ML24313A066

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LLC - Response to NRC Request for Additional Information No. 015 (RAI-10133 R1) on the NuScale Standard Design Approval Application
ML24313A066
Person / Time
Site: 05200050
Issue date: 11/08/2024
From: Shaver M
NuScale
To:
Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML24313A065 List:
References
RAIO-175675
Download: ML24313A066 (1)
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RAIO-175675 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com November 08, 2024 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information No. 015 (RAI-10133 R1) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter to NuScale, Request for Additional Information No. 015 (RAI-10133 R1), dated March 15, 2024 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The enclosure to this letter contains NuScale's response to the following RAI question from NRC RAI-10133 R1:

3.13-1 is the proprietary version of the NuScale response to NRC RAI No. 015 (RAI-10133 R1, Question 3.13-1). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirement of 10 CFR § 810. Enclosure 2 is the nonproprietary version of the NuScale response.

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

If you have any questions, please contact Elisa Fairbanks at 541-452-7872 or at efairbanks@nuscalepower.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 08, 2024.

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

RAIO-175675 Page 2 of 2 11/08/2024 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Prosanta Chowdhury, Senior Project Manager, NRC

NuScale Response to NRC Request for Additional Information RAI-10133 R1, Question 3.13-1, proprietary : NuScale Response to NRC Request for Additional Information RAI-10133 R1, Question 3.13-1, nonproprietary : Affidavit of Mark W. Shaver, AF-175676

RAIO-175675 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10133 R1, Question 3.13-1, proprietary

RAIO-175675 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10133 R1, Question 3.13-1, nonproprietary

Response to Request for Additional Information Docket: 052000050 RAI No.: 10133 Date of RAI Issue: 03/15/2024 NRC Question No.: 3.13-1 Regulatory Basis

  • 10 CFR Part 50 Appendix-A, GDC 4, Environmental and dynamic effects design bases, states that structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
  • 10 CFR 50.46 Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors describe, in part, that emergency core cooling systems cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.

Issue Standard Review Plan (SRP) Section 3.13, Threaded Fasteners - ASME Code Class 1, 2, and 3, provides guidance for reviewing and evaluating the adequacy of an applicants criteria in regard to selection of materials, design, inspection, and testing of its threaded fasteners (i.e.,

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threaded bolts, studs, etc.) prior to initial service and during service in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1, 2 or 3 systems.

The staff noted that a change in the design of connections with Reactor Pressure Vessel from the DCA to the SDA. The DCA design utilized primarily welded connections with a few bolted connections for a limited number of connections, while the SDAA design utilizes primarily bolted connections with threaded inserts. This change impacts safety-related ASME BPV Code Section III Class 1 components that constitute the reactor coolant pressure boundary. Due to the safety significant nature of these connections and the significant change in the SDAA design from that reviewed in the DCA, the staff needs additional information to support its safety review of the structural integrity of these connections.

The RVV and RRV connections to the RPV are the most safety significant valve connections in the ECCS. Gross failure at these locations would result in a significant challenge to fuel integrity and would result in unanalyzed dynamic loads that may challenge the containment boundary. If breaks at these connections are not assumed, then appropriate justification must be provided.

Using risk-informed considerations, the decrease in defense-in-depth due to not postulating losses of coolant at these locations can be compensated, in part, by a high confidence in the structural integrity of the bolted connections. According to GDC 4, dynamic effects associated with postulated pipe ruptures can be excluded from the design basis when analyses that demonstrate that the probability of fluid system piping rupture is extremely low under design basis conditions are reviewed and approved by the Commission. This can be satisfied, in part, by confirming that stress and cumulative usage factors (CUF) margins meet at least the BTP3-4 criteria using acceptable analysis methods, having appropriate augmented measures in materials, fabrication, and preservice and inservice inspections, and continuing to meet the guidelines and recommendations in EPRI NP-5769 [Degradation and failure of Bolting in Nuclear Power Plants].

