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LLC - Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Sections 3.7, 3.8, 3.9.2 and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report Ad the Density Wave.
	ML25029A202
Person / Time
Site:	05200050
Issue date:	01/29/2025
From:	Griffith T; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-178797 PM-178795, Rev 0
	Download: ML25029A202 (1)
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LO-178797 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 29, 2025 Docket No. 052-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Sections 3.7, 3.8, 3.9.2, and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report and the Density Wave Oscillation Safety Case),

PM-178795, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on February 4th, 2025. The materials support NuScales presentation of the subject sections, and technical report for the US460 Standard Design Approval Application.

The enclosure to this letter is the nonproprietary presentation entitled Sections 3.7, 3.8, 3.9.2, and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report and the Density Wave Oscillation Safety Case), PM-178795, Revision 0.

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

If you have any questions, please contact Jim Osborn at 541-360-0693 or at josborn@nuscalepower.com.

Sincerely, Thomas Griffith Director, Regulatory Affairs NuScale Power, LLC Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Engineer, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC

LO-178797 Page 2 of 2 01/29/2025 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

ACRS Subcommittee Meeting (Open Session) Sections 3.7, 3.8, 3.9.2, and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report and the Density Wave Oscillation Safety Case), PM-178795, Revision 0, Nonproprietary

LO-178797 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com ACRS Subcommittee Meeting (Open Session) Sections 3.7, 3.8, 3.9.2, and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report and the Density Wave Oscillation Safety Case), PM-178795, Revision 0, Nonproprietary

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)

February 4th, 2025 Sections 3.7, 3.8, 3.9.2, and Chapter 5 (Including the Pressure and Temperature Limits Methodology Technical Report and the Density Wave Oscillation Safety Case)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.

This presentation was prepared as an account of work sponsored by an agency of the United States (U.S.)

Government. Neither the U.S. 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 U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 February 4, 2025 Haydar Karaoglu and Emily Larsen Chapter 3 Design of Structures, Systems, Components and Equipment (Sections 3.7, 3.8, and 3.9.2)

Presenters:

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 3 - Design of Structures, Systems, Components and Equipment Section 3.7 - Seismic Design Section 3.8 - Design of Category I Structures Section 3.9.2 - Mechanical Systems and Components - Dynamic Testing and Analysis of Systems, Components, and Equipment Note: The presentation does not include Section 3.8.1, Concrete Containment, and Section 3.8.3, Concrete and Steel Internal Structures of Steel or Concrete Containments. The US460 NuScale Power Plant design does not use concrete containments or internal structures.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Overview of Key Design Features and Updates The Standard Design Approval Application (SDAA) is a derivative of the certified design.

SDAA structures reflect 6 modules (12 modules in the DC), which necessitated updated structural analyses.

For the SDAA, the Reactor Building (RXB) uses steel-plate composite (SC) walls along with reinforced concrete (RC) members.

The site layout in the SDAA reflects the updated building designs.

Seismic analyses for the SDAA are performed for a double-building model, featuring the RXB and Radioactive Waste Building (RWB) and a separate surface-based Control Building (CRB) model, while the design certification (DC) used a triple-building model and individual building models.

Presentation will focus on high level design and methodology changes and important audit questions and requests for additional information (RAIs).

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design Section 3.7.1 - Seismic Design Parameters Percentage of Critical Damping o The DC used separate fully cracked and fully uncracked models, and the RC members had the same damping ratio of 7 percent.

o The SDAA employs hybrid models with both cracked and uncracked members. The damping in the structural members varies based on their cracking status and whether the calculation is for developing in-structure response spectra (ISRS) or performing design calculations.

Building Design and Analysis Methodology for Safety-Related Structures, TR-0920-71621-P-A Supporting Medium o The DC included four generic soil profiles, Soil-7 (Rock), Soil-8 (Firm Soil/Soft Rock), Soil-9 (Hard Rock), and Soil-11 (Soft Soil).

o In the SDAA, Soil-8 is removed and the soil-separation scenario with the Soil-7 profile is introduced.

No audit questions or RAIs for Section 3.7.1

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design (Continued)

Section 3.7.2 - Seismic System Analysis Seismic Analysis Method o In the DC, soil-structure interaction (SSI) analyses were performed using the extended subtraction method with SASSI.

o In the SDAA, the SSI analyses are performed using the soil library methodology, a robust approach equivalent to the direct method. The soil libraries are built using SASSI and the simulations are performed with ANSYS.

