3.2 Adjusted Reference Temperature

There is no ART for the NPM because there is no need to adjust the RT NDT for fluence because the peak neutron fluence at the top of the lower flange of the RPV is

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(( }}2(a),(c),ECI, which is less than the 1E+17 n/cm2, E > 1 MeV regulatory limit.

Appendix H of Reference 6.1.3 requires beltline material surveillance if the portions of the RPV experience a maximum fluence greater than 1.0E+17 n/cm 2, E > 1 MeV; however, the portion of the RPV experiencing the highes t neutron fluence is the lower RPV, which is made of austenitic stainless steel. The ASME BPVC Section III, NB-2311, does not require impact testing for austenitic stainless steels because they do not undergo ductile-to-brittle transition temperature and have higher toughness than ferritic materials used for ASME BPVC Section III Class 1 pressure-retaining components. Without impact testing, RTNDT cannot be calculated for austenitic stainless steel, and thus ART is not applicable. The NRC endorsed ASME BPVC Section III in 10 CFR 50.55a. Since the upper RPV is the only part of the RPV made of ferritic materials, an evaluation of the upper RPV experiencing the highest design lif e peak fluence indicates that the upper RPV neutron fluence would have to increase by a factor of (( }} 2(a),(c),ECI to experience a fluence greater than 1.0E+17n/cm 2, E > 1 MeV; therefore, there is no need to adjust the reference temperatures.

3.3 Scope of Pressure-Temperature Limits Analysis

In order to develop a P-T limits methodology for the NPM, this report calculates minimum P-T limits for the NPM upper RPV design based on the requirements of Appendix G of Reference 6.1.3 and based on the methodologies in ASME BPVC Section XI, Appendix G (Reference 6.1.5). Finite elemen t models simulate thermal transient stress and analyze fracture mechanics.

Selected transients bound the rate of temperature change experienced during normal operation. These bounding rates are either (1) a nominal maximum rate of temperature change greater than expected operation, such as 100 degrees F per hour for the average coolant temperature and 200 degrees F per hour for the pressurizer liquid temperature; or (2) the maximum rate possible with the plant-specific equipment. The maximum rates of temperature change translate to operational constraints on the temperature rate of change through limiting conditions of operation in the technical specifications that relate to the P-T limits.

This method of transient selection and enforce ment of associated operational constraints, through the technical specifications, ensures that the rates of change in temperature applied to the RPV used in developing the P-T limits exceed the rates of change permitted during plant operation.

3.3.1 Thermal Transients

Thermal transients, in the context of this evaluation, include two heat transfer mechanisms: convection and radiation.

Convection is considered on the following surfaces: internal surfaces of the RPV (free and forced convection)

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the annulus between the RPV and CNV when flooded during the heatup and cooldown transients (free convection) the outside of the CNV for locations submerged in the pool (free convection)

Radiation is considered in the following regions: between the RPV outer surface and the CNV inner surface between the lower and upper RPV in the gap in the RPV flange between the lower and upper CNV in the gap in the CNV flange between the upper CNV and the control rod drive mechanism access cover at the closure surface

Convection driven by condensation in the upper pressurizer is also a driving heat transfer mechanism that occurs during these transients when the pressurizer wall temperature dips below saturation temperature. This occurrence can increase the convective film coefficients.

The four thermal transients considered in this evaluation include: heatup. power ascent. power descent. cooldown.

This report creates P-T limit curves for the following transient conditions: heatup, including power ascent. The heatup transient begins with the annulus between the RPV and CNV flooded with water. cooldown, starting with power descent. The cooldown transient includes the annulus between the RPV and the CNV flooded with water. inservice leak and hydrostatic testing (ISLH). The ISLH considers both steady state and heatup/cooldown transient conditions.

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3.3.1.1 Heatup Transient

Figure 3-4 shows the heatup transient.

Figure 3-4 Transient Temperature for Heatup ((

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

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3.3.1.2 Power Ascent Transient

Figure 3-5 shows the power ascent transient.

Figure 3-5 Power Ascent Transient Definition - Temperatures and Convection Coefficients ((

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

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3.3.1.3 Power Descent Transient

Figure 3-6 shows the power descent transient.