During an audit meeting, NuScale stated that they would complete a Finite Element Analyses (FEA) to determine the stress and CUF values for bolts, threaded inserts, and other components in these connections, but has indicated that this analysis would not be completed until the ITAAC phase. The staff does not consider regulatory commitments to complete the design evaluation and meet the ASME BPV Code acceptance criteria through an ITAAC as an appropriate resolution for these highly safety significant ECCS RVV and RRV connections due to the far-reaching implications from 10 CFR Appendix A, GDC 4, and 10 CFR 50.46 perspectives.

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In order to reach a safety finding for the SDAA review, the NRC is requesting a summary of results for the current design configuration Due to as-built configurations, the NRC realizes that minor changes may occur in the final design of ECCS valve bolted connections. Therefore, the final as-built FEA analysis or a reconciliation for the ECCS Valve bolted connections analyses can be submitted during the ITAAC phase.

Information Requested Provide a summary of the method and assumptions used in the FEA (or other analysis methodology if FEA is not used). Additionally, provide a summary of preliminary analysis results of representative bolted connection from each group (RSV, RRV, RVV, I&C, feed plenum access port, PZR Heater cover, and MS Plenum). Include ASME BPV Code Section III Class 1 stress intensities and CUFs for all the key components of the bolted connections (bolting, threaded inserts, flanges, and others), based on the current design of the bolted connections.

This FEA evaluation (or other analysis methodology) should be made available for staff audit.

The discussion should address the inputs, boundary conditions, and margins in stresses and CUF values for threaded inserts, bolts, flanges, and other key components. Revise the FSAR as appropriate with the summary of the analysis results.

NuScale Response:

Summary and Preface:

As discussed in the response to RAI 3.9.4-9, only the control rod drive mechanism (CRDM) pressure housing and CRDM support structure connections changed from a welded connection in the US600 design to a bolted connection in the US460 design. The threaded insert design for bolted flange connections maintain the same basis as established in the US600 certified design as shown in Table 1 of RAI 3.9.4-9. NuScale makes this clarification because the following statement made in this Request for Additional Information (RAI) 3.13-1 incorrectly compares the US600 to US460 design:

The DCA design utilized primarily welded connections with a few bolted connections for a limited number of connections, while the SDAA design utilizes primarily bolted connections with threaded inserts.

In a May 7th and 8th 2024 meeting discussing the potential closure of RAI 3.13-1 and 3.9.4-9, NuScale and the staff agreed that a demonstration of the emergency core cooling system NuScale Nonproprietary NuScale Nonproprietary

(ECCS) reactor vent valve (RVV) and reactor recirculation valve (RRV) threaded insert performance to code stress and fatigue limits closes the concern of approving a standard design usage of threaded inserts. The results of the finite element model (FEM) for the US600 include stress classification lines (SCLs) for the threaded insert locations in the inputs. NuScale extracted this information from the US600 calculation, and applied US460 specific transient loads for fatigue and stress calculations. This engineering evaluation demonstrates margin on the connections as an example of application of the threaded insert standardized design for staff assurance and the corresponding safety finding.

The FEM analysis of the RVV and RRV connections demonstrate a representative analysis to assure design finality consistent with the Commissions standardization objectives. For the purposes of review, the other connections described in the RAIthe reactor safety valve (RSV),

instrumentation and controls (I&C), feed plenum access port, pressurizer (PZR) heater cover, and main steam (MS) plenumare examples of components whose for which code compliance if adequate to demonstrate their safety function and achieve design finality. The RRV and RVV assemblies, however, provide a unique safety-related function for the US600 and US460 which requires assurance via preliminary analysis for standardization objectives.

For the connections between components such as welding and flanges, the code directs the owner to select the appropriate connection type and then evaluate the specified design against established acceptance criteria. The usage of threaded inserts is permitted by the ASME Code, and the design analysis provides confidence the final assembly of the system will fulfill design functions and maintain structural and functional integrity over the life of the plant.