Improvements in Frequency Domain Soil-Structure-Fluid Interaction Analysis, TR-0118-58005-P-A Three Components of Earthquake Motion o In the DC, the maximum responses were calculated using the square-root-of-the-sum-of-the-squares method.

o In the SDAA, the SSI responses from the three, statistically independent-components of the ground motion are algebraically added.

TR-0118-58005-P-A, Figure 4-1:

Idealized Soil, Structure, and Fluid Substructures

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design (Continued)

Section 3.7.2 - Seismic System Analysis (Continued)

SSI Numerical Models o In the SDAA, the reactor pool is modeled with FLUID elements of ANSYS and using the fluid-structure interaction (FSI) technology.

The 6 NuScale Power Modules (NPMs) are modeled in detail using advanced features of ANSYS.

o In the DC, the pool was modeled as distributed mass. The 12 NPMs were modeled using mass, spring, and beam elements (simplified beam model).

Audit Responses 33 questions resolved in audit, resulting in the following details and updates added to the Final Safety Analysis Report (FSAR) o modal analysis, double building model dimensions, and pool sloshing No RAIs for Section 3.7.2 Figure 3.7.2-2a: Isometric View of the Double Building (DB) Model

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design (Continued)

Section 3.7.3 - Seismic Subsystem Analysis The SDAA includes updates to major subsystems, including the bioshields, the reactor building crane, and the NPMs.

Three different NPM models have been developed o Simplified NPM model is used in SSI analyses to calculate seismic responses on RC and SC structural members.

o A detailed NPM model is used in SSI analyses to calculate the seismic response around the pool.

o A detailed NPM model with the use of the superelement technology of ANSYS is used for the nonlinear transient analysis.

(content reflected in Appendix 3A)

US460 NuScale Power Module Seismic Analysis, TR-121515-P Simplified NPM Model (TR-121515 Figure 3-1)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design (Continued)

Section 3.7.3 - Seismic Subsystem Analysis (Continued)

In the SDAA, the nonlinear NPM seismic analyses are conducted using a local model that includes the 6 NPMs, the pool, and the surrounding structural members.

US460 NuScale Power Module Seismic Analysis, TR-121515-P In the DC, the NPM seismic analyses were conducted using a local model that included only one NPM at a time, the pool, and a rigid plane under the NPM.

NuScale Power Module Seismic Analysis, TR-0916-51502-P-A Audit Responses 4 questions resolved in audit, resulting in additional bioshield details in the FSAR No RAIs for Section 3.7.3 Figure 3.7.2-7: NPMs within UHS (Local Seismic Model)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.7 - Seismic Design (Continued)

Section 3.7.4 - Seismic Instrumentation In the SDAA, the locations and descriptions of the seismic instrumentations are updated due to the new layout of the buildings.

No audit questions or RAIs for Section 3.7.4

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.8 - Design of SC-I Structures Section 3.8.2 - Steel Containment Differences from DC o Increase in design pressure and temperature for power uprate o Material change from carbon steel with cladding to combination of austenitic and martensitic stainless steels o Changed pre-service/in-service inspections from Class 1 to Class MC vessel with augmented requirements in some areas o Removed hydrogen detonation from load combinations because of added passive autocatalytic recombiners (Chapters 6 and 15) o Majority of nozzles changed from welded to integrally forged Audit Responses o 12 questions resolved in audit No RAIs for Section 3.8.2

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.8 - Design of SC-I Structures (Continued)

Section 3.8.4 - Other SC-I Structures In the SDAA o The RXB incorporates SC walls designed according to AISC N690-18 using element-and panel-based approaches.

o The RC members are designed according to ACI 349-13 using the section-cut forces at critical locations.

o The forces are calculated from numerical models with different cracked states associated with different load combinations.

o The simulations are performed using ANSYS with the use of SASSI for soil library calculations.

(content is also reflected in Appendix 3B)

Building Design and Analysis Methodology for Safety-Related Structures, TR-0920-71621-P-A In the DC, the major structural members were of RC type and designed according to ACI 349-06 using an element-based approach. The simulations were performed using SASSI and SAP2000.

Audit Responses 15 questions resolved in audit, resulting in the following updates to the FSAR o dynamic soil pressure, differential settlement analysis, definition of the supporting medium used for calculating the static load demands, and the design and analysis procedure (Appendix 3B)

No RAIs for Section 3.8.4

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.8 - Design of SC-I Structures (Continued)

Section 3.8.5 - Foundations Differences from DC o In the SDAA, the nonlinear stability analysis is performed only for the SC-I portion of the surface-based CRB.

o In the SDAA, the peak bearing pressure values are calculated using a methodology tailored to the capabilities of the software utilized, ANSYS.