Figure 3-6 Power Descent Transient Definition - Temperatures and Convection Coefficients ((

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

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3.3.1.4 Cooldown Transient

Figure 3-7 shows the cooldown transient.

Figure 3-7 Cooldown Transient Definition - Temperatures and Convection Coefficients ((

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

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3.3.2 Heatup and Cooldown Rates

The Heatup and Cooldown transients target a temperature change rate of 100 degrees F per hour for the average coolant temperature and 200 degrees F per hour for the pressurizer liquid temperature. These target rates are exceeded over short periods of time (on the order of five minutes); however, the target rate is met over longer durations (on the order of an hour). The transients, including the brief periods where the rates of temperature change are higher than the target rate of temperature change, are used to generate the P-T limits, so the P-T limits account for these briefly higher rates of change. The possible heating rate is faster at lower temperature and slower at higher temperature. During plant operation, the temperature change rates are limited by operational constraints through limiting condition for operation 3.4.3.

3.3.3 Fracture Mechanics

This report analyzes axial and circumferential flaw locations at the most limiting thermal and pressure stress locations. Fracture mechanics analyses consider postulated flaws as follows: axial flaws: one-fourth thickness from the inner surface and one-fourth thickness from the outer surface. circumferential flaws: one-fourth thickness from the inner surface and one-fourth thickness from the outer surface.

3.3.4 Pressure and Temperature Limit Methodology

3.3.4.1 Pressure Boundary Components

In accordance with Appendix G of Reference 6.1.3, the calculations in this report bound the pressure-retaining components of the RCPB. The lower RPV is austenitic stainless steel (Table 3-1). Section 3.1 discusses the material distribution for the RPV and CNV. This evaluation considers the upper RPV because it contains ferritic materials.

3.3.4.2 Maximum Postulated Cracks

The methods of Appendix G, Article G-2214.1, of Reference 6.1.5 postulate the existence of a sharp surface crack in the RPV that is normal to the direction of the maximum stress. As specified in paragraph G-2120 of Reference 6.1.5, the crack depth is one-fourth of the RPV wall thickness, and the crack length is 1.5 times the wall thickness. This report considers both inside and outside surface cracks in axial and circumferential directions individually. For crack evaluations, a single crack is present in the RPV.

Figure 3-8 and Figure 3-9 show representations of circumferential and axial cracks, respectively.

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Figure 3-8 Representation of Postulated Semi-Elliptical Circumferential Cracks in Reactor Pressure Vessel Wall

Figure 3-9 Representation of Postulated Semi-Elliptical Axial Cracks in Reactor Pressure Vessel Wall

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3.3.4.3 Fracture Toughness

Appendix G, Article G-2110, of Reference 6.1.5 requires use of the critical stress intensity factor (SIF), KIC, defined by Equation 3-1, in P-T limits calculations.

K IC 33.2 20.734 exp 0.02 TR=+[]-()T NDTEq. 3-1

Where:

K IC = Critical SIF measuring fracture toughness (ksi in 0.5).

T = Temperature at crack tip (degrees F).

RT NDT = Reference temperature for nil-ductility transition (degrees F).

The conservative limit on upper shelf fracture toughness, K IC from Equation 3-1,

has an upper bound value of, which is slightly lower than the upper 200 ksi in 0.5 cutoff of lower bound in Appendix G, Article G-2212 of Reference 6.1.5. The K IC crack-tip temperatures needed for these fracture toughness calculations are from transient thermal analysis.

3.3.4.4 Fracture Mechanics Analysis

3.3.4.4.1 Calculation of Stress Intensity Factors due to Internal Pressure

Appendix G, Article G-2214.1, of Reference 6.1.5 provides a method to calculate KIm corresponding to membrane tension for postulated axial and circumferential cracks. This method applies to locations away from geometric discontinuity where calculation of hoop stress and axial stress occurs directly through an influence coefficient Mm (Mm_axial for axial cracks and Mm_circ for circumferential cracks).

For postulated axial cracks:

K Im_axial M m_axial pR i t=()Eq. 3-2

Where:

p = internal pressure (ksi).

