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LLC Submittal of Technical Report TR-1015-18177, Pressure and Temperature Limits Methodology, Revision 1
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Issue date:	04/06/2018
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LO-0418-59348 April 6, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Technical Report TR-1015-18177, Pressure and Temperature Limits Methodology, Revision 1

REFERENCES:

1. Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, Submittal of Technical Reports Supporting the NuScale Design Certification Application, dated December 30, 2016 (ML17005A112)

2. NuScale Technical Report Pressure and Temperature Methodology Revision 0, TR-1015-18177, dated December 2016 (ML17005A112)

In a letter dated December 30, 2016 (Reference 1), NuScale Power, LLC (NuScale) submitted the technical report titled Pressure and Temperature Limits Methodology, Revision 0 (Reference 2).

NuScale hereby submits Revision 1 of the Pressure and Temperature Limits Methodology, (TR-1015-18177). The purpose of this submittal is to request that the NRC review and approve the methodology presented in the report. contains the proprietary version of the report entitled Pressure and Temperature Limits Methodology, Revision 1 (TR-1015-18177-P, 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 2 contains the nonproprietary version of the report entitled Pressure and Temperature Limits Methodology, (TR-1015-18177-NP, Revision 1).

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

If you have any questions, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com if you have any questions.

Sincerely, Thomas A. A Bergman Vice President, Presi siident, Regulatory de Regula ato t ry Affairs NuScale Power, Powewe LLC err,, LLC Distribution: 5REHUW7D\ORU, NRC, OWFNH4A

Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury, OWFN-8G9A Bruce Bavol, NRC, OWFN-8G9A NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0418-59348 Page 2 of 2 04/06/2018 Enclosure 1: Pressure and Temperature Limits Methodology, TR-1015-18177-P, Revision 1 proprietary version Enclosure 2: Pressure and Temperature Limits Methodology, TR-1015-18177-NP, Revision 1 non-proprietary version Enclosure 3: Affidavit of Thomas A. Bergman, AF-0418-59349 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0418-59348 :

Pressure and Temperature Limits Methodology, TR-1015-18177-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-0418-59348 :

Pressure and Temperature Limits Methodology, TR-1015-18177-NP, Revision 1 non-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

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report Pressure and Temperature Limits Methodology April 2018 Revision 1 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 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 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.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report Department of Energy Acknowledgement and Disclaimer 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.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report CONTENTS Abstract ....................................................................................................................................... 1 Executive Summary .................................................................................................................... 2 1.0 Introduction ..................................................................................................................... 3 1.1 Purpose ................................................................................................................. 3 1.2 Scope .................................................................................................................... 3 1.3 Abbreviations and Definitions ................................................................................ 4 2.0 Regulatory Considerations ............................................................................................ 6 2.1 General Design Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary ................................................................................................ 6 2.2 General Design Criterion 32 - Inspection of Reactor Coolant Pressure Boundary ............................................................................................................... 6 2.3 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation .................................. 6 2.4 10 CFR 50 Appendix G - Fracture Toughness Requirements .............................. 6 2.5 10 CFR 50 Appendix H - Reactor Vessel Material Surveillance Program Requirements ........................................................................................................ 6 2.6 Generic Letter 96 Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits ................ 7 2.7 Regulatory Guide 1.99 - Radiation Embrittlement of Reactor Vessel Materials ................................................................................................................ 7 3.0 Design Inputs .................................................................................................................. 8 3.1 Reactor Pressure Vessel Beltline and Beltline Materials ....................................... 8 3.1.1 Definition of Reactor Pressure Vessel Beltline per 10 CFR 50 .................. 8 3.1.2 NuScale Reactor Pressure Vessel Beltline Extent and Materials .............. 8 3.2 Materials .............................................................................................................. 11 3.3 Heatup Transient ................................................................................................. 12 3.4 Cold Shutdown Valve Alignment Temperature .................................................... 13 3.5 Preservice Hydrostatic Test Pressure and Temperature ..................................... 14 3.6 Inservice Leak and Hydrostatic Test Pressure .................................................... 14 4.0 Components of the Pressure-Temperature Calculations .......................................... 15 4.1 Neutron Fluence .................................................................................................. 15 4.2 Evaluation of Adjusted Reference Temperature .................................................. 17 4.2.1 Adjusted Reference Temperature Definition ............................................ 17

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report 4.2.2 Initial Reference Temperature for Nil-Ductility Transition ......................... 17 4.2.3 Calculation of RTNDT .............................................................................. 17 4.2.4 Margin ...................................................................................................... 20 4.2.5 Adjusted Reference Temperature Calculation and Comparison to RG 1.99 Acceptance Criteria ................................................................... 20 4.3 Analysis of Cracks ............................................................................................... 22 4.4 Stress Analysis .................................................................................................... 25 4.4.1 Thermal Stress Analysis .......................................................................... 25 4.4.2 Fracture Toughness ................................................................................. 26 4.4.3 Calculation of Stress Intensity Factors due to Internal Pressure ............. 26 4.4.4 Calculation of Stress Intensity Factor KIT due to Thermal Stress ............ 28 4.4.5 Stress Intensity Factors due to Unit Internal Pressure ............................ 29 4.4.6 Transient Thermal Stresses ..................................................................... 30 5.0 ASME Code Section XI Appendix G Limits ................................................................. 33 5.1 Preservice Hydrostatic Test ................................................................................. 34 5.2 Allowable Pressure for Normal Heatup and Cooldown ....................................... 35 5.3 Allowable Pressure for Inservice Leak and Hydrostatic Tests ............................. 36 6.0 10 CFR 50 Appendix G Pressure and Temperature Limits ........................................ 37 7.0 Low Temperature Overpressure Protection Limits .................................................... 39 8.0 Surveillance Program ................................................................................................... 41 8.1 Fluence for RTNDT = 50°F .................................................................................. 41 8.2 Surveillance Capsule Withdrawal Schedule ........................................................ 41 9.0 Pressure-Temperature Curves ..................................................................................... 43 10.0 Summary and Conclusions .......................................................................................... 48 11.0 References ..................................................................................................................... 49 Appendix A. Thermal Stress Analysis in ANSYS ................................................................. 50 Appendix B. Crack Models ..................................................................................................... 56

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report TABLES Table 1-1 Abbreviations ......................................................................................................... 4 Table 1-2 Definitions .............................................................................................................. 5 Table 3-1 Lower reactor pressure vessel geometry ............................................................ 11 Table 3-2 Reactor pressure vessel beltline material chemistry ........................................... 11 Table 4-1 Fast neutron (energy > 1 MeV) fluence in reactor pressure vessel at 57 EFPY . 16 Table 4-2 Reactor pressure vessel beltline material properties........................................... 17 Table 4-3 RTNDT at 0-T ...................................................................................................... 19 Table 4-4 RTNDT at 1/4-T ..................................................................................................... 19 Table 4-5 RTNDT at 3/4-T ..................................................................................................... 19 Table 4-6 Calculation of Margin ........................................................................................... 20 Table 4-7 1/4-T adjusted reference temperature results at 57 effective full-power years fluence ................................................................................................................. 21 Table 4-8 3/4 -T adjusted reference temperature results at 57 effective full-power years fluence ................................................................................................................. 21 Table 4-9 Postulated cracks and applicable RTNDT ............................................................. 25 Table 4-10 Stress intensity factors due to unit loads, psi in ................................................. 29 Table 4-11 Comparison of KIm by formulation and finite element analysis, psi in ............... 30 Table 5-1 Allowable pressure for preservice hydrostatic test .............................................. 34 Table 6-1 Pressure and temperature requirements for the reactor pressure vessel ........... 37 Table 6-2 Pressure-temperature limits for NuScale reactor pressure vessel per 10 CFR 50, Appendix G .................................................................................................... 38 Table 7-1 Example variable low temperature overpressure protection pressure setpoint as a function of reactor coolant system cold temperature ................................... 40 Table 8-1 Fluence for RTNDT = 50°F .................................................................................. 41 Table 8-2 Estimated surveillance capsule withdrawal schedule .......................................... 42 Table 9-1 Pressure-temperature limits for normal heatup and cooldown ............................ 46 Table 9-2 Pressure-temperature limits for inservice leak and hydrostatic test .................... 47 Table A-1 Acronyms used in Appendix A ............................................................................. 50 Table A-2 Heat transfer K values for natural convection on flooded reactor pressure vessel outside diameter surfaces ........................................................................ 53 FIGURES Figure 3-1 Lower reactor pressure vessel ........................................................................... 10 Figure 3-2 Transient temperature for heatup ........................................................................ 12 Figure 3-3 Coolant temperature for cooldown ...................................................................... 13 Figure 3-4 CNV flooding water temperature for cooldown .................................................... 14 Figure 4-1 Postulated crack locations ................................................................................... 23 Figure 4-2 Postulated semi-elliptical circumferential cracks in reactor pressure vessel wall ...................................................................................................................... 24 Figure 4-3 Postulated semi-elliptical axial cracks in reactor pressure vessel shell............... 24 Figure 4-4 Temperature distribution in the lower reactor pressure vessel ............................ 31 Figure 4-5 Thermal hoop stress distribution in the lower reactor pressure vessel................ 32 Figure 5-1 Bounding allowable pressure for heatup and cooldown transients ..................... 35 Figure 5-2 Allowable pressure for inservice leak and hydrostatic test .................................. 36 Figure 9-1 Pressure-temperature limits for normal heat up and criticality limit ..................... 44 Figure 9-2 Pressure-temperature limits for normal cooldown with decay heat removal system and containment vessel flooding ............................................................. 45