Audit Responses o 12 questions resolved in audit No RAIs for Section 3.8.5

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.9.2 -

Dynamic Testing and Analysis of Systems, Components, and Equipment Differences from DC o Updated requirements from Regulatory Guide 1.20 Revision 3 to 1.20 Revision 4 o Updated requirements from the American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM)

Code, 2012 Edition to ASME OM Code, 2017 Edition o Comprehensive vibration assessment program (CVAP) startup instrumentation changed from strain gauges and accelerometers to dynamic pressure sensors o Removed Combined Operating License (COL) Item 3.9-14 (DC density wave oscillation (DWO) carveout) o Reactor vessel internals (RVI) were evaluated for updated US460 loads o Revised flow-induced vibration (FIV) analyses with US460 design changes and updated flowrates and operating conditions o Added inlet flow restrictor (IFR) cavitation evaluations with consideration of DWO to CVAP analysis report o Added an analysis case of both reactor vent valves (RVVs) actuating to TR-121517-P, "NuScale Power Module Short-Term Transient Analysis"

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.9.2 -

Dynamic Testing and Analysis of Systems, Components, and Equipment (Continued)

Audit Responses o 35 audit questions resolved

Added reference to startup test abstracts from Section 14.2 to FSAR 3.9.2.1

Updated language of NPM prototype classification options to match TR-121353-P, NuScale Comprehensive Vibration Assessment Program Analysis Technical Report

Provided summary of TF-3 (steam generator fluid-induced vibration (SGFIV)) flow testing results for review

Provided tube sliding and wear evaluation caused by the DWO transient

Provided DWO fatigue usage for tube-to-tubesheet weld, tubes, and tubesheet in the feedwater plenum RAI Results o RAI 10111 (Question 3.9.2-1) - Confirmation that steam generator (SG) integrity is maintained during Service Level D events

Provided preliminary Service Level D fatigue results for RVI and SG components

Resulted in no changes to the SDAA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 3.9.2 -

Density Wave Oscillation 10 audit questions resolved o 1 in Section 3.9.1, 9 in Section 3.9.2 No DWO RAIs in Chapter 3 Analyses o Section 3.9.1

DWO Service Level A Transient

NPM lifetime limit for time in DWO o Section 3.9.2

Structural integrity of steam generator during DWO

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ASME American Society of Mechanical Engineers COL Combined Operating License CRB Control Building CVAP Comprehensive Vibration Assessment Program DB Double Building DC Design Certification DWO Density Wave Oscillation FIV Flow-Induced Vibration FSI Fluid-Structure Interaction IFR Inlet Flow Restrictor ISRS In-Service Response Spectra ITP Initial Test Program NPM NuScale Power Module NRC Nuclear Regulatory Commission OM Operations and Maintenance RAI Request for Additional Information RC Reinforced Concrete RVI Reactor Vessel Internals RVV Reactor Vent Valve RWB Radioactive Waste Building RXB Reactor Building SC Steel-Plate Composite SG Steam Generator SGFIV Steam Generator Fluid-Induced Vibration SSI Soil-Structure Interaction SDAA Standard Design Approval Application

NuScale Nonproprietary Template #: 0000-21727-F01 R10 February 4, 2025 Wendy Reid and Erin Whiting Chapter 5 Reactor Coolant System and Connecting Systems Presenters:

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 5 - Reactor Coolant System and Connecting Systems Section 5.1, Summary Description Section 5.2, Integrity of Reactor Coolant Boundary Section 5.3, Reactor Vessel o Pressure and Temperature Limits Methodology Technical Report (TR-130877-P, Revision 1)

Section 5.4, Reactor Coolant System Component and Subsystem Design

¹ Denotes changes made in revision 2 of the Standard Design Approval Application (SDAA) Final Safety Analysis Report (FSAR)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 5 and Pressure and Temperature Limits Technical Report Review Audit Questions o 59 questions in Chapter 5 o 20 questions on Pressure and Temperature Limits Methodology Technical Report (PTLR)

Request for Additional Information (RAI) o 1 RAI in Chapter 5 o No RAIs on PTLR

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.1 - Summary Description Change in primary and secondary operating pressures, temperatures, and flow rates as a result of the power uprate Design pressure is the same for primary (inside the reactor vessel) and secondary (inside the steam generator tubes. Both design pressures changed from 2100 psi to 2200 psi Classification change for upper steam generator (SG) support for manufacturing concerns, requirements are consistent with American Society of Mechanical Engineers (ASME) code. ¹ Reactor coolant system (RCS) volume change ¹