R i = vessel inner radius (inches).

t = vessel wall thickness (inches).

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On the inside surface:

1.85 for t 4 in< M m_axial = Eq. 3-3 0.926 t for 4 in t 12 in 3.21 for t 12 in>

On the outside surface:

1.77 for t 4 in< M m_axial = Eq. 3-4 0.893 t for 4 in t 12 in 3.09 for t 12 in>

And for postulated circumferential cracks on the inside or outside surface:

K Im_circ M m_circ pR i t=()Eq. 3-5

0.89 for t 4 in< M m_circ = Eq. 3-6 0.443 t for 4 in t 12 in 1.53 for t 12 in>

Equation 3-2 through Equation 3-6 are not valid for cracks postulated at locations with a geometric discontinuity. A finite element analysis crack model calculates the SIFs due to pressure for all locations. A unit pressure (1 psig) is applied to the RPV inner surface. The SIFs for the crack tip node at the deepest point are calculated for five contours. The maximum value from contours two through five for the deepest point is the maximum SIF ( ) for K Im this evaluation. The first contour is not used because it is not accurate due to numerical inaccuracies in the stresses and strains at the crack tip.

3.3.4.4.2 Calculation of Stress Intensity Factors due to Thermal Stress

The hoop and axial thermal stresses are curve-fit to third order polynomial functions, which calculate thermal stress intensity factors KIT. The format of the polynomial function is:

c 0 c 1 x=++ + c 2 x--- 2 c 3 x--- 3---Eq. 3-7 a a a

Where,,, c 0 c 1 c 2 and c 3 are coefficients.

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=hoop stress or axial stress used to calculate SIF for postulated axial or circumferential crack (psi).

a =crack depth (inches).

x =distance from the appropriate (i.e., inside or outside) surface with xa= at the deepest crack tip (inches).

Appendix G, Article G-2214.3(b) of Reference 6.1.5 provides generic equations to calculate for radial thermal gradient for any thermal stress K IT distribution. For postulated axial and circumferential cracks away from geometry discontinuity, the following equations calculate SIFs.

For an inside surface crack during a cooldown transient:

K IT 1.0359 c 0 0.6322 c 1 0.4753 c 2 0.3855 c 3=() a+++Eq. 3-8

For an outside surface crack during a heat up transient:

K IT 1.043 c 0 0.630 c 1 0.481 c 2 0.401 c 3=() a+++Eq. 3-9

Where a is the crack depth (inches), and,, and are coefficients of c 0c 1c 2c 3 the third order polynomial equation for hoop or axial thermal stresses.

Equation 3-8 and Equation 3-9 are not accurate for cracks postulated at locations with a geometric discontinuity. A finite element analysis crack model calculates the SIFs due to transient thermal stresses by the superposition principle. To do so, a unit pressure (1psig) is applied to the crack top face and crack bottom face in four separate steps.

1. Constant unit pressure, set,, and in c 0 1=c 1 0=c 2 0=c 3 0=

Equation 3-7. The calculated SIF is.K It_c 0

2. Linear pressure along the crack depth direction, set,, c 0 0=c 1 1=

c 2 0= and in Equation 3-7. The calculated SIF is.c 3 0=K It_c 1

3. Quadratic pressure along the crack depth direction, set,, c 0 0=c 1 0=

c 2 1= and in Equation 3-7. The calculated SIF is.c 3 0=K It_c 2

4. Cubic pressure along the crack depth direction, set,, c 0 0=c 1 0=

c 2 0= and in Equation 3-7. The calculated SIF is.c 3 1=K It_c 3

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The SIFs for the crack tip node at the deepest point are calculated for five contours. The maximum value from the integrals of contour paths two through five is the maximum SIF. The proposed crack-specific equation to calculate SIFs for any axial/circumferential inside/outside surface cracks is:

K IT c 0 K It_c 0 c 1 K It_c 1 c 2 K It_c 2 c 3 K It_c 3=+++Eq. 3-10

Where,,, and are the actual coefficients of the 3 rd order c 0c 1c 2c 3 polynomial equation. If is negative, the allowable pressure calculation K IT uses a zero value.