In the current preliminary design phase, demonstration of connection performance for the RSV, I&C, feed plenum access port, PZR heater cover, and MS plenum is not critical for the standardization of the design. The US600 design certification came to the safety conclusion for threaded inserts without preliminary FEM calculations at these locations for CUF or stress, which provides a valid example for the US460 review for the other flange connections and corresponding threaded inserts. The final ASME design and data report for Class 1, 2, and 3 SSC completes assurance of a final plant-specific designs safety, which is further discussed in RAI 3.9.4-9. Inspection, Test, and Analysis Acceptance Criteria (ITAAC) 02.01.01.ii and 02.01.02 ensure the final plant specific design and installed components fulfill the requirements as approved by the staff review in the US460 SDAA.

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ASME Evaluation Study To provide reasonable assurance on the design of the reactor pressure vessel (RPV) bolted connections, a representative threaded insert stress and fatigue evaluation was performed for the safety-significant emergency core cooling system (ECCS) reactor vent valves (RVVs) and reactor recirculation valves (RRVs). The evaluation utilized the US600 RPV geometry with updates for US460 materials and loadings. A reconciliation was performed to demonstrate that the US600 geometry produces stresses in the threaded inserts, which are similar to the US460 geometry. Previous analyses conducted during the US600 review demonstrated with the finite element models (FEMs) and corresponding stress classification lines (SCLs) sufficient margin of safety for the preliminary design. These US600 FEMs can be used for the representative US460 threaded inserts evaluation with a reconciliation performed for the evaluated differences in geometries. Specifically, the RVVs on the US460 connect to a flat top head, while the US600 connect to an elliptical head. A comparison of the geometric parameters relevant to the stress distributions in the threaded inserts for these connections between the US600 and US460 designs shows that the variation in geometric parameters does not exceed 8% between the configurations. The use of the US600 geometry is therefore considered representative of the US460 design.

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Table 1: Valve Geometry Comparison ((2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 1: US600 Reactor Recirculation Valve Connection (( }}2(a),(c),ECI NuScale Nonproprietary NuScale Nonproprietary

Figure 2: US600 Reactor Vent Valve Connection (( }}2(a),(c),ECI The US460 and US600 reactor pressure vessels use the same materials with the exception of the threaded inserts. In the evaluation using the US600 model the threaded insert material was changed to reflect the US460 material. NuScale Nonproprietary NuScale Nonproprietary

Table 2: Vessel and Valve Material Comparison Valve Component US600 Material US460 Material RRV RPV Shell SA-508 Grade 3 Class 2 SA-508 Grade 3 Class 2 RPV Shell Cladding 308L/309L 308L/309L Valve Body SA-182 F316/316L SA-182 F316/316L Studs, Nuts, and Washers SB-637 Alloy 718 SB-637 Alloy 718 Threaded Inserts SA-479 Type 304 (0.03% Max Carbon) SA-193 Grade B8R RVV RPV Shell SA-508 Grade 3 Class 2 SA-508 Grade 3 Class 2 RPV Shell Cladding 308L/309L 308L/309L Valve Body SA-182 F316/316L SA-182 F316/316L Studs, Nuts, and Washers SB-637 Alloy 718 SB-637 Alloy 718 Threaded Inserts SA-479 Type 304/304L SA-193 Grade B8R The fatigue evaluation considers Level A and B transients according to XIII-3500 of the ASME BPVC, Section III, Mandatory Appendix XIII. The analysis conservatively includes all Level C transients. NuScale Nonproprietary NuScale Nonproprietary