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Licensing Technical Report Figure 9-3 Pressure-temperature limits for transient inservice leak and hydrostatic (composite of heatup and cooldown) and steady-state inservice leak and hydrostatic test .................................................................................................... 45 Figure A-1 ANSYS model for thermal analysis ..................................................................... 51 Figure A-2 Thermal boundary conditions .............................................................................. 52 Figure A-3 Containment vessel cavity volume slicing and flooding time ............................... 54 Figure A-4 Mechanical boundary conditions ......................................................................... 55 Figure B-1 Crack #1: Inside diameter circumferential crack at core support lower edge ...... 56 Figure B-2 Crack #2: Inside diameter circumferential crack at core support upper edge ..... 57 Figure B-3 Crack #3: Outside diameter circumferential crack at reactor pressure vessel alignment feature ................................................................................................. 57 Figure B-4 Crack #4: Outside diameter circumferential crack at reactor pressure vessel weld ..................................................................................................................... 58 Figure B-5 Crack #5: Inside diameter circumferential crack at reactor pressure vessel weld ..................................................................................................................... 58 Figure B-6 Crack #6: Outside diameter circumferential crack at flange ................................ 59 Figure B-7 Crack #7: Outside diameter circumferential crack at reactor pressure vessel shell ..................................................................................................................... 59 Figure B-8 Crack #8: Inside diameter circumferential crack at reactor pressure vessel shell ..................................................................................................................... 60 Figure B-9 Crack #9: Outside diameter axial crack at reactor pressure vessel shell ............ 60 Figure B-10 Crack #10: Inside diameter axial crack at reactor pressure vessel shell ............. 61 Figure B-11 Crack model for crack #2 mesh sensitivity study................................................. 62

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Abstract This report describes the methodology used to develop the pressure-temperature (P-T) limits and the low temperature overpressure protection (LTOP) set point for the NuScale Power, LLC (NuScale) standard plant. Plant operation within these limits protects the reactor coolant pressure boundary (RCPB) from non-ductile fracture.

The P-T limits developed in this report are based on the requirements and the methodologies in 10 CFR 50, Appendix G and ASME Section XI, Appendix G, and account for vessel embrittlement due to neutron fluence in accordance with RG 1.99, Revision 2. Representative P-T limits for the NuScale standard plant are presented as tables and figures displaying maximum allowable reactor coolant system (RCS) pressure as a function of RCS temperature.

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

Representative limits developed in this report are based on the projected 57 effective full-power years (EFPY) neutron fluence over the 60-year design life. The P-T limits and LTOP setpoints applicable to operating units are unit-specific based on material properties of as-built reactor vessels. These limits will be provided by plant licensees and can be based on the methods provided in this report.

The NuScale reactor vessel contains samples of the material used in the construction of the lower reactor pressure vessel (RPV). These samples are located inside the reactor vessel adjacent to the vessel wall at the beltline elevation, which is the level where the greatest neutron exposure is expected. The samples are exposed to approximately the same temperature as the vessel beltline. Owing to their location, the samples accumulate neutron damage at a faster rate than does the reactor vessel and, therefore, can be used as predictors of neutron damage to the vessel. Future changes in material properties of the vessel can be estimated by removing and testing a representative number of these samples. This report summarizes the reactor vessel material surveillance program.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Executive Summary There are a number of U.S. Nuclear Regulatory Commission (NRC) regulations related to reactor coolant pressure boundary (RCPB) integrity, including general design criterion (GDC) 31, GDC 32, 10 CFR 50.60, 10 CFR 50 Appendix G, and 10 CFR 50 Appendix H. Collectively, these regulations require a licensee to

ensure the RCPB is designed with sufficient margin to prevent non-ductile failure during all phases of operation, including postulated accident conditions, 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, which are limitations on reactor operating pressure as a function of reactor coolant temperature for various operating conditions.

develop and maintain a surveillance program to monitor reduction in material toughness over the life of the reactor vessel.

This report presents a methodology to demonstrate compliance with these requirements and provides a representative set of calculations and results for the NuScale standard reactor vessel.

Historically, P-T limits were included in the plants technical specifications. NRC guidance provided in GL 96-03 provides a means of relocating the P-T limits to a pressure-temperature limits report (PTLR), which facilitates modifications to P-T limits as they are needed over the life of the plant. Moving the P-T limits to the PTLR requires the licensee to develop methods and programs to address each of the following aspects:

1. neutron fluence calculation method

2. adjusted reference temperature (ART) calculation method to account for the effects of neutron embrittlement

3. limiting ART

4. minimum temperature requirements for the reactor vessel during various operational and testing modes

5. reactor vessel surveillance program (RVSP)

6. low temperature overpressure protection (LTOP) setpoint calculation method This report addresses each of these topics. 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.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 1.0 Introduction 1.1 Purpose The purpose of this report is to describe the methodology used to develop the NuScale Power, LLC (NuScale) standard plant heatup and cooldown curves (pressure-temperature curves) and low temperature overpressure protection (LTOP) set points.

Operation within these limits protects the reactor vessel from brittle fracture.

This report also provides an embrittlement analysis in accordance with Regulatory Guide (RG) 1.99 (Reference 11.3), and describes the reactor vessel surveillance program (RVSP) to be developed and implemented by the licensee.

1.2 Scope This report provides a methodology for development of pressure and temperature limits for the NuScale standard plant reactor coolant pressure boundary (RCPB) including:

heatup and cooldown curves and pressure-temperature (P-T) limits for normal operation

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

the LTOP set points In addition, this report provides values for each of these items based on assumed material properties at an exposure of 57 effective full-power years (EFPY) fluence, which represents the end-of-design-life neutron exposure based on a 60-year design life and an assumed 95 percent capacity factor.

The report does not provide pressure and temperature limits for use in an as-built NuScale Power Module (NPM); these limits must be created on a unit-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 unit-specific pressure-temperature limits report (PTLR), or may choose to develop an alternative methodology.

In accordance with GL 96-03 (Reference 11.4), this report addresses the following six methodology aspects:

1. neutron fluence calculation method

2. Adjusted reference temperature (ART) calculation method accounting for neutron embrittlement in accordance with Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99

3. limiting ART

4. minimum reactor vessel temperature as a function of pressure based on Appendix G to 10 CFR 50 (Reference 11.1)

5. the RVSP

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1

6. the LTOP setpoint calculation method 1.3 Abbreviations and Definitions Table 1-1 Abbreviations Term Definition APDL ANSYS parametric design language ART adjusted reference temperature ASME American Society of Mechanical Engineers CNV containment vessel EFPY effective full-power years GDC General Design Criterion ID inside diameter ISLH inservice leak and hydrostatic KIC critical stress intensity factor measuring fracture toughness KIM stress intensity factor due to membrane stress KIT thermal gradient stress intensity factor LTOP Low temperature overpressure protection MCNP Monte Carlo N-Particle Transport Code NPM NuScale Power Module NRC U.S. Nuclear Regulatory Commission OD outside diameter P-T pressure-temperature PTLR pressure-temperature limits report RAI request for additional information RCPB reactor coolant pressure boundary RCS reactor coolant system

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 RPV reactor pressure vessel RTNDT reference temperature for nil-ductility transition RTNDT shift in reference temperature for nil-ductility transition RVSP reactor vessel surveillance program SIF stress intensity factor Table 1-2 Definitions Term Definition 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 Beltline selection of the most limiting material with regard to radiation damage. This includes all regions of the reactor vessel for which the end of life fluence of neutrons with energy greater than 1 MeV is greater than 1017 n/cm2. Beltline is further defined in Section 3.0 of this document.