3.3.4.5 American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Appendix G, Limits

This section documents the Appendix G of Reference 6.1.5 methodology for calculating the RPV allowable pressure for preservice hydrostatic test, normal heatup and cooldown transients, and ISLH conditions. This report documents development of a representative set of P-T calculations.

The ASME BPVC allowable pressure is part of the Appendix G of Reference 6.1.3 requirements. Except for the preservice hydrostatic test, the requirement of Appendix G of Reference 6.1.3 is that the test temperature must be greater than 50 degrees F.

The fundamental equation that is used to calculate P-T limits with a required safety margin is given by:

K I applied K IC= Eq. 3-11

Where is the lower bound crack initiation fracture toughness factor for the K IC material as represented in Equation 3-1, and is the stress intensity factor K I applied due to pressure and thermal gradient loads at the tip of the one-fourth T postulated cracks.

K I applied SF M m pR i t=() K IT+Eq. 3-12

Where is the required structural factor applied to the pressure loading, and SF dependent on which P-T limits curve is being evaluated, is the influence M m coefficient from Section 3.3.4.4.1, and is calculated using the K IT Section 3.3.4.4.2 methodology.

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The allowable pressure associated with a specified temperature along a P-T limits curve is:

() t- - P K IC K IT==------------------------------ K IC K IT---------------------------------------------------Eq. 3-13 SF M m R iSF K Im p()p1= psi

The appropriate and values used for various conditions are:K ITSF

For preservice hydrostatic tests, a steady-state condition ( ) is applied, K IT 0= and the required structural factor.SF 1=

P K IC t==------------------ K IC----------------------------------------Eq. 3-14 M m R i()p1K Im p= psi

Performance of the allowable pressure calculation occurs for the crack with highest that bounds other cracks. The basis for the preservice limiting M m pressure is NUREG-0800, Section 5.3.2 (Reference 6.1.4).

For the heat up and cooldown transients, the thermal SIF calculation occurs K IT at selected time points, and the required structural factor.SF 2=

() t- - P K IC K IT==------------------------------ K IC K IT-------------------------------------------Eq. 3-15 2 M m R i 2 K Im p()p1= psi

For ISLH, the SIF from heat up and c ooldown transients conservatively apply K IT to the most limiting crack, and the required structural factor.SF 1.5=

() t- - P K IC K IT==------------------------------ K IC K IT------------------------------------------------Eq. 3-16 1.5 M m R i 1.5 K Im p()p1= psi

3.3.4.6 10 CFR 50, Appendix G, Pressure and Temperature Limits

Appendix G of Reference 6.1.3 requires that the P-T limits are at least as conservative as limits obtained by following the Appendix G of Reference 6.1.5, methods presented in Section 3.3.4.5. Additionally, Table 1 of Appendix G of Reference 6.1.3 requires further limitations (Table 3-3).

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Table 3-3 Pressure and Temperature Requirements for the Reactor Pressure Vessel

Vessel Requirements for Minimum Temperature Operating Condition Pressure(1) Pressure-Temperature Requirements Limits Hydrostatic Pressure and Leak Tests (core is not critical) Fuel in the Vessel 20% ASME BPVC § XI App. G Limits (2) Fuel in the Vessel > 20% ASME BPVC § XI App. G Limits (2) + 90 degrees F (5) No Fuel in the Vessel (preservice hydrostatic all Not Applicable (3) + 60 degrees F test) Normal Operation (including heatup and cooldown), Including Anticipated Operational Occurrences Core Not Critical 20% ASME BPVC § XI App. G Limits (2) Core Not Critical > 20% ASME BPVC § XI App. G Limits (2) + 120 degrees F (5) Core Critical 20% ASME BPVC § XI App. G maximum of (4) or Limits + 40 degrees F ((2) + 40 degrees F) Core Critical > 20% ASME BPVC § XI App. G maximum of (4) or Limits + 40 degrees F ((2) + 160 degrees F) Notes:

1. Percent of the preservice syst em hydrostatic test pressure.

2. The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload.

3. The highest reference temperature of the vessel.

4. The minimum permissible temperature for the in-service system hydros tatic pressure test.

5. Lower temperatures are permissible if they can be justified by showing that the margins of safety of the controlling region are equivalent to those required for the beltline when it is controlling.