Table 3: Transients Applicable to the Reactor Recirculation Valves and Reactor Vent Valves Valve ID Name Service Level Events RRV HTS Reactor Heatup to Hot Standby A 200 RCD Reactor Cooldown from Hot Standby A 200 PAC Power Ascent from Hot Standby A 700 PWD Power Descent to Hot Standby A 300 LFW Load Following A 19,750 RIL Load Ramp Increase A 2000 RDL Load Ramp Decrease A 2000 SIL Step Load Increase A 3000 SDL Step Load Decrease A 3000 LLD Large Step Load Decrease A 200 BLT (1) Bolting/Unbolting A 80 FTD Decrease in Feedwater Temperature B 180 ISF Increase in Secondary Flow B 30 TTX Turbine Trip without Bypass B 90 TTB Turbine Trip with Bypass B 180 LOP Loss of Normal AC Power B 60 MC2 Inadvertent MSIV Closure B 30 IOD Inadvertent Operation of the DHRS B 15 TFW Reactor Trip from Full Power - FW B 100 TD1 Reactor Trip from Full Power - DHRS B 20 TD2 Reactor Trip from Full Power - DHRS no Decay Heat B 5 DRP Control Rod Misoperation B 60 PZM Inadvertent Pressurizer Spray B 15 COP Cold Overpressure Protection B 30 CMT CVCS Malfunctions - Trip B 15 OBE (2) Operating Basis Earthquake B 312 (3) RRV Spurious ECCS Valve Actuation C 5 (4) RVV RSV Inadvertent Opening of a RSV C 5 SML CVCS Pipe Break C 5 SFb Steam Generator Tube Failure C 5 RVV HTS Reactor Heatup to Hot Standby A 200 RCD Reactor Cooldown from Hot Standby A 200 PAC Power Ascent from Hot Standby A 700 NuScale Nonproprietary NuScale Nonproprietary

Valve ID Name Service Level Events PWD Power Descent to Hot Standby A 300 RIL Load Ramp Increase A 2000 RDL Load Ramp Decrease A 2000 SIL Step Load Increase A 3000 SDL Step Load Decrease A 3000 LLD Large Step Load Decrease A 200 BLT (1) Bolting/Unbolting A 80 FTD Decrease in Feedwater Temperature B 180 ISF Increase in Secondary Flow B 30 TTX Turbine Trip without Bypass B 90 TTB Turbine Trip with Bypass B 180 LOP Loss of Normal AC Power B 60 MC2 Inadvertent MSIV Closure B 30 IOD Inadvertent Operation of the DHRS B 15 TD1 Reactor Trip from Full Power - DHRS B 20 TD2 Reactor Trip from Full Power - DHRS no Decay Heat B 5 PZM Inadvertent Pressurizer Spray B 15 CMT CVCS Malfunctions - Trip B 15 COP Cold Overpressure Protection B 30 DRP Control Rod Misoperation B 60 TFW Reactor Trip from Full Power - FW B 100 OBE (2) Operating Basis Earthquake B 312 (3) RRV Spurious ECCS Valve Actuation C 5 (4) RVV RSV Inadvertent Opening of a RSV C 5 SML CVCS Pipe Break C 5 SFb Steam Generator Tube Failure C 5 Notes: (1) The BLT transient is a zero-stress state transient to account for stud tensioning and de-tensioning cycles. (2) The RPV ASME Design Specification states that seismic loads shall be evaluated at normal operating conditions. The stress and temperature results for the PZM transient at 10 seconds are used to represent OBE in the fatigue evaluation since this transient at this time point occurs at normal operating conditions. (3) The RPV ASME Design Specification states that fatigue analyses may consider 20 full SSE vibratory cycles or 312 fractional amplitude SSE cycles (i.e., OBE cycles). (4) According to the RPV Loading Specification, there are a total of five events for all ECCS valve actuations. At locations where data is provided for both RRV and RVV actuation, use the most limiting data for five events only. NuScale Nonproprietary NuScale Nonproprietary

The fatigue analysis is based on the rules provided in Subsubarticle XIII-4230 of the ASME BPVC, Section III, Mandatory Appendix XIII. The fatigue curves applicable to this calculation follow those taken for SA-193 Grade B8R alloy as represented in Table I-9.0 and Figure I-9.4 of the ASME BPVC, Section III, Mandatory Appendix I. These fatigue curves are modified during post processing to consider the effect of elastic modulus per Subparagraph XIII-4230(d). Specifically, the alternating stress is multiplied by the ratio of the elastic modulus given on the fatigue curve to the elastic modulus at the analysis temperature. A fatigue strength reduction factor (FSRF) of 4.0 is also applied to the alternating stress at the inner and outer surfaces of the threaded inserts according to Subparagraph XIII-4230(c). The fatigue evaluation does not account for the effects of the light water reactor (LWR) water environments according to NRC RG 1.207 and NUREG/CR-6909. These define a LWR water environment as any transient or steady state environment in a light water commercial nuclear power plant where the component of interest is exposed to water above 50 C (122 F). Equation 44 of NUREG/CR-6909 Rev 1 reports for temperatures below 100 C (212 F) the T* variable in the austenitic stainless steel Fen equations is equal to zero. The threaded inserts are not exposed to LWR water during steady-state operation and most Levels A and B transient conditions. They may be exposed to water during transients where the containment vessel (CNV) is flooded during the Reactor Heatup to Hot Standby transient (HTS) or the Reactor Cooldown from Hot Standby (RCD) transients; however, this water is not LWR water. The threaded inserts may be exposed to LWR water during ECCS actuation. The following transients in which the threaded inserts may be exposed to water are:

Reactor Heatup to Hot Standby (HTS)

Reactor Cooldown from Hot Standby (RCD)

Cold Overpressure Protection (COP)

Loss of Normal AC Power (LOP)

Reactor Trip from Full Power - DHRS (TD1) (10 cycles only)

Spurious ECCS Valve Actuation (RRV, RVV)

Inadvertent Opening of a RSV (RSV)

CVCS Pipe Break (SML) The temperatures when the CNV is flooded due to HTS or RCD is below the 212 F temperature threshold such that the environmental fatigue correction factor (Fen) is equal to 1.0. Jet loading due to ECCS actuation that is considered will result in a high strain rate such that the environmental fatigue correction factor for the RVV for this condition would be equal to NuScale Nonproprietary NuScale Nonproprietary

1.0. Therefore, no environmental fatigue correction factors (Fen) are considered in this evaluation. Subparagraph XIII-4230(e) of the ASME BPVC, Section III, Mandatory Appendix XIII requires that the CUF for Levels A and B loadings shall not exceed 1.0. In addition, the maximum nominal stress shall not exceed 2.7 times the allowable stress intensity for the material at service temperature (2.7Sm) to enable the use of the higher of the two design fatigue curves in Figure I-9.4 of the ASME BPVC, Section III, Mandatory Appendix I. Membrane and bending stresses are evaluated along selected path lines through the thickness of components, which are referred to as stress classification lines (SCLs). The location and orientation of SCLs comply with the guidelines in Annex 5-A of ASME BPVC, Section VIII, Division 2. The SCL locations were identified based on a review of stress contour plots. Figure 3 provides example stress intensity contour plots from the RVV FEM for sample peak stress time from a pressure plus thermal stress structural analysis and pressure stress structural analysis, respectively. The maximum peak stresses were observed to occur at the top of the threaded inserts, which are in contact with the surface of the RPV cladding. However, SCLs were placed two elements from the top of the inserts based on the following reasoning:

The peak stresses are elevated on the top of the insert due to the sharp edges at the material discontinuity between the bolts, inserts, RPV cladding, and RPV base metal.

Thread engagement at the top of the inserts, which are in contact with cladding, are not critical to the structural integrity of the joint. However, thread engagement at locations of the inserts in contact with the RPV base metal, where the SCLs in Figure 4 are defined, are critical to the structural integrity of the joint. NuScale Nonproprietary NuScale Nonproprietary

Figure 3: Example Reactor Vent Valve Finite Element Model Stress Intensity Contour (( }}2(a),(c),ECI NuScale Nonproprietary NuScale Nonproprietary