End of life For purposes of this report, end of life of the reactor vessel is 57 effective full-power years.

RTNDT The reference temperature for a reactor vessel material, under any conditions. For the reactor vessel beltline materials, reference temperature for nil-ductility transition (RTNDT) must account for the effects of neutron radiation.

RTNDT The transition temperature shift, or change in RTNDT, due to neutron radiation effects, that is evaluated as the difference in the 30 ft-lb (41 J) index temperatures from the average Charpy curves measured before and after irradiation.

Steady-state ISLH testing Inservice leak and hydrostatic (ISLH) testing performed after a one-hour soak period, during which the reactor coolant system (RCS) TAVG, THOT and TCOLD do not vary more than +/- 5 degrees Fahrenheit from the mean.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 2.0 Regulatory Considerations 2.1 General Design Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary General Design Criterion (GDC) 31 requires the RCPB to be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions: (1) the boundary behaves in a non-brittle manner, and (2) the probability of rapidly propagating fracture is minimized. Changes in material properties must account for service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining: (1) material properties; (2) the effects of irradiation on material properties; (3) residual, steady state, and transient stresses; and (4) size of flaws.

2.2 General Design Criterion 32 - Inspection of Reactor Coolant Pressure Boundary General design criterion 32 requires licensees to develop and maintain a material surveillance program for the RCPB.

2.3 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation 10 CFR 50.60 requires licensed light water reactors to meet the fracture toughness and material surveillance program requirements for the RCPB set forth in Appendices G and H of 10 CFR 50. Proposed alternatives to the described requirements in Appendices G and H or portions thereof may be used when an exemption is granted by the NRC under

§ 50.12.

2.4 10 CFR 50 Appendix G - Fracture Toughness Requirements 10 CFR 50 Appendix G 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, including anticipated operational occurrences and system hydrostatic tests, which the pressure boundary may be subjected to over its service lifetime.

2.5 10 CFR 50 Appendix H - Reactor Vessel Material Surveillance Program Requirements 10 CFR 50 Appendix H requires licensees to establish and maintain a material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors that result from exposure of these materials to neutron irradiation and the thermal environment.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 2.6 Generic Letter 96 Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits Generic letter 96-03 provides information that is needed to describe the methodology that licensees may use to create PTLRs, and therefore directly applies to this report.

2.7 Regulatory Guide 1.99 - Radiation Embrittlement of Reactor Vessel Materials Regulatory Guide 1.99 provides general procedures for calculating the effects of neutron radiation embrittlement of low-allow steels used for light-water reactor vessels.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 3.0 Design Inputs In accordance with 10 CFR 50, Appendix G, the calculations in this report bound the ferritic materials for pressure-retaining components of the RCPB. The lower RPV (i.e.,

the region below the upper flange) is exposed to higher neutron fluence than other portions of the RCPB, which results in higher ART and reduced fracture toughness. The lower RPV also experiences higher thermal stress than other RCPB portions during cooldown transients with CNV flooding. In addition, cracks are postulated in the lower RPV, including locations of geometric discontinuity, to conservatively account for high local stress concentration. Combining these factors, the lower reactor vessel was determined to be more susceptible to failure than other portions of the reactor coolant pressure boundary, and the P-T limits calculation is focused on the lower RPV.

3.1 Reactor Pressure Vessel Beltline and Beltline Materials 3.1.1 Definition of Reactor Pressure Vessel Beltline per 10 CFR 50 Regulation 10 CFR 50 Appendix H provides the following threshold fluence before an RVSP is required:

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 all uncertainties in the measurements, that the peak neutron fluence at the end of the design life of the vessel will not exceed 1E+17 n/cm2, E > 1 MeV.

Regulation 10 CFR 50 Appendix G provides the following definition of 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.

When the RVSP requirements in 10 CFR 50, Appendix G and Appendix H, are taken together, the RPV beltline refers to the portion of the RPV whose design-life peak fluence exceeds 1E+17 n/cm2, E > 1 MeV. That is, the RPV ferritic materials (ferritic base metal, weld metal, and heat affected zone) within this RPV beltline must be monitored by an RVSP for irradiation effect on RTNDT and Charpy upper shelf energy.

3.1.2 NuScale Reactor Pressure Vessel Beltline Extent and Materials Figure 3-1 illustrates the lower RPV configuration. The lower RPV has no vertical welds and only one circumferential weld, which is between the lower RPV shell and RPV bottom head. Fluence calculations determined that the fluence (n/cm2, E >1MeV) at

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 57 EFPY was less than 1E+17 for the entire upper RPV. Therefore, the lower RPV beltline consists of the following three items:

lower RPV shell

lower RPV weld

RPV bottom head

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Lower RPV Shell SA-508 Grade 3 Class 1 Lower RPV Weld RPV Bottom Head SA-508 Grade 3 Class 1 Figure 3-1 Lower reactor pressure vessel

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 3-1 and Table 3-2 present the geometry and material chemistry of the lower reactor vessel.

Table 3-1 Lower reactor pressure vessel geometry Parameter Value RPV ID of cladding surface 96 RPV Base Metal Thickness 4.25 RPV ID Cladding Thickness 0.25 RPV OD Cladding Thickness 0.125 3.2 Materials Table 3-2 Reactor pressure vessel beltline material chemistry Maximum Limit, wt%

Material Cu Ni P S Co Lower RPV Shell SA-508 Grade 3 Class 1 0.06 0.85 0.010 0.010 0.05 Lower RPV Weld Low Alloy Steel Weld 0.06 0.85 0.012 0.015 0.05 RPV Bottom Head SA-508 Grade 3 Class 1 0.06 0.85 0.010 0.010 0.05

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 3.3 Heatup Transient The heatup transient assumed in this report is shown in Figure 3-2. In addition, a bounding heatup rate of 100 degrees Fahrenheit per hour is analyzed in this report.

Figure 3-2 Transient temperature for heatup

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 3.4 Cold Shutdown Valve Alignment Temperature The RCS cooldown transient assumed in this report is shown in Figure 3-3. At the time the reactor is cooled down to 350 degrees Fahrenheit, the CNV is flooded with reactor pool water at 40 degrees Fahrenheit. The temperature transient for water in the CNV is shown in Figure 3-4.

Figure 3-3 Coolant temperature for cooldown

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 3-4 CNV flooding water temperature for cooldown 3.5 Preservice Hydrostatic Test Pressure and Temperature In accordance with the requirements of ASME Section III, NB-6000 (Reference 11.7), a preservice hydrostatic test pressure will be performed at no less than 1.25 times the system design pressure, which is 2100 psia, resulting in a minimum preservice hydrostatic test pressure of 2625 psia. However, the test pressure has been conservatively defined as 2625 psig. The minimum specified hydrostatic test temperature of 70 degrees Fahrenheit is used in this report.

3.6 Inservice Leak and Hydrostatic Test Pressure The ISLH test pressure is assumed to be 2035 psia, which is ten percent above the normal operating pressure of 1850 psia as required by ASME Section XI (Reference 11.8), Table IWB-5230-1.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 4.0 Components of the Pressure-Temperature Calculations 4.1 Neutron Fluence This section provides a summary of the NuScale neutron fluence methodology, which is consistent with the guidance provided in RG 1.190 (Reference 11.5) with exceptions as described in NuScale Power technical report TR-0116-20781-P, Fluence Calculation Methodology and Results (Reference 11.10).

A Monte Carlo method using Monte Carlo N-Particle Transport Code (MCNP) 6.1 (Reference 11.9) was chosen to develop the fluence profile. MCNP6.1 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. The ENDF-B/VII.I pointwise (continuous energy) cross-section data from ENDF-B/VII.I are available with the MCNP6.1 package used in this analysis.

The neutron flux in the lower RPV, in units of n/cm2-sec, is calculated by using MCNP cylindrical mesh tallies of the neutron flux with a defined mesh structure superimposed over the region of interest. Each data block provides the mesh central coordinates in three dimensions as well as tally results and its relative error.

The fast neutron (E > 1 MeV) fluence (n/cm2) in the RPV is calculated using flux tallies with a 1 MeV energy cutoff. Neutron fluence is evaluated at several locations, including the surveillance capsules, by using MCNP mesh tallies.