3.4 Reactor Vessel Surveillance Program Consideration

Appendix H of Reference 6.1.3 states:

The purpose of the material surveillance program required by this appendix is to monitor changes in the fracture toughness proper ties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment.

No material surveillance program is required for reactor vessels for which it can be conservatively demonstrated by analytical methods applied to experimental data and tests performed on comparable vessels, making appropriate allowances for uncertainties in the measurements, that the peak neutron fluence at the end of the design life of the vessel does not exceed 1E+17 n/cm 2, E > 1 MeV.

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Because Appendix H of Reference 6.1.3 applies to ferritic materials with peak neutron fluence at the end of the design life above 1E+17 n/cm 2, E > 1 MeV, the NPM reactor pressure vessel has no RVSP requirement in accordance with Appendix H of Reference 6.1.3 because the lower RPV is made of austenitic stainless steel and the maximum design life peak fluence of the ferri tic portion of the RPV is below 1E+17 n/cm2, E > 1 MeV. The NPM design supports an exemption to 10 CFR 50.60 due to the absence of ferritic materials in the RPV beltline region.

3.5 Low Temperature Overpressure Protection

The NPM reactor vessel uses LTOP systems for protection against failure during reactor start-up and shutdown operation due to LTOP ev ents classified as service level A or B events. Per Appendix G, paragraph G-2215, of Reference 6.1.5, LTOP systems must be effective at coolant temperatures less than 200 degrees F or at coolant temperatures corresponding to a reactor vessel metal temperature less than RT NDT + 50 degrees F, whichever is greater. ((

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

3.6 Pressurized Thermal Shock

Regulation 10 CFR 50.61 requires the PTS screening for the RPV beltline region of PWRs. A PTS event means an event or transi ent in PWRs causing severe overcooling (thermal shock) concurrent with, or followed by, significant pressure within the RPV. The 10 CFR 50.61 definition of beltline is:

(The) RPV beltline means the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

Regulation 10 CFR 50.61 (Reference 6.1.3) does not define significant radiation damage. However, Appendix H of Reference 6.1.3 requires the monitoring of ferritic RPV beltline materials with peak neutron fluence at the end of the design life exceeding 1E+17 n/cm 2, E>1MeV.

The 10 CFR 50.61 (Reference 6.1.3) PTS sc reening methodology is based on calculating the reference temperature for PTS (RT PTS). The RTPTS means RTNDT evaluated for the end of design life peak fluence for each of the vessel beltline materials using the 10 CFR 50.61 (Reference 6.1.3) procedures per the following 10 CFR 50.61 equation:

RT PTS RT NDT(u) RT NDT Margin=++Eq. 3-17

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The RTNDT(U) is the reference temperature RTNDT before service (unirradiated condition) established by impact testing per NB-2311 of Reference 6.1.6.

The 10 CFR 50.61(b)(2) (Reference 6.1.3) acceptance criteria for passing the PTS screening are: RTPTS not to exceed 270 degrees F for plates, forgings, and axial welds, and RTPTS not to exceed 300 degrees F for circumferential welds.

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4.0 Results

4.1 Adjusted Reference Temperature

There is no ART because there is no need to adjust RT NDT for fluence because the peak fluence at the top of the lower flange of the RPV is ((

}}2(a),(c),ECI, which is less than 1E+17n/cm2, E > 1 MeV.

Appendix H of Reference 6.1.3 requires survei llance of ferritic materials that experience a maximum fluence greater than 1.0E+17 n/cm 2, E > 1 MeV. The upper RPV fluence would have to increase by a factor of (( }}2(a),(c),ECI to experience a fluence greater than 1.0E+17n/cm2, E > 1 MeV; therefore, there is no need to adjust the reference temperatures.

4.2 Pressure Temperature Limits

The Appendix G of Reference 6.1.3 P-T limits are based on the requirements presented in Table 3-3. ((

}}2(a),(c),ECI Table 4-1 presents the application of Appendix G, Table 1 of Reference 6.1.3 to the NPM reactor pressure vessel. Figure 4-1 through Figure 4-7 present the uncorrected P-T limits curves.