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.2 - Integrity of Reactor Coolant Boundary Adopted 2017 ASME Boiler and Pressure Vessel and Operation and Maintenance Codes Change to leakage detection sensitivity requirement o No change to the equipment or system capabilities o No change to Technical Specifications for RCS leakage Change from three to two reactor vent valves The set points and design of the reactor safety valves (RSVs) changed o Setpoints increased with the design pressure increase and staggered o Minimum design capacity per valve increased¹ o Design change from pilot operated to spring operated RSVs Added the containment isolation test fixture (CITF) ¹ Augmented preservice examination for the Class 1 containment isolation valves (CIVs) and CITF on each of the four chemical and volume control system lines ¹ Augmented examinations applied to welds between containment vessel (CNV) and CIVs to support Branch Technical Position 3-4 requirements as discussed in Section 3.6 ¹ Low temperature overpressure protection setpoints changed due to material change for lower reactor pressure vessel (RPV)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.2 - Integrity of Reactor Coolant Boundary (Continued)

Changes to Table 5.2-3 reporting materials for reactor coolant pressure boundary components and support materials Lower RPV change discussed in Section 5.3 Added additional permissible materials to increase manufacturing flexibility for the combined license applicant Changes for consistency and completeness in response to audit questions Reconciled naming conventions with internal design documents

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.3 - Reactor Pressure Vessel Material change for the lower RPV to FXM-19 austenitic stainless steel o Change reflected in the PTLR methodology Technical Report o Upper RPV limiting ferritic component susceptible to fluence effects ¹ o Expansion to Combined Operating License (COL) Item 5.3-1 for PTLR ¹ o Exemptions for 10 CFR 50.60 fracture toughness (Appendices G and H) for and 10 CFR 50.61 pressurized thermal shock o Use of austenitic stainless steel in lower RPV o Superior ductility compared to ferritic materials o Less susceptible to the effects of neutron and thermal embrittlement than ferritic materials o Regulatory beltline concerns not an issue o No Appendix H material surveillance program required Removal of COL Item concerning onsite cleaning of the RPV during construction Removal of the flow diverter ¹ Change to seismic restraint feature between lower CNV and lower RPV ¹

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.4 - Reactor Coolant System Component and Subsystem Design Decay heat removal system (DHRS) o System size change o Credited in safety analysis; required for containment peak pressure response to a loss-of-coolant accident (LOCA)

(added to Chapter 5)¹ o Details on emergency core cooling system (ECCS) venting to limit hydrogen accumulation in the RPV during containment isolation ¹ o Design meets the intent of SECY 94-084 by achieving passively cooled, safe shutdown conditions within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ¹

DHRS performance cases achieve a passively cooled, safe shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Added off-nominal cases, including the worst case DHRS case (single train, high inventory), which provides sufficient cooling to below 450 degrees Fahrenheit RCS average temperature in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

o Actuation valve accumulator pressure details added ¹ Expanded description of SG supports ¹ Added description of flow paths between the riser and downcomer ¹ SG tube plugging criterion description changed due to bracketed value in Technical Specifications ¹

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 5.4 - Reactor Coolant System Component and Subsystem Design (Continued)

Design Certification (DC) approach o Ensure density wave oscillation (DWO) preclusion with inlet flow restrictor (IFR) sizing o DWO onset evaluation subject to future analysis o SG integrity to be determined during operation with DWO o COL Item 3.9-14 (DC DWO carveout)

DWO Safety Case ¹ o Three pillars provide defense-in-depth safety case o Real-Time Monitoring

Approach temperature description and figure

Link to Section 13.5.2 procedure development o Physical Inspections

Augmented examination requirements for SG tubes and IFRs o Added IFR loss coefficient range

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ASME American Society of Mechanical Engineers CITF Containment Isolation Test Fixture CIV Containment Isolation Valve CNV Containment Vessel COL Combined Operating License DC Design Certification DHRS Decay Heat Removal System DWO Density Wave Oscillation ECCS Emergency Core Cooling System FSAR Final Safety Analysis Report IFR Inlet Flow Restrictor LOCA Loss-of-Coolant Accident NRC Nuclear Regulatory Commission PTLR Pressure-Temperature Limits Report RAI Request for Additional Information RCS Reactor Coolant System RPV Reactor Pressure Vessel RSV Reactor Safety Valve SG Steam Generator SDAA Standard Design Approval Application

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