The total fission neutron source intensity S in the NuScale module at a given power is determined by the following equation.

Eq. 4-1 10

=

1.602 x 10 where

Average number of neutrons produced per fission, P : Fission power defined (MW),

Keff : Effective multiplication factor (= 1.000), and Qave : The average recoverable energy per fission for all materials (MeV).

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Using the above equation, the calculated fission neutron intensity of the NuScale module operating at 160 MW is 1.24x1019 n/s at the beginning of the first operating cycle.

Exposure-averaged neutron flux profiles over multiple cycles were used in the best estimate calculation.

The conversion of neutron flux to an accumulated 57-EFPY fluence (in units of n/cm2) is completed by multiplying the calculated MCNP neutron flux at different operational cycles (in units of n/cm2-sec) with the time constant of each fuel cycle for a total of 57 EFPY.

Fluences, shown in Table 4-1, were determined at the surfaces using the Monte Carlo method described above. At depths of one-fourth and three-fourths the thickness of the material, fluence was determined using the method of RG 1.99:

= exp(0.24 ) Eq. 4-2 where

The depth (in inches) below 0-T.

Table 4-1 Fast neutron (energy > 1 MeV) fluence in reactor pressure vessel at 57 EFPY 57-EFPY, n/cm2, E > 1 MeV 0-T (a) 1/4-T(b) 3/4-T Lower RPV Shell, Flange (c) ((

Lower RPV Shell, Beltline Lower RPV Weld RPV Bottom Head2(a),(c) (a) 0-T = interface between ID cladding and ferritic material (ferritic base metal or ferritic weld metal) outside the ID cladding. (b)T refers to the thickness of the material. 1/4-T means 1/4 of the thickness of the material, as measured from the inner surface. (c) Because the attenuation equation (Eq. 4-2) is not suitable for the flange, the peak fluence for the entire flange volume is conservatively used to bound the flange fluence at 0-T, 1/4-T, and 3/4-T © Copyright 2018 by NuScale Power, LLC 16

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 4.2 Evaluation of Adjusted Reference Temperature In order to demonstrate that the reactor vessel will not become embrittled beyond acceptable limits, the ART is calculated and compared against the acceptance criterion, 200 degrees Fahrenheit, found in RG 1.99 and stated as follows: The copper content should be such that the calculated adjusted reference temperature at the 1/4-T position in the vessel wall at end of life is less than 200 degrees F. 4.2.1 Adjusted Reference Temperature Definition The ART is defined by the methodology in regulatory position 1.1 of RG 1.99: ART = Initial RT + RT + Margin Eq. 4-3 4.2.2 Initial Reference Temperature for Nil-Ductility Transition Initial RTNDT is the non-irradiated RTNDT. A set of initial RTNDT values for the standard plant is shown in the following table. Table 4-2 Reactor pressure vessel beltline material properties Location Material Initial RTNDT Lower RPV Shell SA-508 Grade 3 Class 1 -10°F max Lower RPV Weld Low Alloy Steel Weld -20°F max RPV Bottom Head SA-508 Grade 3 Class 1 -10°F max 4.2.3 Calculation of RTNDT 4.2.3.1 Regulatory Guide 1.99 Method The NRC-endorsed method for calculating the shift in RTNDT (RTNDT) is described in RG 1.99 Rev. 2 (Reference 11.3) and 10 CFR 50.61. Although they have different purposes, RG 1.99 Rev. 2 and 10 CFR 50.61 use the same RTNDT methodology as described below.

                                                   ,   .       =                                 Eq. 4-4

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 where CF = chemistry factor based on Cu and Ni contents, and is given by Tables 1 and 2 of RG 1.99 Rev. 2 or 10 CFR 50.61. FF = fluence factor calculated per the following equation: ( . . ) Eq. 4-5

f = fluence in units of 1E+19 n/cm2 The fluence for PTS screening is the clad-to-base-metal interface fluence (0-T fluence) per 10 CFR 50.61. The 1/4-T and 3/4-T fluences for the beltline region ART calculations are determined per Eq. 4-2. 4.2.3.2 Adjustment for Irradiation Temperature The RPV irradiation temperature is 497 degrees Fahrenheit for NuScale. Because 497 degrees Fahrenheit is below the applicable temperature range of RG 1.99 Rev. 2, adjustment is required by the following statement in RG 1.99 Rev 2: The [RG 1.99 Rev. 2] procedures are valid for a nominal irradiation temperature of 550°F. Irradiation below 525°F should be considered to produce greater embrittlement, and irradiation above 590°F may be considered to produce less embrittlement. The correction factor used should be justified by reference to actual data. Currently, there is no actual NuScale RPV data to justify a correction factor for RTNDT. The Nuclear Regulatory Commission's RAI No. 234 (Reference 11.11) requested that RTNDT be increased by one degree Fahrenheit for each degree Fahrenheit below 525 degrees Fahrenheit (1°F/1°F). Because NuScale Tcold is 28 degrees Fahrenheit below 525 degrees Fahrenheit, the adjustment can be expressed by the following equation for the NuScale RPV:

                                               =         ,   .         + 28°                         Eq. 4-6 RTNDT at depths of 0-T, 1/4-T, and 3/4-T are shown in the following tables.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 4-3 RTNDT at 0-T 0-T Fluence RTNDT, 57-EFPY Fluence n/cm2, Cu Ni FF CF RG1.99 Rev2 RTNDT (°F) E > 1 MeV (°F) Lower RPV Shell, Beltline (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Lower RPV Weld (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 82 (( }}2(a),(c) RPV Bottom Head (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Table 4-4 RTNDT at 1/4-T 1/4-T Fluence RTNDT, 57-EFPY Fluence n/cm2, Cu Ni FF CF RG1.99 Rev2 RTNDT (°F) E > 1 MeV (°F) Lower RPV Shell, Flange (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Lower RPV Shell, Beltline (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Lower RPV Weld (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 82 (( }}2(a),(c) RPV Bottom Head (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Table 4-5 RTNDT at 3/4-T 3/4-T Fluence RTNDT, 57-EFPY Fluence n/cm2, Cu Ni FF CF RG1.99 Rev2 RTNDT (°F) E > 1 MeV (°F) Lower RPV Shell, Flange (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Lower RPV Shell, Beltline (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) Lower RPV Weld (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 82 (( }}2(a),(c) RPV Bottom Head (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) © Copyright 2018 by NuScale Power, LLC 19

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 4.2.4 Margin The margin term is calculated per the following RG 1.99 equation: Eq. 4-7

                                                                         =2     U  +

where, U is the standard deviation for initial RTNDT and is assigned 0 degrees Fahrenheit when the initial RTNDT is measured from actual RPV materials. This is appropriate because actual RTNDT information will be available for the as-built reactor vessels, and is the standard deviation for RTNDT. Per RG 1.99, is assigned 17 degrees Fahrenheit for base metal and 28 degrees Fahrenheit for weld material, not to exceed one-half of RTNDT. The calculation of margin is summarized in Table 4-6. Table 4-6 Calculation of Margin Location U RTNDT/2 used* Margin Lower RPV Shell, Beltline 0 17 (( }}2(a),(c) Lower RPV Weld 0 28 (( }}2(a),(c) RPV Bottom Head 0 17 (( }}2(a),(c)

As mentioned in the discussion of margin, this is the lesser of or one-half of the RTNDT.