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Table 4-1 Pressure-Temperature Limits for NuScale Power Module Reactor Pressure Vessel per 10 CFR 50, Appendix G

Requirements for Minimum Operating Condition Vessel Pressure (1) Pressure-Temperature Limits Temperature Requirements Hydrostatic Pressure and Leak Tests (core is not critical) Fuel in the Vessel 20 percent 535.3 psig ASME BPVC § XI App. G Limits 0 degrees F Fuel in the Vessel >20 percent 535.3 psig ASME BPVC § XI App. G Limits 90 degrees F No Fuel in the Vessel (preservice hydrostatic test) ALL 535.3 psig Not Applicable 60 degrees F Normal Operation (including heatup and cooldown), Including Anticipated Operational Occurrences Core Not Critical 20 percent 535.3 psig ASME BPVC § XI App. G Limits 0 degrees F Core Not Critical > 20 percent 535.3 psig ASME BPVC § XI App. G Limits 120 degrees F Core Critical 20 percent 535.3 psig ASME BPVC § XI App. G Limits + 40degrees F 90 degrees F Core Critical > 20 percent 535.3 psig ASME BPVC § XI App. G Limits + 40degrees F 160 degrees F Notes:

1. Percent of the preservice system hydrostatic test pressure.

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Table 4-2 Summary of Pressure-Temperature Limits - Normal

Composite Normal Composite Normal Normal Combined (Core Critical with RPV (Core Critical with RPV Normal Combined Heatup and Power Pressure = 20 Percent Pressure > 20 Percent Power Descent and Ascent Transient Pressure = 535.3 psig) Pressure = 535.3 psig) Cooldown (Core Not Critical) (Minimum core critical temperature determined from the steady state and transient ISLH curves) Fluid Pressure Fluid Pressure Fluid Pressure Fluid Pressure Temperature psig Temperature psig Temperature psig Temperature psig degrees F degrees F degrees F degrees F 65 535 Reactor is not permitted to be Reactor is not permitted to be 600 3260 critical below 90°F if ISLH critical below 160°F if ISLH 120 535 220 3260testing is performed at testing is performed at 120 2230 210 2400steady-state or transient steady-state or transient conditions. conditions. 150 2230 90 0 160 0 150 1875 200 2285 90 535 160 1875 120 1875 300 2475 160 535 190 1875 120 535 600 2475 160 1875 240 2285 65 535 190 1875 340 2475 240 2285 640 2475 340 2475 640 2475 Note: Linear interpolation can be used to calculate the allowable pressures for the temperatures not listed in the table.

Table 4-3 Summary of Pressure-Temperature Limits - Inservice Leak and Hydrostatic Testing

ISLH for Combined ISLH for Combined Transient ISLH Heatup and Power Power Descent and (Bounding) Steady-State ISLH Ascent Transient Cooldown Transient Fluid Pressure Fluid Pressure Fluid Pressure Fluid Pressure Temperature psig Temperature psig Temperature psig Temperature psig degrees F degrees F degrees F degrees F 65 535 600 4350 65 535 65 535 90 535 220 4350 90 535 90 535 90 2980 210 3200 90 2500 90 3660 150 2980 150 2500 150 2500 95 3960 200 3050 90 2500 200 3050 100 4300 300 3300 90 535 300 3300 105 4610 600 3300 65 535 600 3300 600 4610 Note: Linear interpolation can be used to calculate the allowable pressures for the temperatures not listed in the table.