Figure 4: Reactor Vent Valve Finite Element Model Stress Classification Line Locations (( }}2(a),(c),ECI The selected SCL locations represent the highest stress locations observed in the portions of the inserts in contact with the RPV base metal. The evaluation of fatigue at these SCLs ensure proper thread engagement between the bolts, threaded inserts, and RPV base metal and structural integrity of the joint over the design life. Table 4 summarizes the results for the most limiting RRV and RVV threaded insert locations. Table 4: Summary of Threaded Insert Stress and Fatigue Results (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Operational Inspections and Programs The threaded inserts are designed to ensure a reliable connection to the RPV throughout the 60-year service life of the US460 design. Their design incorporates a stainless steel material and a seal weld that is resistant to boric acid corrosion, similar to the stainless steel composition of the RPV exterior adjacent cladding. Given the above results and expected stresses under normal operations, there is no postulated degradation mechanism for the threaded inserts and corrosion resistant barrier. In order to provide reasonable assurance the threaded inserts do not degrade over time, augmented examination requirements are applied to the threaded inserts and seal welds. A VT-1 examination shall be performed in accordance with ASME BPVC, Section XI, IWA-2211 for surface defects or potential corrosion whenever an ASME Class 1 component with a threaded insert or an ASME Class 1 support with a threaded insert is disassembled. Further, a Boric Acid Control Program shall be implemented by a COL applicant as designated in COL Item 5.2-3. Leakage will be monitored with visual or nondestructive inspections during plant operations. Conclusion The use of threaded inserts for the RVV and RRV bolted connections did not change from the US600 and US460 designs. The design evaluation presented above demonstrates sufficient design margin for the US460 RVV and RRV threaded inserts. This analysis represents an example connection in the US460 design showing design margin and can be considered representative for other US460 bolted connections with threaded inserts. In addition, the Boric Acid Control Program and threaded insert VT-1 examinations during operations provide assurance that degradation does not occur at the connections utilizing threaded inserts and the reactor coolant pressure boundary integrity is maintained. Impact on US460 SDAA: FSAR Section 3.13 has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Threaded Fasteners (American Society of Mechanical Engineers Code Class 1, 2, and 3) NuScale US460 SDAA 3.13-4 Draft Revision 2 Class 1, 2, and 3 threaded fasteners are retained in accordance with 10 CFR 50.71. 3.13.2 Inservice Inspection Requirements Audit Question A-15.6.5-13 Inservice Inspection for ASME Class 1, 2, and 3 threaded fasteners is in accordance with the ASME BPVC, Section XI (Reference 3.13-5), as required by 10 CFR 50.55a, except where specific written relief is granted by the Nuclear Regulatory Commission. or where augmented inspection requirements that exceed the ASME Code requirements are specified. COL Item 3.13-1: An applicant that references the NuScale Power Plant US460 standard design will provide an inservice inspection program for American Society of Mechanical Engineers Class 1, 2, and 3 threaded fasteners. The program will identify the applicable edition and addenda of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI and ensure compliance with 10 CFR 50.55a. RAI 3.13-1 3.13.2.1 Augmented Examination Requirements The NuScale design applies augmented examination requirements to ASME Class 1 threaded inserts and seal welds. A VT-1 examination shall be performed in accordance with ASME BPVC, Section XI, IWA-2211 for surface defects or potential corrosion whenever an ASME Class 1 component with a threaded insert or an ASME Class 1 support with a threaded insert is disassembled. VT-1 examinations on the Class 1 threaded inserts or Class 1 seal welds that reveal cracks or corrosion shall be extended to include additional examinations in accordance with IWB-2430(a)(1). The additional examinations shall require a population of the CRDM, RRV, and RVV connection threaded inserts and seal welds. 3.13.3 References 3.13-1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 edition, Section III, Rules for Construction of Nuclear Facility Components, New York, NY. 3.13-2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 edition, Section II, Materials, New York, NY. 3.13-3 McIlree, A.R., Degradation of High Strength Austenitic Alloys X-750, 718 and A286 in Nuclear Power Systems, Proceedings of the 1st International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, National Association of Corrosion Engineers, 1984.

RAIO-175675 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Affidavit of Mark W. Shaver, AF-175676

AF-175676 Page 1 of 2

NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the response by which NuScale develops its NuScale Power, LLC Response to NRC Request for Additional Information (RAI-10133 R1, Question 3.13-1) on the NuScale Standard Design Approval Application. NuScale has performed significant research and evaluation to develop a basis for this response 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, NuScales 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 NuScales 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 response to NRC Request for Additional Information RAI-10133 R1, Question 3.13-1. 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-175676 Page 2 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 NuScales technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on November 08, 2024. Mark W. Shaver}}

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