4.2.5 Adjusted Reference Temperature Calculation and Comparison to RG 1.99 Acceptance Criteria Using Eq. 4-3, the ARTs at depths of 1/4-T and 3/4-T and an exposure of 57 EFPY fluence were calculated. The resulting ARTs at 1/4-T, tabulated in Table 4-7, are less than the RG 1.99 limit of 200 degrees Fahrenheit. Table 4-8 provides the results of the ART calculations at 3/4-T. © Copyright 2018 by NuScale Power, LLC 20

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 4-7 1/4-T adjusted reference temperature results at 57 effective full-power years fluence Initial Location RTNDT Margin ART (°F) RG 1.99 Criterion RTNDT Lower RPV Shell, Flange -10 (( }}2(a),(c) < 200°F Lower RPV Shell, Beltline -10 (( }}2(a),(c) < 200°F Lower RPV Weld -20 (( }}2(a),(c) < 200°F RPV Bottom Head -10 (( }}2(a),(c) < 200°F Table 4-8 3/4 -T adjusted reference temperature results at 57 effective full-power years fluence Initial Location RTNDT Margin ART (°F) RTNDT Lower RPV Shell, Flange -10 (( }}2(a),(c) Lower RPV Shell, Beltline -10 (( }}2(a),(c) Lower RPV Weld -20 (( }}2(a),(c) RPV Bottom Head -10 (( }}2(a),(c) © Copyright 2018 by NuScale Power, LLC 21

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 4.3 Analysis of Cracks The methods of ASME Code, Section XI, Appendix G, for protection against failure of the vessel postulate the existence of a sharp surface crack in the RPV that is normal to the direction of the maximum stress. As specified in ASME Section XI, paragraph G-2120, the crack depth is assumed to be one-fourth of the RPV wall thickness, and the crack length is 1.5 times the wall thickness. Both inside and outside surface cracks in axial and circumferential directions are evaluated. Figure 4-1 illustrates the postulated crack locations, numbered to correspond with Table 4-9. Typical circumferential and axial cracks are shown in Figure 4-2 and Figure 4-3. For the RPV shell, axial cracks are postulated on both the inside and outside surfaces. For the circumferential weld connecting the RPV shell and the RPV bottom head, circumferential cracks are postulated on both the inside and outside surfaces. For locations with geometric discontinuity such as the RPV wall connection to the RPV flange, core support blocks, and RPV bottom head alignment feature, inside or outside cracks are postulated accordingly. Table 4-9 lists the postulated cracks along with the initial RTNDT and ART values from Table 4-7 and Table 4-8. © Copyright 2018 by NuScale Power, LLC 22

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 4-2 Postulated semi-elliptical circumferential cracks in reactor pressure vessel wall Figure 4-3 Postulated semi-elliptical axial cracks in reactor pressure vessel shell © Copyright 2018 by NuScale Power, LLC 24

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 4-9 Postulated cracks and applicable RTNDT Initial 57-EFPY Crack Type Location Notes RTNDT, °F RTNDT, °F Core support block lower

Geometric discontinuity 1 Circ. -10 (( }}2(a),(c) edge where pressure and Core support block upper thermal transient may 2 Circ. -20 (( }}2(a),(c) edge create high tensile stress
Core support block upper edge is close to the RPV weld so the weld ART is 3 Circ. RPV alignment feature -10 (( }}2(a),(c) conservatively used
Crack k-factor calculations are not covered by G-2214 of ASME Section XI 4 Circ. RPV weld OD -20 (( }}2(a),(c)
Weld with higher RTNDT and lower fracture 5 Circ. RPV weld ID -20 (( }}2(a),(c) toughness
Geometric discontinuity where pressure and thermal transient may 6 Circ. RPV OD at flange -10 (( }}2(a),(c) create high tensile stress
Crack k-factor calculations are not covered by G-2214 of ASME Section XI 7 Circ. RPV shell OD -10 (( }}2(a),(c)
RPV shell within the beltline has higher RTNDT 8 Circ. RPV shell ID -10 (( }}2(a),(c) and lower fracture toughness 9 Axial RPV shell OD -10 (( }}2(a),(c)
Hoop stress by pressure is 2 times the axial stress which may create high stress intensity factor for axial crack 10 Axial RPV shell ID -10 (( }}2(a),(c)
Cracks in the cylindrical shell are used as benchmark against the G-2214 formulations 4.4 Stress Analysis 4.4.1 Thermal Stress Analysis A thermal transient analysis is conducted first for each transient. Appendix A provides a description of the model used. The temperature field from the thermal transient analysis is then applied to the static structural model for stress analysis. The hoop and axial

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 stresses at the crack locations are curve-fit to third-order polynomial functions that are used to calculate thermal stress intensity factors (SIFs) KIT. The format of the polynomial function is: Eq. 4-8

                              =    +      +         +

where , , and are coefficients, and

                      = hoop stress or axial stress used to calculate the SIF for postulated axial or circumferential crack,
                      = crack depth, and
                       = distance from the appropriate (i.e., inside or outside) surface with          =    at the deepest crack tip.

4.4.2 Fracture Toughness The ASME Section XI, Appendix G, requires application of the critical SIF , in P-T limit calculations:

                           = 33.2 + 20.734exp[0.02(         )]                                      Eq. 4-9 where, K     = Critical SIF measuring fracture toughness,             ,

T = Temperature at crack tip, and RT = Reference temperature for nil-ductility transition. Consistent with industry practice, the upper shelf fracture toughness K from Eq. 4-9 is conservatively limited to an upper bound value of 200 . The crack-tip temperature needed for these fracture toughness calculations is obtained from transient thermal analysis as described in Section 4.4.1. 4.4.3 Calculation of Stress Intensity Factors due to Internal Pressure ASME Section XI, paragraph G-2214.1 (Reference 11.8) provides a method to calculate KIm , the stress intensity factor corresponding to membrane tension for postulated axial and circumferential cracks that can be used for locations away from geometric discontinuities where hoop stress and axial stress can be calculated directly through an influence coefficient ( _ for axial cracks, and _ for circumferential cracks). © Copyright 2018 by NuScale Power, LLC 26

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 For postulated axial cracks, _ = _ ( ) Eq. 4-10 on inside surface: 1.85 <4 Eq. 4-11 _ = 0.926 4 12 3.21 > 12 on outside surface: 1.77 <4 Eq. 4-12 _ = 0.893 4 12 3.09 > 12 and for postulated circumferential cracks on an inside or outside surface: _ = _ ( ) Eq. 4-13 0.89 <4 Eq. 4-14 _ = 0.443 4 12 1.53 > 12 where,

                    = internal pressure, ksi,
                     = vessel inner radius, inches, and
                    = vessel wall thickness, inches.

For cracks postulated at locations with geometric discontinuities, the above equations are not valid. A finite-element analysis crack model is used to calculate the SIFs due to pressure. A unit pressure (1 psi) is applied to the lower RPV inside surface. 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 2 through 5 is the maximum SIF ( ). The influence coefficient can be calculated for both axial and circumferential cracks, as: © Copyright 2018 by NuScale Power, LLC 27

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1

                      =                                                                               Eq. 4-15 Note that the above method to calculate SIF due to pressure can also capture any contribution from primary bending stresses at locations of geometric discontinuity.

4.4.4 Calculation of Stress Intensity Factor KIT due to Thermal Stress ASME Section XI, paragraph G-2214.3(b), provides equations to calculate KIT for radial thermal gradient for any thermal stress distribution. For postulated axial and circumferential cracks away from geometry discontinuities, the SIFs are calculated by the following equations. For an inside surface crack during a cooldown transient,

                           = (1.0359    0 + 0.6322  1 + 0.4753  2 + 0.3855 3 )                        Eq. 4-16 For an outside surface crack during a heatup transient,
                       = (1.043    0 + 0.630  1 + 0.481  2 + 0.401 3 )                                Eq. 4-17 where       is the crack depth, and , ,                and      are coefficients of the third-order polynomial equation for hoop or axial stresses calculated in Section 4.4.1.

A finite element analysis crack model is used to calculate the SIFs due to transient thermal stresses by the superposition principle. To do so, a unit pressure (1 psi) is applied to the crack top face and crack bottom face in four separate steps: Step 1: Constant unit pressure, set = 1, = 0, = 0 and = 0 in Eq. 4-8. The calculated SIF is _ . Step 2: Linear pressure along the crack depth direction, set = 0, = 1, =0 and = 0 in Eq. 4-8. The calculated SIF is _ . Step 3: Quadratic pressure along the crack depth direction, set = 0, = 0, = 1 and = 0 in Eq. 4-8. The calculated SIF is _ . Step 4: Cubic pressure along the crack depth direction, set = 0, = 0, =0 and = 1 in Eq. 4-8. The calculated SIF is _ . 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 2 through 5 is the maximum SIF. The proposed crack-specific equation to calculate SIFs for any axial or circumferential, inside- or outside-surface cracks is: © Copyright 2018 by NuScale Power, LLC 28

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1

                             =        _ +  _ +     _  +     _                                  Eq. 4-18 where , , and are the actual coefficients of the third-order polynomial equation. If is negative, a zero value is used conservatively in the allowable pressure calculation.