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Figure 4-1 Pressure-Temperature Limits for Transient Inservice Leak and Hydrostatic Testing (Composite of Transients) In-Service Leak and Hydrostatic Test - Transient

PWD/RCD - LPZR PWD/RCD - Thot PWD/RCD - Tcold HTS/PAC - LPZR HTS/PAC - Thot HTS/PAC - Tcold Simplified Composite Bounding ISLH Transient

6,000

5,000

4,000

3,000

2,000

1,000

0 0 100 200 300 400 500 600 700 Temperature (°F)

Notes:

1. This image is intended to be viewed in color.

2. The following are abbreviations used in the figure above:

a. PWD: power descent

b. RCD: cooldown

c. LPZR: lower pressurizer region

d. Thot: reactor coolan t system hot temperature

e. Tcold: reactor coolant system cold temperature

f. HTS: heatup transient

g. PAC: power ascent

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Figure 4-2 Pressure-Temperature Limits for Steady-State Inservice Leak and Hydrostatic Testing ((

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

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Figure 4-3 Pressure-Temperature Limits for Bounding Heatup and Power Ascent Transient Inservice Leak and Hydrostatic Testing In-Service Leak and Hydrostatic Test - Heatup and Power Ascent Combined Transient HTS/PAC - LPZR HTS/PAC - Thot HTS/PAC - Tcold Simplified Bounding ISLH HTS/PAC Transient

4,000

3,500

3,000

2,500

2,000

1,500

1,000

500

0 0 100 200 300 400 500 600 700 Temperature (°F)

Notes:

1. This image is intended to be viewed in color.

2. The following are abbreviations used in the figure above:

a. HTS: heatup

b. PAC: power ascent

c. LPZR: lower pressurizer region

d. Thot: reactor coolan t system hot temperature

e. Tcold: reactor coolant system cold temperature

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Figure 4-4 Pressure-Temperature Limits for Bounding Power Descent and Cooldown Transient Inservice Leak and Hydrostatic Testing In-Service Leak and Hydrostatic Test - Power Descent and Cooldown Combined Transient

PWD/RCD - LPZR PWD/RCD - Thot PWD/RCD - Tcold Simplified ISLH Bounding PWD/RCD Transient 6,000

5,000

4,000

3,000

2,000

1,000

0 0 100 200 300 400 500 600 700 Temperature (°F)

Notes:

1. This image is intended to be viewed in color.

2. The following are abbreviations used in the figure above:

a. PWD: power descent

b. RCD: cooldown

c. LPZR: lower pressurizer region

d. Thot: reactor coolan t system hot temperature

e. Tcold: reactor coolant system cold temperature

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Figure 4-5 Pressure-Temperature Limits for Bounding Normal Heatup and Power Ascent Transient ((

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

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Figure 4-6 Pressure-Temperature Limits for Bounding Normal Power Descent and Cooldown Transient ((

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

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Figure 4-7 Pressure-Temperature Limits for Core Critical Heatup/Power Ascent and Power Descent/Cooldown Transients ((

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

4.3 Low Temperature Overpressure Protection Setpoint Limits

The LTOP setpoint limits the maximum pressure in the reactor vessel to less than the pressure limit curves in Figure 4-8. It uses the minimum pressure from the heatup and cooldown curves. Overpressure protection occurs by opening the two reactor vent valves (RVVs) located on the head of the reactor vessel when exceeding the LTOP pressure setpoint. For a given cold temperature, a pressurizer pressure above the LTOP setpoint causes the module protection system to send a RVV open signal. Above (( }}2(a),(c),ECI, the reactor safety valves provide overpressure protection.

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The LTOP logic and components c an continue to perform their function in the event of a single active failure.

The reactor safety valves do not lift when LTOP is enabled. This calculation accounts for pressure and temperature measurement uncertainties, the static pressure difference between the pressure measurement and the bottom of the RPV, the maximum delay in the RVV opening, and the delay in sensor response and module protection system processing time.

The pressurizer pressure determines the recommended LTOP setpoint; the LTOP setpoint has a conservative bias for the elevation head to the bottom of the RPV. Table 4-4 shows the recommended LTOP pressure setpoint as a function of reactor coolant system (RCS) cold temperature.

Table 4-4 Recommended Low Temperature O verpressure Protection Pressure Setpoint as a Function of Cold Temperature Tcold (degrees F) Pressurizer Pressure (psia)

<146.0 420.0 146.0 1750.0 175.0 1750.0 210.0 2025.0 290.0 2025.0
>290.0 LTOP not enabled

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Figure 4-8 shows the recommended LTOP setpoint along with the saturation pressure curve.