Similarly, the above method to calculate SIF due to thermal transients can also capture any contribution from secondary bending stresses at locations of geometric discontinuity. 4.4.5 Stress Intensity Factors due to Unit Internal Pressure The SIFs due to unit internal pressure (i.e., membrane tension), denoted by , and the components of the calculation of , denoted by _ , _ , _ and _ , are calculated using the crack models for the postulated cracks in Table 4-9. See Appendix B for the crack models. Table 4-10 lists these values, which will be used to calculate thermal SIFs and allowable pressures in the following sections. Also shown in Table 4-10 (as Crack #11) are the results for Crack #2 with finer mesh, indicating good agreement with the coarser mesh. Table 4-10 Stress intensity factors due to unit loads, ((

                                                                                                   }}2(a),(c)

For the cracks in the RPV shell sufficiently far away from geometric discontinuities, the SIF can also be calculated using the formulation provided in ASME Section XI, paragraph G-2214.1, as shown in Section 4.4.3. A comparison of the calculated is provided in Table 4-11, showing good agreement between the finite element analysis method as summarized in Table 4-10 and the formulation methods. © Copyright 2018 by NuScale Power, LLC 29

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 4-11 Comparison of by formulation and finite element analysis, ((

                                                                                                       }}2(a),(c) 4.4.6    Transient Thermal Stresses The transient thermal stresses are calculated for the heatup and cooldown transients with the appropriate film coefficients on the RPV inside diameter (ID) and outside diameter (OD). The transient temperature results at selected time points of the thermal analysis are then applied to the static structural analysis to calculate the normal stresses in hoop and axial directions. The time points are selected based on the coolant temperatures that are used to construct the P-T limit curves, and also the time points that give maximum thermal gradient during heatup and cooldown transients.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 4-4(a) shows the transient temperature distribution at 43,800 seconds for the heatup transient. Figure 4-4(b) shows the transient temperature distribution at 26,500 seconds for the cooldown transient, which is about half an hour after CNV flooding is initiated. ((

                                                                                                }}2(a),(c)

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 4-5(a) shows the hoop stress distribution at 43,800 seconds for the heatup transient. Figure 4-5(b) shows the hoop distribution at 26,500 seconds for the cooldown transient. ((

                                                                                                 }}2(a),(c)

Figure 4-5 Thermal hoop stress distribution in the lower reactor pressure vessel The axial and hoop stresses are extracted from the stress analysis. The stresses in OD and ID 1/4-T are then calculated as a function of the radial distance from appropriate surfaces. The radial distance is normalized so that the value is 0 at the surface, and 1 at the cracks deepest point (the 1/4-T location). The stresses are then curve-fit to the third-order polynomial Eq. 4-8. © Copyright 2018 by NuScale Power, LLC 32

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 5.0 ASME Code Section XI Appendix G Limits This section documents the ASME Section XI, Appendix G, methodology for calculating the RPV allowable pressure for preservice hydrostatic test, normal heatup and cooldown transients, and ISLH conditions. A representative set of P-T calculations is developed. The ASME code allowable pressure is part of the 10 CFR 50 Appendix G requirements, as summarized in Table 6-1. Except for the preservice hydrostatic test, the only requirement of 10 CFR 50, Appendix G, is that the test temperature must be greater than 50 degrees Fahrenheit. The fundamental equation that is used to calculate P-T limits with a required safety margin is given by: Eq. 5.1

where is the lower-bound crack initiation fracture toughness for the material as represented in Eq. 4-9, and is the SIF due to pressure and thermal gradient loads, at the tip of the 1/4-T postulated cracks, Eq. 5-2

                                       =         (    )+

where is the required structural factor applied to the pressure loading and, dependent upon which P-T curve is being evaluated, is the influence coefficient given in Section 4.4.3, and is calculated by the methodology discussed in Section 4.4.4. The allowable pressure associated with a specified temperature along a P-T limit curve is given by: Eq. 5-3 ( )

                             =                 =

The appropriate and values used for various conditions are provided below:

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

Eq. 5-4

                             =            =

The allowable pressure is calculated for the crack with highest that bounds the other cracks. The preservice limiting pressure is based on NUREG-0800, Section 5.3.2 (Reference 11.2).

For the heatup and cooldown transients, the thermal SIF is calculated at selected time points, and the required structural factor = 2.

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 ( ) Eq. 5-5

                             =                 =

2 2

For ISLH tests, the SIF from heatup and cooldown transients are conservatively applied for the most limiting crack, and the required structural factor = 1.5.

( ) Eq. 5-6

                             =                 =

1.5 1.5 5.1 Preservice Hydrostatic Test The allowable pressure during the preservice hydrostatic test is calculated using Eq. 5-4. The fracture toughness is calculated for initial RTNDT provided in Table 4-9 and temperature of 70 degrees Fahrenheit. The in Table 4-10 is used for each postulated crack. The limiting allowable pressure is (( }}2(a),(c) psi for the preservice hydrostatic test (based on fracture toughness only; other requirements may apply to lower the limiting pressure). Table 5-1 Allowable pressure for preservice hydrostatic test ((

                                                                                                         }}2(a),(c)

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 5.2 Allowable Pressure for Normal Heatup and Cooldown The allowable pressure for normal heatup, postulated heatup at 100 degrees Fahrenheit per hour, and cooldown transients is calculated using Eq. 5-5. The fracture toughness is calculated for 57-EFPY RTNDT provided in Table 4-9 and the transient temperature at the postulated crack tip. The bounding curves for heatup and cooldown transients are plotted in Figure 5-1. Figure 5-1 Bounding allowable pressure for heatup and cooldown transients © Copyright 2018 by NuScale Power, LLC 35

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 5.3 Allowable Pressure for Inservice Leak and Hydrostatic Tests The allowable pressure for ISLH tests is calculated using Eq. 5-6. The bounding curves for heatup and cooldown transients in Figure 5-1 are scaled by the ratio of the safety factor for the heatup and cooldown conditions to that for the ISLH conditions. The allowable pressure for ISLH tests is shown in Figure 5-2. The transient ISLH curve is the lower bounding of the ISLH heatup and cooldown curves. The minimum permissible temperature corresponding to the ISLH test pressure of 2,035 psi is 287 degrees Fahrenheit for the transient ISLH curve, and 185 degrees Fahrenheit for steady-state ISLH curve, as indicated in Figure 5-2. Figure 5-2 Allowable pressure for inservice leak and hydrostatic test © Copyright 2018 by NuScale Power, LLC 36

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 6.0 10 CFR 50 Appendix G Pressure and Temperature Limits Appendix G to 10 CFR Part 50 requires that the P-T limits be at least as conservative as limits obtained by following the ASME Section XI, Appendix G, methods presented in Section 5.0. In addition, Appendix G to 10 CFR Part 50 specifies P-T limits and minimum temperatures for operation of a reactor vessel, dependent upon pressure, criticality, and the presence or absence of fuel. The requirements applicable to the NuScale design are presented in Table 6-1. Table 6-1 Pressure and temperature requirements for the reactor pressure vessel Minimum Vessel Operating Condition P-T Limits Temperature Pressure (1) Requirements Hydrostatic Pressure and Leak Tests (Core is not critical) Fuel in the vessel 20% ASME Appendix G Limits (2) Fuel in the vessel > 20% ASME Appendix G Limits (2) + 90°F(5) No fuel in the vessel ALL Not Applicable (3) + 60°F (Preservice Hydrotest) Normal Heatup and Cooldown, including CNV Flooding during Cooldown Core not critical 20% ASME Appendix G Limits (2) Core not critical > 20% ASME Appendix G Limits (2) + 120°F(5) ASME Appendix G Limits + The maximum of ((4)) Core critical 20% 40°F or ( (2)+40°F) ASME Appendix G Limits + The maximum of ((4) Core critical > 20% 40°F or ((2)+160°F) Notes: (1) Percent of the preservice system 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 leak and hydrostatic 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. ((

                                                                                        }}2(a),(c)

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 ((

                                                                                                          }}2(a),(c)