Figure 4-8 Recommended Low Temperature Overpressure Protection Setpoint 2400 2200 2000 1800 LTOP Variable 1600 Setpoint (PZR Pressure) 1400 Saturation Curve 1200 1000 800 600 400 200 0 50 100 150 200 250 300 350 400 450 Temperature ( °F)

4.3.1 Pressurized Thermal Shock Screening

The RTNDT(U) is the RTNDT before service (unirradiated condition) established by impact testing per NB-2331 of Reference 6.1.6.

The 10 CFR 50.61(b)(2) (Reference 6.1.3) acceptance criteria for passing the PTS screening are as follows: RTPTS not to exceed 270 degrees F for plates, forgings, and axial welds; and RTPTS not to exceed 300 degrees F for circumferential welds.

Per NB-2311 of Reference 6.1.6, austenitic stainless steels are exempt from impact test requirements and therefore are exempt from RT NDT requirements of NB-2331 of Reference 6.1.6. While 10 CFR 50.61 (Reference 6.1.3) does not specifically state that it applies only to ferritic materials, the chemistry factors in Table 1 and Table 2 of 10 CFR 50.61 (Reference 6.1.3) were derived for ferritic materials. Therefore, the PTS screening requirements in 10 CFR 50.61 (Reference 6.1.3) do not apply to the austenitic stainless steel used in the lower RPV of the NPM (Table 3-1). While there

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are ferritic materials in the upper RPV of the NPM, the 57 EFPY design life peak fluence for the top surface of the lower RPV flange is ((

}}2(a),(c),ECI. Hence, the design life peak fluence for the upper RPV shell is below the Appendix H of Reference 6.1.3 threshold value of 1E+17 n/cm 2, E > 1 MeV for the RPV beltline. Therefore, PTS screening of the upper RPV is not required, and PTS screening does not apply to the lower RPV shell.

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5.0 Summary and Conclusions

This report contains methodology based on Appendix G of Reference 6.1.3 and Appendix G of Reference 6.1.5 for the RCPB and P-T limits applicable to the NPM. An example set of P-T curves applicable to the NPM included in this report use these methods. These limits account for the effect s of neutron-induced embrittlement up to an exposure of 57 EFPY fluence. Curves developed include transient ISLH. steady-state ISLH. bounding heatup and power ascent transient ISLH. bounding power descent and cooldown transient ISLH. bounding normal heatup and power ascent. bounding normal power descent and cooldown. core critical.

The NPM design does not necessitate an RVSP to ensure adequate fracture toughness.

This report contains the LTOP limits and methodology for the NPM.

Using the material properties of an as-built reactor vessel, the licensee may use the methods developed in this report to develop their P-T limits and LTOP setpoints.

The PTS screening requirement of 10 CFR 50.61 (Reference 6.1.3) does not apply to the NPM reactor pressure vessel.

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6.0 References

6.1 Source Documents

6.1.1 U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Revision 2, May 1988.

6.1.2 U.S. Nuclear Regulatory Commission, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, Generic Letter 96-03, January 1996.

6.1.3 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy, (10 CFR 50).

6.1.4 U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, Revision 2, June 1987.

6.1.5 American Society of Mechanical Engi neers, Boiler and Pressure Vessel Code, 2017 Edition, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, New York, NY.

6.1.6 American Society of Mechanical Engi neers, Boiler and Pressure Vessel Code, 2017 Edition, Section III, Rules for Construction of Nuclear Facility Components, New York, NY.

Affidavit of Mark W. Shaver, AF-175290

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 Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale.

(2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following:

(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas.

(3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the process by which NuScale develops its Pressure and Temperature Limits Methodology.

NuScale has performed significant research and evaluation to develop a basis for this process 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, Pressure and Temperature Limits Methodology, TR-130877, Revision 1. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document.

AF-175290 Page 1 of 2 (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4).

(6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld:

(a) The information sought to be withheld is owned and has been held in confidence by NuScale.

(b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality.

(c) The information is being transmitted to and received by the NRC in confidence.

(d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence.

(e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 29, 2024.

Mark W. Shaver

AF-175290 Page 2 of 2}}

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2.0 Background

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