Table 6-2 Pressure-temperature limits for NuScale reactor pressure vessel per 10 CFR 50, Appendix G Minimum Vessel Operating Condition P-T Limits Temperature Pressure Requirements Hydrostatic Pressure and Leak Tests (Core is not critical) Fuel in the vessel 525 psig ASME Appendix G Limits 56°F Fuel in the vessel > 525 psig ASME Appendix G Limits 146°F (1) No fuel in the vessel ALL Not Applicable 50°F (Preservice Hydro test) Normal Heatup and Cooldown, including CNV Flooding during Cooldown Core not critical 525 psig ASME Appendix G Limits 56°F Core not critical > 525 psig ASME Appendix G Limits 176°F (2) Core critical (transient ISLH) ALL ASME Appendix G Limits + 40°F 287°F (3) Core critical (steady-state ALL ASME Appendix G Limits + 40°F 185°F (3) ISLH) Notes: (1) The ISLH pressure shall not exceed 525 psig when the temperature is below 146°F. The ISLH pressure shall not exceed the ASME Appendix G ISLH pressure limits for any temperature above 146°F. (2) During normal heatup and cooldown, the pressure shall not exceed 525 psig when the temperature is below 176°F. The pressure shall not exceed the ASME Appendix G pressure limits for any temperature above 176°F. (3) The core shall not be critical when the temperature is below the minimum temperature. For temperature above the minimum with the core critical, the pressure shall not exceed the ASME Appendix G pressure limits with a 40°F shift to the right hand side. © Copyright 2018 by NuScale Power, LLC 38

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 7.0 Low Temperature Overpressure Protection Limits This section presents an example of LTOP methodology applicable to the NuScale design based on values provided in the preceding sections. Each NuScale reactor vessel uses an LTOP system to provide protection against failure during reactor start-up and shutdown operation due to low temperature overpressure events. Per ASME Code, Section XI, Appendix G, LTOP systems shall be effective at coolant temperatures less than 200 degrees Fahrenheit, or at coolant temperatures corresponding to a reactor vessel metal temperature less than RTNDT+50 degrees Fahrenheit, whichever is greater. The maximum RTNDT is 120.4 degrees Fahrenheit. When the crack-tip temperature is 170.4 degrees Fahrenheit (=RTNDT+50 degrees Fahrenheit), the maximum coolant temperature is 303 degrees Fahrenheit for the postulated cracks, meaning that the LTOP enable temperature shall be no lower than 303 degrees Fahrenheit. The selected LTOP enable temperature would be set higher, for example 325 degrees Fahrenheit, to provide margin for uncertainty in RCS cold temperature measurement and for the possibility of differences in temperature between the sensor and other downcomer locations. The LTOP setpoint is designed to limit the maximum pressure in the reactor vessel to be less than the pressure limit curves in Figure 5-1. The minimum of the pressure from the heatup and cooldown curves is used. Overpressure protection is provided by opening the three reactor vent valves (RVVs) located on top of the reactor vessel when the LTOP pressure setpoint is exceeded. The LTOP pressure setpoint is a function of the wide range RCS cold temperature as defined in Table 7-1. Linear interpolation is used for temperatures between table entries. The mechanical design of the RVVs prevents the valves from opening above a threshold pressure. The constant LTOP pressure setpoint between 280 degrees Fahrenheit and 318 degrees Fahrenheit in this example ensures that the RVVs can open for all temperatures and pressures where LTOP is required. The LTOP setpoint includes margin to the reactor vessel pressure limit for uncertainty in the wide-range RCS pressure measurement, for the pressure difference between the sensor location and the bottom of the RPV, for uncertainty in the wide-range cold temperature measurement, for the temperature difference between the sensor and other downcomer locations, for delays in the pressure sensor and signal processing, and for delays for the RVVs to open. The pressure increase during the pressure sensor, signal processing, and valve-opening delays is determined using a bounding pressurization transient where the pressurizer is water solid with heat addition to the RCS from decay heat and the pressurizer heaters. The RCS inventory addition is not included in the pressurization transient as the RCS injection line is isolated when the pressurizer level is greater than 80 percent for modes of operation that require LTOP. © Copyright 2018 by NuScale Power, LLC 39

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 7-1 Example variable low temperature overpressure protection pressure setpoint as a function of reactor coolant system cold temperature RCS Cold Leg Temperature LTOP Pressure Setpoint

          °F                          psia 40                          350 190                         350 250                         745 318                         745
          >318                        LTOP not enabled Based on the parameters described above, an example variable LTOP pressure setpoint as a function of wide-range RCS cold temperature is defined in Table 7-1.

Linear interpolation is used for temperatures between table entries. If the wide range RCS pressure measurement is larger than the interpolated LTOP pressure setpoint, the module protection system opens the RVVs. The LTOP enable temperature in this example is 318 degrees Fahrenheit. © Copyright 2018 by NuScale Power, LLC 40

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 8.0 Surveillance Program 8.1 Fluence for RTNDT = 50°F The NuScale RVSP follows ASTM E185-82 as required by 10 CFR 50 Appendix H. Per Table 1 of ASTM E185-82, the first capsule withdrawal depends on when RTNDT of capsule specimens reaches 50 degrees Fahrenheit. Per Table 8-1, the fluence to cause 50 degrees Fahrenheit shift in weld metal is 4.27E+17 n/cm2, E > 1 MeV using the RTNDT methodology in Section 4.2.3. Because the capsule lead factor is 2.5, the corresponding EFPY can be determined with the following: 4.27 + 17 /

                                              = 0.55 1.77 + 19 /

(2.5) 57 where 1.77E+19 n/cm2 is the 57 EFPY peak fluence from Table 4-3 It is noted that 28 degrees Fahrenheit of the 50-degree Fahrenheit shift is due to the 1°F/1°F adjustment method requested by the NRC. The 28 degrees Fahrenheit adjustment is fixed irrespective of changes in fluence, therefore, overcompensates at low fluence levels. As 28 degrees Fahrenheit accounts for most of the 50-degree Fahrenheit shift, it is considered unlikely that weld metal specimens will reach 50 degrees Fahrenheit shift at 4.27E+17 n/cm2, E > 1 MeV. Table 8-1 Fluence for RTNDT = 50°F Fluence RTNDT, RG1.99 Rev2 n/cm2, Cu Ni FF CF (°F) RTNDT (°F) E > 1 MeV (=CF

FF)

Base Metal Specimens (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 37 (( }}2(a),(c) 50.0 Weld Metal Specimens (( }}2(a),(c) 0.06 0.85 (( }}2(a),(c) 82 (( }}2(a),(c) 50.0 8.2 Surveillance Capsule Withdrawal Schedule NuScale surveillance capsules contain specimens from the material used in the construction of the lower RPV shell. These specimens are located inside the vessel at the vessel beltline and are, therefore, exposed to a higher neutron flux and approximately the same temperature as is the lower RPV shell. Specimens are periodically removed and tested in order to monitor changes in fracture toughness in accordance with ASTM E185-82 (Reference 11.6) as required by 10 CFR

© Copyright 2018 by NuScale Power, LLC 41

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 50, Appendix H. The estimated withdrawal schedule for the surveillance capsules is shown in Table 8-2. The specimen lead factor is defined as the ratio of average fluence of specimens inside the capsule to the RPV inside surface peak fluence, and is approximately 2.5 for the design analyzed in this report. Therefore, the lead factor is consistent with ASTM E185-82 recommendation of less than 3.0. Table 8-2 Estimated surveillance capsule withdrawal schedule Sequence ASTM E185-82(a) Estimated Withdrawal 1st Whichever comes first First refuel outage (~2 EFPY) when highest

6-EFPY predicted RTNDT > ~50°F(b)
Capsule fluence > 5E+18 n/cm2, E > 1 MeV
Highest predicted RTNDT > ~50°F of all encapsulated materials 2nd Whichever comes first ~13 EFPY for capsule fluence to reach peak
15 EFPY 32-EFPY RPV inside surface fluence
Capsule fluence > peak 32-EFPY RPV
32 EFPY/2.5 = ~13 EFPY inside surface fluence 3rd Capsule fluence is between 1 and 2 times of peak Between ~13 EFPY and ~26 EFPY 32-EFPY inside surface fluence
32 EFPY/2.5 = ~13 EFPY
64 EFPY/2.5 = ~26 EFPY 4th Not required by ASTM E185-82
Withdrawal of the 4th capsule will be determined by applicable regulation at the time of license extension application.

(a) The withdrawal schedule for the first three capsules is in accordance with ASTM E185-82 for the initial 40-year operating license period. The 4th capsule is not required by E185-82 for the 40-year license. (b) The highest predicted RTNDT material to reach 50°F is the limiting weld metal at 4.27+17 n/cm2, E > 1 MeV (or 0.55 EFPY), as shown in Section 8.1. However, 28°F of the 50°F shift is due to the 1°F/1°F below 525°F irradiation temperature adjustment method requested by NRC RAI No. 234. The 28°F adjustment is fixed irrespective of changes in fluence, therefore, overcompensates at low fluence levels. It is considered unlikely that weld metal specimens will reach 50°F shift at 0.55 EFPY. Therefore, the first capsule can be withdrawn at the end of the first fuel cycle (2EFPY or 1.55E+18 n/cm2, E > 1 MeV), instead of mid-fuel-cycle. © Copyright 2018 by NuScale Power, LLC 42

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 9.0 Pressure-Temperature Curves Using the methodology provided in ASME Code, Section XI, Appendix G, and the requirements in 10 CFR 50, Appendix G, a representative set of P-T limits at 57-EFPY fluence was developed for various conditions. The P-T limits for normal heatup (core critical and core not critical), normal cooldown, and ISLH tests are provided in Figure 9-1, Figure 9-2, and Figure 9-3, respectively. The corresponding numerical values are listed in Table 9-1 and Table 9-2. These P-T curves meet the pressure and temperature requirements listed in Table 1 of 10 CFR 50, Appendix G. The RCS pressure should be maintained below the limit of the P-T limit curves to ensure protection against brittle failure. Acceptable pressure and temperature combinations for RCPB operation are below and to the right of the applicable P-T curves. The reactor is not permitted to be critical until the P-T combinations are to the right of the criticality curve shown in Figure 9-1. These P-T curves do not include location correction or instrument uncertainty. For the purpose of location correction, the allowable pressure in the P-T curves can be taken as the pressure at the RPV bottom. © Copyright 2018 by NuScale Power, LLC 43

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 9-1 Pressure-temperature limits for normal heat up and criticality limit The low temperature overpressure protection pressure limit at 57-EFPY is 502 psig for a fixed set point LTOP system or any pressure below the P-T limits curves in Figure 5-1 and Figure 5-2 for a variable set point LTOP system. The allowable pressure does not include any corrections for location or instrument uncertainty. The minimum 57-EFPY enable temperature is 303 degrees Fahrenheit, which does not include any instrument uncertainty. © Copyright 2018 by NuScale Power, LLC 44

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Figure 9-2 Pressure-temperature limits for normal cooldown with decay heat removal system and containment vessel flooding Figure 9-3 Pressure-temperature limits for transient inservice leak and hydrostatic (composite of heatup and cooldown) and steady-state inservice leak and hydrostatic test © Copyright 2018 by NuScale Power, LLC 45

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 9-1 Pressure-temperature limits for normal heatup and cooldown Normal Heat up (Core Critical) (Minimum core Normal Heat up (Minimum core critical critical temperature Normal Cooldown (Core Not Critical) temperature determined determined from the from the Steady-State transient ISLH ISLH curve) curve) Fluid Press. Fluid Press. Fluid Temp. Press. Fluid Press. Temp. (°F) (psig) Temp. (°F) (psig) (°F) (psig) Temp. (°F) (psig) 40 515 (Reactor is not permitted 309 2759 52 516 to be critical below 185°F if 300 2187 62 518 ISLH testing is performed 292 1697 78 523 (Reactor is not at steady-state conditions.) 285 1439 94 525 permitted to be critical 276 1280 below 287°F if ISLH 104 525 268 1168 testing is performed at 114 525 Heat up/Cooldown 185 0 260 1071 125 525 transient conditions) 185 150 252 1011 135 525 185 300 243 966 146 525 185 400 239 932 166 525 185 525 235 906 175 525 223 705 232 884 176 525 241 796 229 865 176 676 287 0 258 927 226 848 183 705 287 500 276 1115 223 833 201 796 287 750 295 1414 213 795 218 927 287 1000 314 1864 187 743 236 1115 287 1293 334 2518 176 720 255 1414 295 1414 354 2752 176 525 274 1864 314 1864 374 2750 162 525 294 2518 334 2518 393 2749 156 525 314 2752 354 2752 413 2747 156 525 334 2750 374 2750 426 2745 121 525 353 2749 393 2749 440 2742 118 525 373 2747 413 2747 454 2740 91 525 386 2745 426 2745 462 2735 79 525 400 2742 440 2742 469 2732 73 525 414 2740 454 2740 470 2729 73 525 © Copyright 2018 by NuScale Power, LLC 46

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 Table 9-2 Pressure-temperature limits for inservice leak and hydrostatic test Transient ISLH ISLH for Heat up ISLH for Cooldown (Bounding of Heat Steady-State ISLH Transient Transient up/Cooldown) Fluid Temp. Press. Fluid Temp. Press. Fluid Temp. Press. Fluid Temp. Press. (°F) (psig) (°F) (psig) (°F) (psig) (°F) (psig) 79 525 300 2916 79 525 40 525 91 525 292 2262 91 525 52 525 118 525 285 1918 118 525 62 525 121 525 276 1707 121 525 78 525 146 525 268 1557 146 525 94 525 146 790 260 1429 146 790 104 525 156 823 252 1348 156 823 114 525 156 823 243 1288 156 823 125 525 162 843 239 1243 162 843 135 525 187 968 235 1207 187 968 146 525 213 1185 232 1178 213 1060 146 525 223 1298 229 1153 223 1111 146 1247 226 1343 226 1130 226 1130 166 1556 229 1389 223 1111 229 1153 175 1737 232 1434 213 1060 232 1178 183 1952 235 1479 187 991 235 1207 201 2514 239 1552 162 922 239 1243 218 3312 243 1640 156 886 243 1288 252 1814 156 849 252 1348 260 2042 146 837 260 1429 268 2293 146 525 268 1557 276 2576 121 525 276 1707 285 2960 118 525 285 1918 292 3296 91 525 292 2262 300 3459 79 525 300 2916 Note: Linear interpolation can be used to calculate the allowable pressures for the temperatures not listed in the table. © Copyright 2018 by NuScale Power, LLC 47

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 10.0 Summary and Conclusions A methodology based on 10 CFR 50, Appendix G and ASME Section XI, Appendix G has been presented for the calculation reactor coolant pressure boundary pressure and temperature limits (P-T curves) applicable to the NuScale design. An example set of P-T curves applicable to the NuScale standard plant was developed using these methods. These limits account for the effects of neutron-induced embrittlement up to an exposure of 57 EFPY fluence. Curves developed include:

normal heatup core critical core not critical
normal cooldown
in-service leak and hydro tests In addition, the pressure was established for the preservice hydrostatic leak test.

Calculations were performed to predict the ART of the reactor vessel at 57-EFPY fluence. These calculations, based on the methodology of RG 1.99, were modified to account for the lower temperature at which a NuScale reactor vessel may operate. The results indicate that the NuScale standard reactor vessel 1/4-T ART has substantial margin to the RG 1.99 acceptance criteria of 200 degrees F max for new plants at the end of design life. The LTOP limits were developed for the NuScale standard plant, and a description of the reactor vessel material surveillance program was provided. 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 curves, LTOP setpoints, and reactor vessel surveillance program. © Copyright 2018 by NuScale Power, LLC 48

Pressure and Temperature Limits Methodology TR-1015-18177-NP Rev. 1 11.0 References 11.1 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy, (10 CFR 50). 11.2 U.S. Nuclear Regulatory Commission, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, NUREG-0800, Section 5.3.2, Revision 2, March 2007. 11.3 U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Revision 2, May 1988. 11.4 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. 11.5 U.S. Nuclear Regulatory Commission, Calculation of Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Regulatory Guide 1.190, March 2001. 11.6 ASTM International, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM E185-82, West Conshohocken, PA. 11.7 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition, Section III, Rules for Construction of Nuclear Facility Components, New York, NY. 11.8 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, New York, NY. 11.9 RSICC CODE PACKAGE CCC-810, Monte Carlo N-Particle Transport Code System Including MCNP6.1, MCNP5-1.60, MCNPX-2.7.0 and Data Libraries, Los Alamos National Laboratory. 11.10 NuScale Power, LLC, Fluence Calculation Methodology and Results, TR-0116-20781-P, Revision 0. 11.11 U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 234 (eRAI No. 9118)," dated September 21, 2017. © Copyright 2018 by NuScale Power, LLC 49

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LO-0418-59348 : Affidavit of Thomas A. Bergman, AF-0418-59349 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Thomas A. Bergman I, Thomas A. Bergman, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying technical report reveals distinguishing aspects about the method by which NuScale develops its pressure and temperature limits for the reactor vessel. 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 technical report entitled Pressure and Temperature Limits Methodology, TR-1015-18177-P, 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-0418-59349 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 perjuryy that the foregoing fore egooing n is true and correct. correc Executed on April 6, 2018. Th hom o as A. Bergma Thomas m n Bergman AF-0418-59349 Page 2 of 2}}

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